Screening method for piping wall loss by flow accelerated corrosion

Flow accelerated corrosion (FAC) phenomenon has persisted its impact on plant reliability and personnel safety. Unless we change the operation condition drastically, most parameters affecting FAC will not be effectively controlled. In order to help expand piping inspection coverage, we have developed a screening approach to monitor the wall thinning by direct current potential drop (DCPD) technique. To improve the applicability to the complex piping network such as the secondary cooling water system in PWR's, we devised the equipotential control method that can eliminate undesired leakage currents outside a measurement section. In this paper, we present Wide Range Monitoring (WiRM) and Narrow Range Monitoring (NaRM) with Equipotential Switching Direct Current Potential Drop (ES-DCPD) method to rapidly monitor the thinning of piping. Based on the WiRM results, susceptible locations can be identified for further inspection by ultrasound technique (UT). On-line monitoring of a thinned location can be made by NaRM. Finite element analysis results and a closed-form resistance model are developed for the comparison with measured wall thinning by the developed DCPD technique. Verification experiments were conducted using UT as the reference. The result shows that model predictions and the experimental results agree well to confirm that both WiRM and NaRM based on ES-DCPD can be applicable to FAC management efforts.     

On-line monitoring system development for single-phase flow accelerated corrosion

Aged nuclear piping has been reported to undergo corrosion-induced accelerated failures, often without giving signatures to current inspection campaigns. Therefore, we need diverse sensors which can cover a wide area in an on-line application. We suggest an integrated approach to monitor the flow accelerated corrosion (FAC) susceptible piping. Since FAC is a combined phenomenon, we need to monitor as many parameters as possible and that cover wide area, since we do not know where the FAC occurs. For this purpose, we introduce the wearing rate model which focuses on the electrochemical parameters. Using this model, we can predict the wearing rate and then compare testing results. Through analysis we identified feasibility and then developed electrochemical sensors for high temperature application; we also introduced a mechanical monitoring system which is still under development. To support the validation of the monitored results, we adopted high temperature ultrasonic transducer (UT), which shows good resolution in the testing environment. As such, all the monitored results can be compared in terms of thickness. Our validation tests demonstrated the feasibility of sensors. To support direct thickness measurement for a wide-area, the direct current potential drop (DCPD) method will be researched to integrate into the developed framework.     

Stress analysis of non-uniform thickness piping system with general piping analysis software

Most general piping analysis software can only perform ASME design stage type code compliance analysis with uniform pipe wall thickness. However, non-uniform wall thickness, commonly on elbows or bends, can be found in many industrial applications. A typical example is thinned non-uniform thickness at bends or elbows caused by flow accelerated corrosion (FAC). In this paper, an analysis procedure is introduced to enable a general piping software to conduct ASME III class 1 piping analysis with non-uniform wall thickness. The demonstration is performed on CANDU (Canadian Deuterium Uranium) feeder pipes, which have been subjected to FAC caused wall thinning. The results are compared with both conventional uniform thickness piping analysis and non-uniform thickness solid finite element analysis. The comparison shows the validity of the proposed “average-minimum-average” approach by employing the general piping analysis software. The approach remains conservative compared to the benchmark solid finite element analysis results. Meanwhile it provides lower acceptable thickness than the conventional piping analysis.     

Correction Techniques of Radiation Induced Errors for Raman Distributed Temperature Sensor and Experiment at the Experimental Fast Reactor: JOYO

Optical fibers have advantages like flexible configuration, intrinsic immunity for electromagnetic fields etc., and they have been used as optical fiber sensors. By some of these techniques, continuous or discrete distribution of physical parameters can be measured. Here, in order to apply Raman distributed temperature sensor (RDTS) to the monitoring of nuclear facilities: some correction techniques for radiation induced errors were investigated. It has been shown that, when uniform loss distribution can be assumed, simple correction technique with two thermocouples can be applied. Moreover, if loop type arrangement is applied, even when the loss distribution is not uniform, radiation induced errors can be canceled.

