Flow-Accelerated Corrosion Susceptibility Prediction of Recirculating Steam Generator Internals

2008 ◽  
Author(s):  
John M. Pietralik ◽  
Kevin L. Heppner
Keyword(s):  
Water Chemistry ◽  
Steam Generator ◽  
Transport Model ◽  
Safety Issue ◽  
Original Design ◽  
Accelerated Corrosion ◽  
Essential Components ◽  
Flow Accelerated Corrosion ◽  
Secondary Side ◽  
Chemistry Module

Steam generator (SG) components are subjected to corrosive solutions in turbulent flow. Under such conditions, actual component lifetimes may be significantly reduced from their original design lifetimes. Premature replacement of steam generator components before their expected lifetime can be very expensive. Furthermore, degradation of essential components can reduce the steam generator efficiency, thus reducing net profits. Moreover, a SG failure can also be a safety issue. One of the degradation mechanisms affecting secondary-side SG internal structural elements, which are referred to as internals, is Flow-Accelerated Corrosion (FAC). The susceptibility to FAC depends on flow parameters, water chemistry, and materials. All SG internals made of carbon steel are susceptible to FAC to varying degrees. For FAC susceptibility prediction, flow velocity, pH, and oxygen distributions are needed. SG codes, including THIRST (Thermal Hydraulic analysis In STeam generators, a computer code developed by AECL), traditionally solve for thermalhydraulic parameters. A new chemistry module has been added to THIRST, which now makes this code useful for the prediction of local water chemistry parameters in the SG. The THIRST chemistry module is comprised of a multicomponent, multiphase mass transport model coupled with a multiphase chemical equilibrium model. As input, the module requires amine concentrations in the feedwater and reheater drains. The module predicts local distributions of amine concentration in the secondary side. The concentrations predicted by the module are used to compute the pH. The chemistry module was verified against results of other work in the literature and against station blowdown data. Flow and chemistry predictions of THIRST were used to predict FAC susceptibility for internals of a SG with an integral preheater and a SG without it. Ranking of SG locations in order of FAC susceptibility was estimated from an empirical, Kastner-Riedle model. The most susceptible internals are predicted to be those in the upper section of the hot side and those on the cold side that are near the SG centre, while SG lower regions, including the integral preheater, if one exists, are better protected.

Supplementary Technical Basis for ASME Section XI Code Case N-597

2005 ◽  
Author(s):  
Doug Scarth
Keyword(s):  
Wall Thickness ◽  
Structural Integrity ◽  
Computer Code ◽  
Nuclear Regulatory Commission ◽  
Original Design ◽  
Acceptance Criteria ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Technical Basis ◽  
Flow Accelerated Corrosion

Efforts to develop clear and conservative methods to measure and evaluate wall thinning in nuclear piping have been underway since the late 1980’s. The Electric Power Research Institute (EPRI) carried out a successful campaign to address programmatic issues, such as locating and predicting flow-accelerated corrosion (FAC) degradation. This included developing a computer code (CHECWORKS), a users group (CHUG), and a comprehensive program guideline document for the effective prediction, identification and trending of flow-accelerated corrosion degradation. U.S. Nuclear Regulatory Commission (NRC) guidelines are provided in the NRC Inspection Manual Inspection Procedure 49001. At the same time, committees under Section XI of the ASME Boiler and Pressure Vessel Code have addressed evaluation of structural integrity of piping subjected to wall thinning. Code Case N-480 of Section XI provided acceptance criteria that focused on primary piping stresses, with evaluation based on a uniform wall thinning assumption for evaluating the minimum wall thickness of the piping. However, when applying this methodology to low pressure piping systems, Code Case N-480 was very conservative. Code Case N-597 was first published in 1998, and supercedes Code Case N-480. The current version is N-597-2. Code Case N-597-2 provides acceptance criteria and evaluation procedures for piping items, including fittings, subjected to a wall thinning mechanism, such as flow-accelerated corrosion. Code Case N-597-2 is a significant improvement over N-480, containing distinct elements to be satisfied in allowing the licensee to operate with piping degraded by wall thinning. The Code Case considers separately wall thickness requirements and piping stresses, and maintains original design intent margins. The Code Case does not provide requirements for locations of inspection, inspection frequency or method of prediction of rate of wall thinning. As described in the original technical basis document published at the 1999 ASME PVP Conference, the piping stress evaluation follows very closely the Construction Codes for piping. Five conditions related to industry use of Code Case N-597-1 have been published by the NRC in Regulatory Guide 1.147, Revision 13. A number of these issues are related to a need for additional explanation of the technical basis for the Code Case, such as the procedures for evaluation of wall thickness less than the ASME Code Design Pressure-based minimum allowable wall thickness. This presentation addresses these NRC conditions by providing additional description of the technical basis for the Code Case.

