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.

Steam Generator In-Service Inspections at Paks NPP

2008 ◽  
Author(s):  
Miklo´s Do´czi
Keyword(s):  
Heat Exchanger ◽  
Steam Generator ◽  
Eddy Current ◽  
Power Plants ◽  
Tube Bundle ◽  
Degradation Mechanism ◽  
Feed Water ◽  
Material Damage ◽  
Steam Generators ◽  
Service Inspection

Steam Generator is one of the most critical components in nuclear power plants. It has of overriding importance from point of view of safe and reliable operation of the whole plant. Variety of degradation mechanisms affecting SG tube bundle may cause different types of material damage. In Paks NPP eddy current in-service inspection have been performed since 1988. In the year 1997 higher number of defected tubes were found in case of Unit#2, compared to results of the previous years. A medium term SG inspection program had been performed in the time period between 1998–2004. Based on the results of eddy current inspections high number of heat exchanger tubes had been plugged. Chemical cleanings of all steam generators were performed aiming to reduce the magnetite, copper deposits and corrosion agents acting on the surface of the tubes. Replacement of the main condensers had been performed to stop the uncontrolled water income caused by the relatively frequent leakages of the condenser tubes. Several tube samples had been cut from the first row of the tube bundles of different steam generators to study the effectiveness of the cleaning process and to determine the composition of deposits on the tube outside surface. Also several tubes with eddy current indications had been pulled out from the steam generators to determine the acting degradation mechanism. Examination of removed tubes can provide opportunity to check the reliability of eddy current inspection using bobbin coil. Also there were tubes pulled out form SG with existing cracks. From the year 2005 new inspection program had been started. As the first results of the new inspection program shows, there is only a few new indications had been found and there is no measurable crack propagation in case of existing indications. During the recent years feed-water collectors were replaced in case of all units of the power plant, because of material damage (erosion corrosion). The paper summarizes the results of eddy current in-service inspection of heat exchanger tubes, results of examinations of removed tubes and also deals with results of visual examination of the feed-water distributor system.

Vibration of Tube Bundles Subjected to Two-Phase Cross-Flow

1985 ◽  
Vol 107 (4) ◽  
pp. 335-343 ◽  
Author(s):  
M. J. Pettigrew ◽  
J. H. Tromp ◽  
J. Mastorakos
Keyword(s):  
Heat Exchangers ◽  
Tube Bundle ◽  
Cross Flow ◽  
Elastic Instability ◽  
Steam Generators ◽  
Two Phase ◽  
Vibration Excitation ◽  
Mass Fluxes ◽  
Excitation Mechanisms ◽  
Shell And Tube

Two-phase cross-flow exists in many shell-and-tube heat exchangers such as condensers, reboilers and nuclear steam generators. Thus we are conducting a comprehensive program to study tube bundle vibrations subjected to two-phase cross-flow. This paper presents the results of experiments on a normal-triangular and a normal-square tube bundle, both of p/d = 1.47. The bundles were subjected to air-water mixtures to simulate realistic vapor qualities and mass fluxes. Vibration excitation mechanisms were deduced from vibration response measurements. Results on damping, hydrodynamic mass, fluid-elastic instability and random turbulence excitation in two-phase cross-flow are presented.

ARTIST: Introduction and First Results

2004 ◽  
Author(s):  
S. Gu¨ntay ◽  
D. Suckow ◽  
A. Dehbi ◽  
R. Kapulla
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Aerosol Deposition ◽  
Severe Accident ◽  
Steam Generators ◽  
Steam Content ◽  
Primary Particles ◽  
Accident Management ◽  
First Results ◽  
Small Flow

ARTIST (Aerosol Trapping In a Steam Generator) is a seven-phase international project (2003–2007) which investigates aerosol and droplet retention in a model steam generator under dry, wet and accident management conditions, respectively. The test section is comprised of a scaled steam generator tube bundle consisting of 270 tubes and 3 stages, one 1:1 separator unit, and one 1:1 dryer unit. As a prelude to the ARTIST project, four tests are conducted in the ARTIST bundle within the 5th EU FWP SGTR. These first tests address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities are sonic, and far away from the break, where the flow velocities are three orders of magnitude lower. With a dry bundle and the full flow representing the break stage conditions, there is strong evidence that the TiO2 aerosols used (AMMD 2–4 μm, 32 nm primary particles) disintegrate into much smaller particles because of the sonic conditions at the break, hence promoting particle escape from the secondary and lowering the overall DF, which is found to be between 2.5 and 3. With a dry bundle and a small flow reproducing the far-field velocities, the overall bundle DF is of the order of 5, implying a DF of about 1.9 per stage. Extrapolating the results of the dry tests, it turns out that for steam generators with 9 or more stages, it is expected that substantial DF’s could be achieved when the break is located near the tube sheet region. In addition, better decontamination is expected using more representative proxies of severe accident aerosols (sticky, multicomponent particles), a topic which is yet to be investigated. When the bundle is flooded, the DF is between 45 and 5740, depending on the mass flow rate, the steam content, and the water submergence. The presence of steam in the carrier gas and subsequent condensation inside the broken tube causes aerosol deposition and blockages near the break, leading to an increase in the primary pressure. This has implications for real plant conditions, as aerosol deposits inside the broken tube will cause more flow to be diverted to the intact tubes, with a corresponding reduction in the source term to the secondary.