For the demonstration of the feasibility of this technique, measurements using a commercial RDTS sys-tem were carried out along the primary piping system of the experimental fast reactor: JOYO. During the continuous measurements with the total dose of more than 10⁷R, the radiation induced errors showed a saturating tendency. The correction technique with two thermocouples was applied and its feasibility has been demonstrated. Although the time response of the system should be improved, the RDTS can be expected as a noble temperature monitor in nuclear facilities.     

Evaluation of the proximity effect on flow-accelerated corrosion

Flow-accelerated corrosion (FAC) is a degradation mechanism that affects carbon steel piping in power plants. The failures and degradation due to FAC have necessitated numerous replacements in many power plants. Several computer codes around the world were developed as part of a systematic program or process to control FAC in power plant utilities. The typical plant model requires the input of the flow parameters, piping configuration and the plant water chemistry. The results on FAC rate are considered the key to proper selection of components for inspection. The lack of information on the effect of the upstream components located in the proximity limited the accuracy of the FAC prediction tools and hence will affect the accuracy in identifying potential inspection locations. In the present study 211 inspection data for 90° carbon steel elbows from several nuclear power plants were used to determine the effect of the proximity between two components on the FAC wear rate. The effect of the velocity as well as the distance between the elbows and the upstream components is discussed in the present analysis. Based on the analyzed trends obtained from the inspection data, significant increase in the wear rate of approximately 70% on average is identified to be due to the proximity.     

A two-phase methodology to predict FAC wear sites in the piping system of a BWR

Flow accelerated corrosion (FAC) wear is a serious degradation problem especially for nuclear power plants since it may result in the plant damage as well as risk to personnel. In this paper, a methodology which includes the two-phase hydrodynamic CFD models and FAC models, is proposed to predict severe FAC wear sites. Based on hydrodynamic simulation results, the present CFD models can precisely capture the two-phase characteristics within the piping system, which include the centrifugal effect, the gravitation effect and the imbalance of phase and mass separation in a T-junction, etc. Coupled with these flow characteristics, the appropriate FAC indicators can predict the possible locations of severe FAC wear. This methodology was validated against the measured results of wear site distributions for the piping system in a boiling water reactor (BWR) power plant. Good agreement between measurements and predictions at severe wear sites indicate that the present models can capture the characteristics of severe FAC wear and can help assist in the pipe wall-monitoring program for a nuclear power plant.     

Cyclic crack growth evaluation of a 20MnMoNi55 piping steel in high-oxygen reactor water

Samples of a low alloy steel piping material taken from the full scale corrosion fatigue test loop of the Heissdampfreaktor (HDR) plant have been tested at 240°C in high oxygen reactor water. The small-scale specimens (CT25) were exposed to a similar loading spectrum to that which has been used in the full-scale corrosion fatigue tests at the HDR-plant. During the autoclave tests cyclic crack growth rates were determined. Fracture surface investigations were performed not only for the laboratory test specimens but also for the fracture surface of a sample taken from the HDR test loop piping after the full scale test. In this paper the autoclave testing results and fracture surface observations are presented and compared to those obtained in the HDR piping tests.     

Development of Sensors for On-Line Monitoring of Nonmetallic Impurities in Liquid Sodium

Sensors for on-line monitoring of hydrogen and carbon in sodium and hydrogen in argon cover gas circuits over sodium have been developed. The performance of these sensors in fast breeder test reactor (FBTR) and large sodium facilities is evaluated. A sensor for monitoring oxygen in sodium is under development. The in-sodium electrochemical hydrogen sensors are found to detect about 10 ppb increase in hydrogen concentration over a background of 50 ppb. The cover gas hydrogen monitoring sensor system is found to sense hydrogen down to 2 vppm in argon over sodium systems. Electrochemical carbon sensors are capable of detecting down to 1 ppm of carbon in sodium.     

