Fitness-for-Service Assessment of Steam Generator Tubing Subject to Dealloying Degradation

2017 ◽  
Author(s):  
Rosita Mousavi ◽  
Xinjian Duan ◽  
Michael Kozluk ◽  
Min Wang ◽  
Yihai Shi
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Life Cycle Management ◽  
Steam Generators ◽  
Nickel Copper ◽  
Steam Generator Tubing ◽  
Management Plans ◽  
Monel 400 ◽  
Technical Basis ◽  
Steam Generator Tubes

A new degradation mechanism has been observed in Monel 400 Steam Generator tubing material, a nickel-copper alloy (63Ni-28Cu-2½Fe) with the ASME material designation SB-163/N04400. The location is above the top preheater support plate of the two re-circulating steam generator in one of the units of the Pickering Nuclear Generating Station. This paper provides a brief description of the regulatory environment, OPG’s steam generator life cycle management plans, the Canadian Industry’s fitness-for-service guidelines for steam generator tubes, and the afflicted steam generators. The paper then goes on to discuss the following activities that were conducted to support the technical basis to justify that the steam generators fit to be returned to service: • Inspection scope expansion, methods, and results. • Examination of removed tubes. • Condition monitoring assessment. • Operational assessment. • Burst-pressure tests of removed tubes and of fabricated test specimens. • Degradation specific flaw model and acceptance standards. • Flaw growth rate predictions. • Plugging limit adopted.

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.

scholarly journals Reliability of Sampling Inspection Schemes Applied to Replacement Steam Generators

2006 ◽  
Vol 129 (1) ◽  
pp. 109-117 ◽  
Author(s):  
Guy Roussel ◽  
Leon Cizelj
Keyword(s):  
Closed Form ◽  
Steam Generator ◽  
The State ◽  
The Other ◽  
Prior Density ◽  
Steam Generators ◽  
Numerical Examples ◽  
Closed Form Solutions ◽  
Sampling Inspection ◽  
Steam Generator Tubing

The basis for determining the size of the random sample of tubes to be inspected in replacement steam generators is revisited in this paper. A procedure to estimate the maximum number of defective tubes left in the steam generator after no defective tubes have been detected in the randomly selected inspection sample is proposed. A Bayesian estimation is used to obtain closed-form solutions for uniform, triangular, and binomial prior densities describing the number of failed tubes in steam generators. It is shown that the particular way of selecting the random inspection sample (e.g., one sample from both SG, one sample from each SG, etc.) does not affect the results of the inspection and also the information obtained about the state of the uninspected tubing, as long as the inspected steam generators belong to the same population. Numerical examples further demonstrate two possible states of the knowledge existing before the inspection of the tubing. First, virtually no knowledge about the state of the steam generator tubing before the inspection is modeled using uniform and triangular prior densities. It is shown that the knowledge about the uninspected part of the tubing strongly depends on the size of the sample inspected. Further, even small inspection samples may significantly improve our knowledge about the uninspected part. On the other hand, rather strong belief on the state of the tubing prior to the inspection is modeled using binomial prior density. In this case, the knowledge about the uninspected part of the tubing is virtually independent on the size of the sample. Furthermore, it is shown qualitatively and quantitatively that such inspection brings no additional knowledge on the uninspected part of the tubing.

Degradation Mechanism of Dissimilar Metal Weld Joints on Steam Generator Collectors VVER 440MW Type

2017 ◽  
Vol 741 ◽  
pp. 134-137
Author(s):  
Lubomír Junek ◽  
Ladislav Jurasek ◽  
Zdeněk Čančura ◽  
Miroslava Ernestová ◽  
Zuzana Skoumalová
Keyword(s):  
Steam Generator ◽  
Experimental Analysis ◽  
Degradation Mechanism ◽  
Steam Generators ◽  
Dissimilar Metal ◽  
Weld Joints ◽  
Dissimilar Metal Weld ◽  
Metal Weld

Indications were detected on dissimilar metal welds (DMW) of steam generators (SG) after 20 years of operation during NDT inspections. Indications started slowly growth every year. DMW on SGs had to be repaired. Paper describes experimental analysis and degradation mechanism of SG weld joints failures.