Flow-Induced Wear in Steam Generator Tubes—Prediction Versus Operational Experience

1998 ◽  
Vol 120 (2) ◽  
pp. 138-143 ◽  
Author(s):  
M. K. Au-Yang
Keyword(s):  
Steam Generator ◽  
Tube Wall ◽  
Tube Bundle ◽  
Cross Flow ◽  
Operational Experience ◽  
Steam Generators ◽  
Support Plate ◽  
Power Operations ◽  
Full Power ◽  
Steam Generator Tubes

Many nuclear steam generators have accumulated more than 10 effective-full-power-years of operation. Eddy-current inspections revealed that a number of these steam generator tubes, notably those located in high local cross-flow regions, have indications of wear at some support plate elevations after 5 to 10 yr of effective-full-power operations. In the last 5 yr, a number of technical papers on nonlinear tube bundle dynamics has been published to address the effect of tube and support plate interactions. At the same time, test data relating wear and tube wall thickness losses for different material combinations and different support plate geometries became available. Based on the available data in the literature, as well as data obtained in the author’s affiliation, this paper assesses the cumulative tube wall wear after 5, 10, and 15 effective-full-power years of operation of a typical commercial nuclear steam generator, using different wear models. It is hoped that this study will shed some light on the probable mechanism that caused the observed wear in today’s operating nuclear steam generators.

Flow-Induced Vibration and Fretting-Wear Performance of CANDU™ Steam Generator U-Tubes: Instrumentation

2009 ◽  
Author(s):  
Atef Mohany ◽  
Victor Janzen
Keyword(s):  
Steam Generator ◽  
Tube Bundle ◽  
Measurement Techniques ◽  
Fretting Wear ◽  
Process Conditions ◽  
Steam Quality ◽  
Steam Generators ◽  
Test Program ◽  
Two Phase ◽  
Support Design

This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU™ steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program is one of the initiatives that AECL is undertaking to demonstrate that the tube support design for future CANDU steam generators meets the stringent requirements associated with a 60 year lifetime. The main objective of the tests is to address the issue of in- and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with AVB supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper.

Steam Generator Blockage: A Thermal-Hydraulic Approach Based on CATHARE 2

2008 ◽  
Author(s):  
Alexis Vergnault ◽  
Robin Dorel ◽  
Fre´de´ric Goux
Keyword(s):  
Pressure Drop ◽  
Steam Generator ◽  
Tube Bundle ◽  
Power Level ◽  
Radiological Protection ◽  
Steam Generators ◽  
Flow Area ◽  
Flow Paths ◽  
Accident Scenarios ◽  
The Impact

In some French steam generators (SG) with broached tube support plates (TSP), the TSP flow paths have become blocked by deposits. Due to these blocked broached TSP, flow paths decrease the flow area and result in increased void fraction in the tube bundle region and pressure drop across the TSP. As a consequence, it also leads to a decrease of the recirculation ratio which can be observed through the SG level measurement. The major impacts of this phenomenon can be classified in three main categories: the mechanical impact on the TSP due to an increase of the pressure drop, the impact on the safety demonstration itself due to the less secondary side inventory that have to be considered for the thermal-hydraulic studies and finally the impact on the normal transients mainly due to the level oscillation encountered in the steam generator. This paper which focuses on the last item presents the Insitute for Radiological Protection and Nuclear Safety (IRSN) approach to calculate with CATHARE 2 [1] code the response of steam generators experiencing the blockage: both are considered transients with only a “stand alone” steam generator and transients performed with our SImulator for Post-Accident so called “SIPA” [2 to 4] which enables to simulate normal and accident scenarios. Sensitivity calculations on the blockage ratio, on the blockage distribution and on the power level are made to be able to assess this phenomenon. Some conclusions are arrived at the power levels that should be adopted to deal with blockage in order to restore the water level stability.

