Cracked Steam Generator Tubes Research Articles

A comparison of failure assessment diagram options for Inconel 690 and Incoloy 800 nuclear steam generators tubes

In the context of structural integrity assessment of cracked Steam Generator Tubes (SGTs), the Failure Assessment Diagram (FAD) has been proposed as a modern tool able to reduce the conservativeness of traditional plastic limit load analyses. As different FAD failure curves are available in the literature, this paper presents a comparison between the so-called Options 1, 2 and 3 FAD curves derived for Inconel 690 and Incoloy 800 nuclear SGTs. Detailed Option 3 curves are obtained from elastoplastic finite element analyses and experimental data from fracture tests. Material-specific Option 2 curves are drawn from stress vs. strain data of uniaxial tests of SGTs. It is shown that the simplest generic Option 1 curve is conservative but appropriate for structural integrity assessments of cracked nuclear SGTs in operation. Alternatively, Option 3 FAD curves must be used when an excessive conservatism shall be avoided.

An experimental investigation of subcooled choked flow in actual steam generator tube cracks

The work presented here is an experimental investigation of the two-phase choked flow of initially subcooled water through a stress corrosion crack on the steam generator tube. The study was designed to provide an extensive experimental data set for predicting leak flow rates of nuclear reactor coolant using actual tube steam generator tubes sample. Five samples of steam generator tubes with stress corrosion axial cracks taken from steam generators of the US pressurized water reactor and Canadian pressurized heavy water reactor (CANDU) were used in this study. Wall thicknesses of the samples vary from 1.1 mm to 1.3 mm with the crack channel length to hydraulic dimeter ratios from 3 to 15. The experiment was carried out for various stagnation pressures up to 6.89 MPa and liquid subcoolings ranging from 5 °C to 70 °C. Parametric dependence of choked flow rate on stagnation pressure, crack channel length to hydraulic diameter ratio, and degree of liquid subcooling was demonstrated for different crack geometries. All the parameters are shown to have significant effects on the choking mass flux. The experimental results indicate that the channel length strongly affects the choking mass flux as well as the correlation between this parameter and the stagnation pressure. These data are unique as they were obtained with actual steam generator tube cracks and hence are advantageous in understanding the choked flow in the very short channel length channel and similar flow geometries.

Numerical validation of burst pressure estimation equations for steam generator tubes with multiple axial surface cracks

This paper provides further validation of the burst pressure estimation equations for multiple axial surface cracked steam generator tubes, recently proposed by the authors based on analytical local collapse load approach against systematic FE damage analysis results of Alloy 690 tubes with twin axial surface cracks. Wide ranges of the relative crack depth and multiple crack configurations are considered. Comparison shows good agreements, giving sufficient confidence of the proposed equations.

Theoretical and Experimental Investigation of Subcooled Flashing Flow through Simulated Steam Generator Tube Cracks

ABSTRACTFlow models and experiments were developed to predict the choking flow of subcooled flashing water through steam generator tube crack. Experiments were conducted for vessel pressures up to 7 MPa with various subcoolings through 7 sets of simulated crack geometries with channel length to diameter ratios from 1.3 to 2, and channel length of 1.3 mm. Mass flux data with respect to subcooling are presented for all seven sets of crack specimens. As the subcooling increased for each specimen the mass flux increased. In general as the mass flux increased the area of the specimens increased. Experimental data were compared with homogeneous equilibrium and non-equilibrium mechanistic models for two-phase choking flow. A comparison of the model results with experimental data shows that the homogeneous equilibrium-based models grossly underpredict choking flow rates in such geometries, while homogeneous non-equilibrium models greatly increase the accuracy of the predictions.

Numerical Prediction of Maximum Load-Carrying Capacity of Cracked Alloy 690TT Steam Generator Tubes

This paper presents a finite element (FE) simulation technique to predict maximum load-carrying capacity of cracked steam generator tubes and its application to Alloy 690TT tubes. The simulation method is based on a simplified version of the stress modified fracture strain model. The damage model is determined from tensile test and one cracked tube test data. Predicted maximum pressures are compared with 23 test data of axial through-wall and surface cracked Alloy 690TT steam generator tubes. Comparison with experimental data shows good agreement.

INFLUENCE OF SIGNAL-TO-NOISE RATIO ON EDDY CURRENT SIGNALS OF CRACKS IN STEAM GENERATOR TUBES

This work presents the influence of noise originating from the tube itself on the detectability and sizing accuracy for laboratory-induced outer diameter axial cracks in nuclear steam generator tubes. The variations of signal amplitude and phase angle of the same cracks were analyzed when increasing the signal-to-noise ratio of the tube itself from 9 to 18. It was experimentally verified that the detectability for small cracks was enhanced by increasing the signal-to-noise ratio. The phase angle also rotated to a value representing the actual position and depth of a crack when increasing the signal-to-noise ratio.