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.

The Effect of Thermal Treatments on Stress Corrosion Cracking of Alloy-800 in a Caustic Environment

CORROSION ◽  
1985 ◽  
Vol 41 (10) ◽  
pp. 575-581 ◽  
Author(s):  
R. S. Pathania ◽  
R. D. Cleland
Keyword(s):  
Stress Corrosion ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Caustic Solution ◽  
Thermal Treatments ◽  
Steam Generators ◽  
Alloy 800 ◽  
Steam Generator Tubing ◽  
Alkaline Environments

Abstract Stress corrosion cracking (SCC) tests under controlled potentials were conducted on plastically stressed C-rings of Alloy-800 to evaluate the effect of thermal treatments (at 500 to 850 C) on SCC of Alloy-800 steam generator tubing. The tests were conducted in a caustic solution at 300 C; such alkaline environments may sometimes form in crevices in nuclear steam generators. The results are presented in the form of a time-temperature-SCC diagram. Thermal treatments generally inhibited the SCC of Alloy-800, and the improvement was substantial in a few cases. Application of anodic potential and increasing stress generally accelerated SCC.

Stress Corrosion Cracking of Steam Generator Tube Materials in Sodium Hydroxide Solutions

CORROSION ◽  
1978 ◽  
Vol 34 (11) ◽  
pp. 369-378 ◽  
Author(s):  
R. S. PATHANIA ◽  
J. A. CHITTY
Keyword(s):  
Stress Corrosion ◽  
Sodium Hydroxide ◽  
Applied Stress ◽  
Steam Generator ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Sodium Hydroxide Concentration ◽  
Inconel 600 ◽  
Steam Generator Tubing ◽  
Monel 400

Abstract Stress corrosion cracking (SCC) tests were carried out on specimens of Monel 400, Inconel 600, and Sanicro 30 steam generator tubing in solutions containing 10 to 500 g NaOH/kg H2O at 300 C for times up to 600 days. Applied stress and sodium hydroxide concentration had a significant effect on the (SCC) resistance of the three materials.

Experimental Investigation of Subcooled Choking Flow in a Steam Generator Tube Crack

2016 ◽  
Author(s):  
Hung Nguyen ◽  
Mark Brown ◽  
Shripad T. Revankar ◽  
Jovica Riznic
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Steam Generator ◽  
Nuclear Power ◽  
Test Facility ◽  
Experimental Program ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
Loss Of Coolant ◽  
Steam Generator Tubes

Steam generator tubes have a history of small cracks and even ruptures, which lead to a loss of coolant from the primary side to the secondary side. These tubes have an important role in reactor safety since they serve as one of the barriers between radioactive and non-radioactive materials of a nuclear power plant. A rupture then signifies the loss of the integrity of the tube itself. Therefore, choking flow plays an integral part not only in the engineered safeguards of a nuclear power plant, but also to everyday operation. There is limited data on actual steam generators tube wall cracks. Here experiments were conducted on choked flow of subcooled water through two samples of axial cracks of steam generator tubes taken from US PWR steam generators. The purpose of the experimental program was to develop database on critical flow through actual steam generator tube cracks with subcooled liquid flow at the entrance. The knowledge of this maximum flow rate through a crack in the steam generator tubes of a pressurized water nuclear reactor will allow designers to calculate leak rates and design inventory levels accordingly while limiting losses during loss of coolant accidents. The test facility design is modular so that various steam generator tube cracks can be studied. Two sets of PWR steam generators tubes were studied whose wall thickness is 1.285 mm. Tests were carried out at stagnation pressure up to 6.89 MPa and range of subcoolings 16.2–59°C. Based on these new choking flow data, the applicability of analytical models to highlight the importance of non-equilibrium effects was examined.