Impact Simulation of Process Equipment Tubes and Support Plates—A Numerical Algorithm

1987 ◽  
Vol 109 (1) ◽  
pp. 70-79 ◽  
Author(s):  
R. G. Sauve´ ◽  
W. W. Teper
Keyword(s):  
Finite Element ◽  
Heat Exchangers ◽  
Tube Bundle ◽  
Fretting Wear ◽  
Design Stage ◽  
Process Equipment ◽  
Integration Scheme ◽  
Convergence Criteria ◽  
Steam Generators ◽  
The Impact

The occurrence of flow-induced vibration fretting wear in process equipment such as heat exchangers and steam generators accounts for the majority of the failures due to vibration. One of the parameters which plays a vital role in the prediction of tube wear rate is the impact force which occurs when the free displacements of the tube exceed the clearance in the support plates, resulting in a collision. To date the determination of these impact forces reported in the literature has been restricted to simplified mathematical models which consider only straight spans of tube with gaps. The need to consider more generalized configurations has led to the development of an analytical method which simulates the nonlinear dynamic-impact response of multi-supported tubes including U-bends and the effect of nonuniform gap clearances at the supports. The approach is incorporated into a computer code based on the finite element and displacement methods using an unconditionally stable numerical integration scheme to solve the nonlinear equations of motion. The algorithm developed includes equilibrium iteration and variable time stepping based on convergence criteria, which ensures that temporal solution errors are minimized. The direct integration of the equations enables all the frequencies (subject to the finite element mesh) to be included. This is necessary since the high-frequency response at impacting may be significant. At present, the method is being used to simulate impact between tubes and support plates in steam generators and heat exchangers in order to determine tube bundle susceptibility to fretting wear failure at the design stage or operational phase. The paper describes the analytical development of the method, verification cases, and applications to the problem of tube/support plate impacting.

A FIRST APPROACH TO MODELLING A STEAM GENERATOR BASED ON HEAT EXCHANGERS NETWORK METHOD

2019 ◽  
Author(s):  
Augusto Delavald Marques ◽  
Caroline Mével ◽  
Paulo Smith Schneider ◽  
Jéssica Duarte ◽  
Guilherme Barth Rossi
Keyword(s):  
Steam Generator ◽  
Heat Exchangers ◽  
Network Method

Parametric Study of External Pressure on Steam Generator Tube Bends

2012 ◽  
Author(s):  
Mitch Hokazono ◽  
Clayton T. Smith
Keyword(s):  
Steam Generator ◽  
Parametric Study ◽  
External Pressure ◽  
Light Water Reactor ◽  
Material Strength ◽  
Steam Generators ◽  
Elliptical Cross ◽  
Light Water ◽  
Water Reactor ◽  
Steam Generator Tubes

Integral light-water reactor designs propose the use of steam generators located within the reactor vessel. Steam generator tubes in these designs must withstand external pressure loadings to prevent buckling, which is affected by material strength, fabrication techniques, chemical environment and tube geometry. Experience with fired tube boilers has shown that buckling in boiler tubes is greatly alleviated by controlling ovality in bends when the tubes are fabricated. Light water reactor steam generator pressures will not cause a buckling problem in steam generators with reasonable fabrication limits on tube ovality and wall thinning. Utilizing existing Code rules, there is a significant design margin, even for the maximum differential pressure case. With reasonable bend design and fabrication limits the helical steam generator thermodynamic advantages can be realized without a buckling concern. This paper describes a theoretical methodology for determining allowable external pressure for steam generator tubes subject to tube ovality based on ASME Section III Code Case N-759-2 rules. A parametric study of the results of this methodology applied to an elliptical cross section with varying wall thicknesses, tube diameters, and ovality values is also presented.

Analysis of phase composition of corrosion-products deposits on the surfaces of a PG-440 steam-generator tube bundle by Mossbauer spectroscopy method

2009 ◽  
Vol 56 (2) ◽  
pp. 162-163
Author(s):  
A. A. Efimov ◽  
V. G. Semenov ◽  
B. A. Gusev ◽  
M. M. Kostin ◽  
I. V. Miroshnichenko ◽  
...  
Keyword(s):  
Mössbauer Spectroscopy ◽  
Phase Composition ◽  
Steam Generator ◽  
Mossbauer Spectroscopy ◽  
Tube Bundle ◽  
Corrosion Products ◽  
Spectroscopy Method ◽  
Steam Generator Tube ◽  
Mößbauer Spectroscopy
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