Fracture mechanics and full scale pipe break testing for the Department of Energy's new production reactor-heavy water reactor

Oak Ridge National Laboratory (ORNL) has completed a major task for the US Department of Energy (DOE) in the demonstration that the primary piping of the proposed new production reactor-heavy water reactor (NPR-HWR), with its relatively moderate temperature and pressure, should not suffer an instantaneous double-ended guillotine break (DEGB) under design basis loadings and conditions. The growth of possible small pre-existing defects in the piping wall was estimated over a plant life of 60 years. This worst-case flaw was then evaluated using fracture mechanics methods. It was calculated that this worst-case flaw would increase in size by at least 14 times before pipe instability during a safe shutdown earthquake (SSE) would even begin to be possible. The approach to showing the improbability of an instantaneous DEGB for HWR primary piping required a major facility (pipe impact test facility, PITF) to apply all possible design loads, including an equivalent major earthquake (called the SSE earthquake). The facility was designed and built at ORNL in 6 months. The test article was 6.1 m long, 406 mm diameter, 13 mm thick pipe of stainless steel 316LN material that was fabricated to exacting standards and inspections following the nuclear industry standard practices. A flaw was machined and fatigued into the pipe at a tungsten inert gas (TIG) butt weld (ER316L weld wire) as an initial condition. The flaw-crack was sized to be beyond the worst-case flaw that HWR piping could see in 60 years of service—if all leak detection systems and if all crack inspection systems failed to notice the flaw's existence. Starting October 1991, the first test article was subjected to considerable overloadings. The pipe was impacted 104 times at levels equal to and well beyond the SSE loadings. In addition, over 560 000 fatigue cycles and numerous purposeful static overloads were applied in order to extend the flaw to establish the data necessary to confirm fracture mechanics theories, and more importantly, to demonstrate simply that instantaneous DEGB is highly improbable for the relatively moderate energy system.     

Integration of nondestructive examination reliability and fracture mechanics

A multi-year program on the Integration of Nondestructive Examination and Fracture Mechanics (NDE/FM) has been funded by the U.S. Nuclear Regulatory Commission at the Pacific Northwest Laboratory. Many activities are being pursued under this program. This paper highlights some of the activities: input to the NRC Pipe Crack Task Group, an evaluation of manual ultrasonic testing of centrifugally cast stainless steel, interaction matrix, advanced UT technique evaluation, qualification document, evaluation of crack characterization techniques, international NDE reliability work, siamese imaging technique for imaging planar-type radial defects in reactor piping, fracture mechanics analysis for PTS-type flaws and piping reliability, and a position paper on piping ISI.     

Current and emerging pressure boundary issues — a US regulatory perspective

Concerns about pressure boundary integrity deal primarily with older plants, and establishing a basis for their continued safe operation. Pressure vessel problems stem from exposure to fast neutrons which changes the Nil-Ductility-Temperature (NDT) and the elevated temperature fracture energy of some vessels. The predicted shift in NDT has increased over the last decade as more has been learned about the effect of impurities (copper) and the synergism between nickel and copper. In PWRs this has lead to concern about excursion in which the a vessel remains at high pressure as the coolant temperature drops rapidly, that is the so-called Pressurized Thermal Shock (PTS) accident. In BWRs one cannot have PTS events, but the more rapid than expected rise in NDT due to irradiation is impacting operations.In another set of PWRs the upper shelf energy of the welds was initially low due to the use of a slag which led to many small inclusions in the weld. Radiation has lowered the Charpy fracture energy of these welds to below the 50 ft lb level at which there is concern that the vessel may undergo low energy ductile failure even if cleavage does not occur.Problems in pressure boundary piping has stemmed primarily from corrosion, that is, IGSCC in BWR recirculation piping, and steam generator tube failures in PWRs. These have made a large contribution to downtime and occupational exposure, but are not seen as significant contributors to risk. There has been some concern about the aging (loss of toughness) of cast stainless components with significant ferrite content, especially because inspection by UT is difficult.     