scholarly journals Improving Safety and Economic Feasibility of NPP by Management of Secondary Side Piping Life

2017 ◽  
pp. 25-29
Author(s):  
V. Kravchenko
Keyword(s):  
Economic Feasibility ◽  
Accelerated Corrosion ◽  
Life Management ◽  
Management Algorithm ◽  
Ultrasonic Thickness Gauging ◽  
The Usa ◽  
Computer Codes ◽  
Flow Accelerated Corrosion ◽  
Ultrasonic Thickness ◽  
Secondary Side

Flow accelerated corrosion (FAC), which caused major accidents at Sarri-2 NPP and Mihama-3 NPP, is the main ageing mechanism of NPP secondary side piping. FAC determines the service life of piping made of carbon steel. Wear prediction using computer codes is considered a tool for life management. The paper presents analysis of computer codes of the USA, Germany and Russia and considers the advantages of computer codes for life management and repair planning. Some problems that arise from using computer codes for analyzing the results of ultrasonic thickness gauging are indicated. The paper provides the algorithm for piping life management at Japanese NPPs and proposes the ways of improving piping life management algorithm, which makes it possible to proceed to the implementation of piping repair concept according to their condition.

Degradation of Alloy 800 Under Steam Generator Secondary Side Crevice Conditions

2009 ◽  
Author(s):  
Y. C. Lu ◽  
G. Goszczynski ◽  
S. Ramamurthy
Keyword(s):  
Steam Generator ◽  
Safety Issue ◽  
Cold Work ◽  
Orientation Imaging Microscopy ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Electrochemical Tests ◽  
Orientation Imaging ◽  
Alloy 800 ◽  
Secondary Side

Alloy 800 is the preferred steam generator (SG) tube materials for CANDU™ reactors and is also used extensively in SGs in some pressurized water reactor (PWR) systems. Degradation of Alloy 800 SG tubing has only been found in a few tubes at a limited number of stations despite the large number of SG tube operating years accumulated to date. Recently, underdeposit corrosion was detected in a few ex-service tubs removed from some CANDU SGs. Pits like wall loss of about 5% to 10% through-wall depth were found in these ex-service tubes. Evidence of intra-tubesheet cracking of Alloy 800 tubes was detected in a few European PWR SGs. There is no degradation in mechanical properties of these ex-service CANDU SG tubes. In addition, the degradation of Alloy 800 tubes observed so far is not a safety issue. However, the findings suggest that Alloy 800 tubing may have some aging degradation susceptibility after many years of service. Whether the degradation of Alloy 800 tubing is due to imperfections in its compositional or metallurgical properties inherent from manufacturing, or due to the aggressive chemistry conditions that should have been precluded by modern chemistry control strategy require clarification. Comprehensive examinations, including metallurgical examinations, orientation imaging microscopy (OIM), surface analyses and electrochemical measurements were performed on the removed ex-service CANDU SG tubes that had some underdeposit corrosion. The results were compared with a reference nuclear grade Alloy 800 tubing and with archive Alloy 800 new SG tubes from several CANDU stations. High-temperature electrochemical tests, scanning vibrating electrode Technique (SVET) measurements as well as C-ring autoclave tests were performed to determine the possible factors leading to Alloy 800 SG tubing degradation. SCC was initiated in a few C-ring specimens in the presence of artificial cold work flaws under simulated acidic SG secondary-side crevices chemistry conditions. OIM and surface analysis were also performed to characterize the degradation initiated in Alloy 800 tubing under the influence of cold work flaws. The possible factors leading to Alloy 800 SG tubing degradation under SG secondary crevices conditions are discussed.

