Diagnosis of Corrosion Processes in Nuclear Power Plants Secondary Piping Structures

Author(s):  
Koushik A. Manjunatha ◽  
Andrea Mack ◽  
Vivek Agarwal ◽  
David Koester ◽  
Douglas Adams

Abstract The current aging management plans of passive structures in nuclear power plants (NPPs) are based on preventative maintenance strategies. These strategies involve periodic, manual inspection of passive structures using nondestructive examination (NDE) techniques. This manual approach is prone to errors and contributes to high operation and maintenance costs, making it cost prohibitive. To address these concerns, a transition from the current preventive maintenance strategy to a condition-based maintenance strategy is needed. The research presented in this paper develops a condition-based maintenance capability to detect corrosion in secondary piping structures in NPPs. To achieve this, a data-driven methodology is developed and validated for detecting surrogate corrosion processes in piping structures. A scaled-down experimental test bed is developed to evaluate the corrosion process in secondary piping in NPPs. The experimental test bed is instrumented with tri-axial accelerometers. The data collected under different operating conditions is processed using the Hilbert-Huang Transformation. Distributional features of phase information among the accelerometers were used as features in support vector machine (SVM) and least absolute shrinkage and selection operator (LASSO) logistic regression methodologies to detect changes in the pipe condition from its baseline state. SVM classification accuracy averaged 99% for all models. LASSO classification accuracy averaged 99% for all models using the accelerometer data from the X-direction.

Author(s):  
Lifei Yang ◽  
Jiang Hong

It is well-known that RCM is an advanced and effective maintenance strategy in practice. With the development of the automation and mechanization in modern industry, RCM method turns to be complex and consumes more resources in real production. However, the development and application of the Streamline RCM (SRCM) has injected new vitality for the new situation, especially in the nuclear power plants. This paper firstly introduces the background, the characteristics of the SRCM and the differences from RCM, and then shows the process in detail as well as the application status of the SRCM in country and abroad. It is proved that SRCM is a unique available method which saves the time and resource consumed, ensuring the integrity and correctness of the classical RCM. Finally, the weak points and the prospect are reviewed and prospected.


2003 ◽  
Author(s):  
J. Guillou ◽  
L. Paulhiac

Several vibration-induced failures at the root of small bore piping systems occurred in French nuclear power plants in past years. The evaluation of the failure risk of the small bore pipes requires a fair estimation of the bending stress under operating conditions. As the use of strain gauges is too time-consuming in the environmental conditions of nuclear power plants, on-site acceleration measurements combined with numerical models are easier to handle. It still requires yet a large amount of updating work to estimate the stress in multi-span pipes with elbows and supports. The aim of the present study is to propose an alternate approach using two accelerometers to measure the local nozzle deflection, and an analytical expression of the bending stiffness of the nozzle on the main pipe. A first formulation is based on a static deformation assumption, thus allowing the use of a simple analog converter to get an estimation of the RMS value of the bending stress. To get more accurate results, a second method is based on an Euler Bernoulli deformation assumption: a spectral analyzer is then required to get an estimation of the spectrum of the bending stress. A better estimation of its RMS value is then obtained. An experimental validation of the methods based on strain gauges has been successfully performed.


2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.


2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.


Author(s):  
A. A. Mikhalevic ◽  
U. A. Rak

The article presents the analysis of the specific features of modeling the operation of energy systems with a large share of nuclear power plants (NPP). The study of operating conditions and characteristics of different power units showed that a power engineering system with a large share of NPP and CHPP requires more detailed modeling of operating modes of generating equipment. Besides, with an increase in the share of installations using renewable energy sources, these requirements are becoming tougher. A review of the literature revealed that most often the curve of the load duration and its distribution between blocks are used for modeling energy systems. However, since this method does not reflect a chronological sequence, it can only be used if there are no difficulties with ensuring power balance. Along with this, when the share of CHP and nuclear power plants is high, to maintain a balance of power one must know the parameters and a set of powered equipment not only currently but, also, in the previous period. But this is impossible if a curve of load duration is used. For modeling, it is necessary to use an hourly load curve and to calculate the state of the energy system for each subsequent hour in chronological order. In the course of a comparative analysis of available computer programs, it was not possible to identify a suitable model among the existing ones. The article presents a mathematical model developed by the authors, which makes us possible to simulate the operation of a power engineering system with a large share of NPP and CHPP while maintaining the power balance for each hour of the forecast period. Verification of the proposed model showed good accuracy of the methods used.


Author(s):  
Pei-Yin Chen ◽  
Patrick Sekerak ◽  
Thomas Scarbrough ◽  
Cheng-Ih Wu

In recent years, the nuclear industry experienced adverse flow effects that caused structural damage to safety-related and nonsafety-related components as a result of flow-induced acoustic resonance in both Boiling Water Reactor (BWR) and Pressurized Water Reactor plants. In particular, fatigue failures and cracks in steam dryers occurred in certain BWR plants during the extended power uprate operation with generation of loose parts that can adversely affect safety-related components within the reactor vessel and the reactor coolant system. The acoustic resonance occurs when the main steam line flow exceeds a critical value such that the vortices over the cavity of the closed side branch pipe are excited by the acoustic modes of the stagnant fluid in the branch. The occurrence of this phenomenon is highly dependent on plant-specific operating conditions and the piping as-built configuration. The U.S. nuclear industry has initiated extensive activities to address this phenomenon in BWR plants. The staff of the U.S. Nuclear Regulatory Commission (NRC) has been monitoring generic industry activities, as well as reviewing the evaluation of potential adverse flow effects that might result from power uprates at current operating plants, and during the design certification and licensing of new reactors. This paper discusses operating experience with adverse flow effects at nuclear power plants from the acoustic resonance phenomenon, industry actions to address and resolve the phenomenon, and NRC staff review activities related to this issue.


Author(s):  
Arnold Gad-Briggs ◽  
Pericles Pilidis ◽  
Theoklis Nikolaidis

Studies are currently on-going on the cycle performance of Generation IV (Gen IV) Nuclear Power Plants (NPPs) for the purpose of determining optimum operating conditions for efficiency and economic reasons. For Gas-cooled Fast Reactors (GFRs) and Very-High Temperature Reactors (VHTRs), the cycle layout is predominantly driven by the choice of components, the component configuration and the coolant. The purpose of this paper to present and review the cycles currently being considered — the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). In all cases, the cycles utilise helium as the coolant in a closed Brayton gas turbine configuration. Comparisons between the cycles are made for Design Point (DP) and Off-Design Point (ODP) analyses to emphasise the benefits and drawbacks of each cycle. The paper also talks about future trends which include higher Core Outlet Temperatures in excess of 1000 degrees Celsius and the proposal of a simplified cycle configuration which eliminates the need for the recuperator.


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