Fitness for service assessment of degraded CANDU feeder piping—Canadian regulatory expectations

Allowance for the continued operation of feeder piping at some Canadian CANDU stations, which is experiencing active degradation mechanisms, has been based primarily on augmented inspection practices and conservative fitness for service assessments. The major degradation mechanisms identified to date are: pipe wall thinning due to Flow Accelerated Corrosion (FAC) and service induced cracking due to Intergranular Cracking due to Stress Corrosion Cracking (SCC) and potentially Low Temperature Creep Cracking (LTCC) mechanisms. Given that currently available industry codes and standards do not provide sufficient guidelines/criteria for assessing the degradation of feeder pipes, the Canadian Nuclear Safety Commission (CNSC) has asked the utilities to establish feeder pipe specific procedures to provide reasonable assurance that the risk associated with the feeder degradation is maintained at an acceptably low level. In response to this requirement, the Canadian CANDU industry has developed and continued to update feeder fitness for service guidelines to provide evaluation procedures and industry standard acceptance criteria for assessing the structural integrity of the feeder pipes. The scope and frequency of inspections are determined based on the results of the fitness for service assessments taking into account the relative susceptibility of feeder pipes to each specific degradation mechanism. While industry practices for the management of degraded feeder pipes have, in general, been complied with the regulatory expectations, outstanding issues still remain. Major regulatory concerns include uncertainties associated with limitations in both the inspection techniques and the mechanistic understanding of the degradation processes, which can impede inspection planning and fitness for service assessments.This paper presents the regulator's view of the current situation with respect to degradation of feeder piping, its implications for nuclear safety and the regulatory expectations on industry's management of the critical ageing phenomena.     

Evaluation of flow accelerated corrosion by coupled analysis of corrosion and flow dynamics. Relationship of oxide film thickness, hematite/magnetite ratio, ECP and wall thinning rate

Systematic approaches to evaluate flow accelerated corrosion (FAC) are desired before discussing application of countermeasures for FAC. First, future FAC occurrence should be evaluated to identify locations where a higher possibility of FAC occurrence exists, and then, wall thinning rate at the identified FAC occurrence zone is evaluated to obtain the preparation time for applying countermeasures.Wall thinning rates were calculated with two coupled models:

1.

static electrochemical analysis and

2.

dynamic oxide layer growth analysis.

The anodic current density and the electrochemical corrosion potential (ECP) were calculated with the static electrochemistry model based on an Evans diagram. The ferrous ion release rate, determined by the anodic current density, was applied as input for the dynamic double oxide layer model. Some of the dissolved ferrous ion was removed to the bulk water and others precipitated on the surface as magnetite particles.The thickness of oxide layer was calculated with the dynamic oxide layer growth model and then its value was used as input in the electrochemistry model. It was confirmed that the calculated results (corrosion rate and ECP) based on the coupled models were in good agreement with the measured ones.Higher ECP was essential for preventing FAC rate. Moderated conditions due to lower mass transfer coefficients resulted in thicker oxide layer thickness and then higher ECP, while moderated corrosion conditions due to higher oxidant concentrations resulted in larger hematite/magnetite rate and then higher ECP.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

Determination of rupture frequency for feeders subject to flow accelerated corrosion

Feeder pipe wall thinning due to flow accelerated corrosion (FAC) has been identified as a degradation mechanism that may affect the operating life of outlet feeder pipes. A large number of feeders are expected to require costly repair or replacement over the remaining life of the station if a conservative FAC rate is used in the deterministic structural integrity assessment. The paper presents a preliminary probabilistic framework for determining the rupture frequency of feeders subject to FAC based on commercial probabilistic fracture mechanics code WinPRAISE 2007. The obtained information can be used as inputs for developing risk-informed feeder life management plans. Darlington Unit 2 is selected to demonstrate the proposed method.     

孔板管道下游流动加速腐蚀速率数值模拟研究

采用计算流体力学方法中的k-ε模型模拟了孔板管道下游管壁与流体间的传质系数分布,并利用Sanchez-Caldera流动加速速率预测模型计算了孔板管道下游的流动加速腐蚀速率分布。结果表明,孔径比的减小会导致流动加速腐蚀敏感部位向孔板下游移动,入口流速的增大对孔板下游流动加速腐蚀敏感部位的位置无明显影响,pH值的增大能有效减小流动加速腐蚀速率。     

Predicting the wall thinning engendered by erosion-corrosion using CFD methodology