Managing Flow Accelerated Corrosion in Carbon Steel Piping in Nuclear Plants

2004 ◽  
Author(s):  
Z. H. Walker
Keyword(s):  
Carbon Steel ◽  
Steam Generator ◽  
Wall Thickness ◽  
Degradation Mechanism ◽  
Reactor Core ◽  
Operating Conditions ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Flow Accelerated Corrosion ◽  
Steel Piping

In 1996, Flow Accelerated Corrosion (FAC) was identified as a degradation mechanism affecting carbon steel outlet feeder pipes in CANDU® (CANadian Deuterium Uranium) reactors. The maximum rate of FAC was estimated to be <0.120 mm/year. In response, wall thickness inspection programs have been implemented to identify and measure the minimum wall thickness in outlet feeder pipes. These data are necessary to ensure fitness-for-service of the feeder pipe. These data, together with the thermalhydraulic and geometric parameters for the measured feeders, are also very useful for developing empirical wall thickness models. Such models can be used to enhance the understanding of feeder wall thinning leading to an improved capability to predict future wall thickness minima and their locations. The determined dependency of the wall-thinning rate on thermalhydraulic conditions can be used to quantify the potential benefits of maintenance activities, such as steam generator cleaning. Activities such as steam generator cleaning are generally viewed as beneficial in recovering lost thermal efficiency, thereby reducing the severity of the thermalhydraulic conditions by reducing the amount of quality (steam phase) exiting the reactor core. Finally, when wall thickness models are applied to data from different plants, there is the potential of identifying operating conditions that can lead to lower rates of wall loss. This paper addresses the aforementioned important issues associated with FAC of ASME PVP Class 1 carbon steel piping.

Steam Generator Analysis Tools and Modeling of Degradation Mechanisms

2004 ◽  
Author(s):  
M. Yetisir ◽  
J. Pietralik ◽  
R. L. Tapping
Keyword(s):  
Nuclear Power ◽  
Fretting Wear ◽  
Design Stage ◽  
Degradation Mechanisms ◽  
Accelerated Corrosion ◽  
Maintenance Activity ◽  
Analysis Tools ◽  
Water Reactors ◽  
Flow Accelerated Corrosion ◽  
Secondary Side

The degradation of steam generators (SGs) has a significant effect on nuclear heat transport system effectiveness and the lifetime and overall efficiency of a nuclear power plant. Hence, quantification of the effects of degradation mechanisms is an integral part of a SG degradation management strategy. Numerical analysis tools such as THIRST, a 3-dimensional (3D) thermalhydraulics code for recirculating SGs; SLUDGE, a 3D sludge prediction code; CHECWORKS a flow-accelerated corrosion prediction code for nuclear piping, PIPO-FE, a SG tube vibration code; and VIBIC and H3DMAP, 3D non-linear finite-element codes to predict SG tube fretting wear can be used to assess the impacts of various maintenance activities on SG thermal performance. These tools are also found to be invaluable at the design stage to influence the design by determining margins or by helping the designers minimize or avoid known degradation mechanisms. In this paper, the aforementioned numerical tools and their application to degradation mechanisms in CANDU® recirculating SGs are described. In addition, the following degradation mechanisms are identified and their effect on SG thermal efficiency and lifetime are quantified: primary-side fouling, secondary-side fouling, fretting wear, and flow-accelerated corrosion (FAC). Primary-side tube inner diameter fouling has been a major contributor to SG thermal degradation. Using the results of thermalhydraulic analysis and field data, fouling margins are calculated. Individual effects of primary- and secondary-side fouling are separated through analyses, which allow station operators to decide what type of maintenance activity to perform and when to perform the maintenance activity. Prediction of the fretting-wear rate of tubes allows designers to decide on the number and locations of support plates and U-bend supports. The prediction of FAC rates for SG internals allows designers to select proper materials, and allows operators to adjust the SG maintenance strategy. CANDU nuclear power plants are pressurized heavy-water reactors that differ in design from pressurized water reactors (PWRs). As a result of this difference, degradation mechanisms in PWRs might be somewhat different; for example, unlike CANDU systems, PWRs do not experience significant primary-side fouling. However, the methodologies presented in this paper are applicable to both CANDU and PWR SGs.

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