Evaluation of Interaction Effects on Part-Through Cracks in Steam Generator Tubes

2013 ◽  
Author(s):  
Yong-Seok Kang ◽  
Hong-Deok Kim ◽  
Kuk-Hee Lee ◽  
Jai-Hak Park
Keyword(s):  
Steam Generator ◽  
Interaction Effect ◽  
Structural Integrity ◽  
Structural Parameters ◽  
Interaction Effects ◽  
Multiple Cracks ◽  
Performance Limit ◽  
Steam Generator Tubing ◽  
Steam Generator Tubes

Degraded steam generator tubing can affect its safety functions. Therefore, its integrity should be maintained for each degradation form and all detected degradation must be assessed to verify that if adequate integrity is retained. Determination of tube integrity limits includes identifying acceptable structural parameters such as flaw length, depth, and amplitude of signals. If we consider just single-cracked tubes, short and deep flaws are not likely to threaten structural integrity of tubes. But if it has multiple-cracks, we have to consider interaction effects of multiple adjacent cracks on its burst pressure. Because adjacent multiple cracks can be merged due to the crack growth then it can challenge against the structural performance limit. There are some studies on the interaction effects of adjacent cracks. However, existing works on the interaction effect consider only through-wall cracks. No study has been carried out on the interaction effects of part-through cracks. Most cracks existing in real steam generator tubing are not through-wall cracks but part-through cracks. Hence, integrity of part-through cracks is more practical issue than that of through-wall cracks. This paper presents experimental burst test results with steam generator tubing for evaluation of interaction effects with axial oriented two collinear and parallel part-through cracks. The interaction effect between two adjacent cracks disappeared when the distance exceeds about 2 mm.

Technical Basis for Structural Integrity Performance Criterion for Steam Generator Tubing

2006 ◽  
Author(s):  
Russell C. Cipolla ◽  
James A. Begley ◽  
Robert F. Keating
Keyword(s):  
Steam Generator ◽  
Nuclear Energy ◽  
Structural Integrity ◽  
Performance Criterion ◽  
Major Fraction ◽  
Reactor Coolant ◽  
Steam Generator Tubing ◽  
Accident Conditions ◽  
Technical Basis ◽  
Repair Techniques

General Design Criteria (GDC) 1, 2, 4, 14, 30, 31 and 32 of 10 CFR Part 50, Appendix A, define requirements for the reactor coolant pressure boundary (RCPB) with respect to structural and leakage integrity [1]. Steam generator tubing and tube repairs constitute a major fraction of the RCPB surface area. Steam generator tubing and associated repair techniques and components, such as sleeves, must be able to maintain reactor coolant inventory and pressure. The Structural Integrity Performance Criterion (SIPC) from Nuclear Energy Institute (NEI) 97-06 [2] was developed to provide reasonable assurance that a steam generator tube will not burst during normal or postulated accident conditions. This paper presents the SIPC and its technical basis.

Probabilistic Assessment of Fretting Wear in Steam Generator Tubes Under Flow Induced Vibrations

2003 ◽  
Author(s):  
Greg D. Morandin ◽  
Richard G. Sauve´
Keyword(s):  
Steam Generator ◽  
Degradation Mechanism ◽  
Probabilistic Methods ◽  
Fretting Wear ◽  
Nonlinear Flow ◽  
Impact Wear ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Wear Damage ◽  
The Impact

Successful life management of steam generators requires an ongoing operational assessment plan to monitor and address all potential degradation mechanisms. A degradation mechanism of concern is tube fretting as a result of flow-induced vibration. Flow induced vibration predictive methods routinely used for design purposes are based on deterministic nonlinear structural analysis techniques. In previous work, the authors have proposed the application of probabilistic techniques to better understand and assess the risk associated with operating power generating stations that have aging re-circulating steam generators. Probabilistic methods are better suited to address the variability of the parameters in operating steam generators, e.g., flow regime, support clearances, manufacturing tolerances, tube to support interactions, and material properties. In this work, an application of a Monte Carlo simulation to predict the propensity for fretting wear in an operating re-circulation steam generator is described. Tube wear damage is evaluated under steady-state conditions using two wear damage correlation models based on the tube-to-support impact force time histories and work rates obtained from nonlinear flow induced vibration analyses. Review of the tube motion in the supports and the impact/sliding criterion shows that significant tube damage at the U-bend supports is a result of impact wear. The results of this work provide the upper bound predictions of wear damage in the steam generators. The EPRI wear correlations for sliding wear and impact wear indicate good agreement with the observed damage and, given the preponderance of wear sites subject to impact, should form the basis of future predictions.

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