Erosion-corrosion (EC) is a serious degradation mechanism of piping, especially for nuclear power plants since it may result in the piping damage, plant shutdown, or personnel injury. The majority of this paper investigates the dependence of wall thinning on the hydrodynamic characteristics using the computational fluid dynamics (CFD) methodology. Four piping systems in a pressurized water reactor (PWR) power plant are selected in this investigation. Based on the plots showing the measured wall thinning with the calculated hydrodynamic parameters, the relationship between them is clearly revealed. Utilizing the characteristics of near-wall turbulence kinetic energy, an envelopment model is proposed herein to conservatively predict the amount of wall thinning distributed on the pipe wall. This estimation model can simply predict the possible distributions of severe EC wear sites and subsequently assist the plant staff to schedule the pipe wall monitoring program in the measured range of pipe wall for the fittings.     

Experimental investigation on heat transfer characteristics in a safety injection nozzle component in PWR

Thermal fatigue is a potentially significant degradation mechanism in Nuclear Power Plants (NPP). For the fatigue analysis, the thermal load information about components must be determined firstly. In this paper, an experimental study was carried out to obtain local fluid temperatures and local heat transfer coefficients for the safety injection nozzle component in reactor coolant system (RCS). In this mixing tee component a hot jet issues into a cold cross-flow stream from an oblique pipe and the turbulent mixing of two fluids induces local cycling stresses on the adjacent piping wall. Experiments were performed using a special-made heat fluxmeter, which can measure the mixed fluid temperature close to the wall and the heat transfer coefficient between the fluid and the wall. Plexiglass and metallic 1/9-scale mockups were manufactured for flow visualization and heat transfer tests, respectively. All tests were conducted at range of 0–40 for the jet-to-cross-flow velocity ratio. The flow visualization test has obtained general pattern of the flow and identified sensitive zones in the component where the jet and cross-flow interact intensively to cause thermal fatigue more possibly. In the heat transfer test, heat fluxmeters were positioned in the wall at these sensitive zones. The measurement results of temperatures and heat transfer coefficients have been discussed in detail in the paper. These experimental results allow us improving the state of knowledge of the thermal load to be used in the industrial mixing tees in operating for long lifetime assessment and for the design in the basic Nuclear Power Plants.     

Ontario hydro's leak before break approach: Application to the darlington (candu) nuclear generating station a

Ontario Hydro has developed a leak before break (LBB) approach for application to the large diameter heat transport piping for Darlington NGS A as an alternative to the provision of pipewhip restraints. This approach has been applied to pipe sizes which are equal to or greater than 530 mm (21 in. NPS). The proposed LBB approach incorporates assessments at several levels to provide assurance against catastrophic rupture. A comprehensive and systematic review of pipe failure mechanisms is considered the first important step in establishing role and applicability of the LBB concept. The elements integral to the approach are those related to demonstration of crack stability utilizing fracture mechanics methods and those related to leak rate predictions and leakage detection capability. For evaluation of crack stability the J-integral/tearing modulus (J/T) method has been selected. Results from an extensive material test program from actual heat transport piping, forgings, associated welds and heat affected zones as inpur to EPFM analyses provide the J-resistance and J—T curves. The details of EPFM analyses for a straight pipe with a circumferential crack and a piping elbow with a central longitudinal throughwall crack are presented here. Additionally, results of crack opening detail, the effects of crack face pressure, the predictions of LEAK RATE code and an assessment of the leakage detection capability are presented.     

A probabilistic model of wall thinning in CANDU feeders due to flow-accelerated corrosion

Wall thinning due to flow-accelerated corrosion (FAC) is a pervasive form of degradation in feeder pipes of the primary heat transport system of CANDU reactors. Prediction of the end-of-life of a feeder from wall thickness measurement data is confounded by the sampling and temporal uncertainties associated with the FAC degradation phenomenon. This paper presents a probabilistic model of wall thinning due to FAC, and calibrates it with a set of feeder wall thickness measurements obtained from a CANDU plant. The proposed model derives the feeder lifetime distribution, which is useful in developing optimum strategies for life-cycle management of the feeder system.