Development of probability of detection functions for evaluating integrity of axial outside diameter stress corrosion cracks in steam generator

The reliability of an eddy current testing (ECT) inspection system depends upon the inspection technique and quality of analyst. In evaluating the integrity of a steam generator (SG) tube, degradation detection and sizing accuracy are considered performance measures of the nondestructive evaluation (NDE) system. A probability of detection (POD) model serves as a functional measure of the ability of an NDE system to detect degradation. It is one of the inputs in the operational assessment, and it is used to estimate the degradation during service via ECT of the SG tube. In this study, the POD functions of the inspection technique and analyst were obtained to quantitatively analyze the ECT bobbin probe for axial outside diameter stress corrosion cracks in SG tubes. This should serve to evaluate the integrity of the SG tubes. The depth and amplitude of defects were used as parameters of the POD model. Hit (detection) and miss (no detection) binary data obtained from destructive and nondestructive inspection of cracked tubes were also used.

Experimental and numerical study on the burst pressure of steam generator tubes with L- and U-type through-wall combination cracks

Several nuclear power plants have reported that they often experience through-wall combination cracks consisting of longitudinal and circumferential cracks in the steam generator tubes. In this study, L- and U-type through-wall combination cracks in a steam generator tube were evaluated using the crack opening angle (COA) in an elastic-plastic analysis. The calculated burst pressures were also compared with the experimental results. In the procedure, the stress intensity factor (KI) in elastic analysis was reviewed to validate COA as an appropriate fracture parameter for the burst pressure of longitudinal through-wall cracks and T-type through-wall combination cracks. From the results, an appropriate length-based burst criterion by using COA can be used for a structural integrity assessment for the through-wall combination cracks in steam generator tubes.

Benchmarking an expert fault detection and diagnostic system on the Three Mile Island accident event sequence

Early fault identification systems enable detecting and diagnosing early onset faults or fault causes which allow maintenance planning on the equipment showing signs of deterioration or failure. This includes valve and leaks and small cracks in steam generator tubes usually detected by means of ultrasonic inspection.We have shown (Cilliers and Mulder, 2012) that detecting faults early during transient operation in NPPs is possible when coupled with a reliable reference to compare plant measurements with during transients. We have also shown (Cilliers, 2013) that by correlating the fault detection information as received from distributed systems it is possible to diagnose the faults in terms of location and magnitude.This paper makes use of the techniques and processes developed in the previous papers and apply it to a case study of the Three Mile Island accident. In this way we can determine how the improved information available could present the operator with a better idea to the state of the plant during situations where a combination of faults and transients prevents the operator and conventional systems to recognise the abnormal behaviour.

Flashing Flow of Subcooled Liquid through Small Cracks

Abstract In Pressurized Water Reactors and in CANDU reactors steam generator (SG) tubes represent a major fraction of the reactor primary coolant pressure boundary. The ability to estimate the leak rates from the through wall cracks in the steam generator tube is important in terms of radiological source terms and overall operational management of steam generators as well as demonstration of the leak-before- break condition. In this study an experimental program and analysis methods were developed to measure and assess the choking flow rate of subcooled water through simulated steam generator tube crack geometries. Experiments were conducted on choking flow for various simulated crack geometries for vessel pressures up to 7 MPa with various subcoolings. Measurements were done on subcooled flashing flow rate through well defined simulated crack geometries with L/D

Correlating hardware fault detection information from distributed control systems to isolate and diagnose a fault in pressurised water reactors

Early fault identification systems enable detecting and diagnosing early onset faults or fault causes which allow maintenance planning on the equipment showing signs of deterioration or failure. This includes valve and leaks and small cracks in steam generator tubes usually detected by means of ultrasonic inspection.We have shown (Cilliers and Mulder, 2012) that detecting faults early during transient operation in NPPs is possible when coupled with a reliable reference to compare plant measurements with during transients. The problem introduced by the distributed application of control systems operating independently to keep the plant operating within the safe operating boundaries was solved by re-introducing the fault information it into the measurement data, thereby improving plant diagnostic performance.This paper introduces the use of improved fault detection information received from all distributed systems in the plant control system and correlating the information to not only detect the fault but also to diagnose it based on the location and magnitude of the fault cause.

Adapting plant measurement data to improve hardware fault detection performance in pressurised water reactors

With the fairly recent adoption of digital control and instrumentation systems in the nuclear industry a lot of research now focus on the development expert fault identification systems. The fault identification systems enable detecting early onset faults of fault causes which allows maintenance planning on the equipment showing signs of deterioration or failure. This includes valve and leaks and small cracks in steam generator tubes usually detected by means of ultrasonic inspection.Detecting faults early during transient operation in NPPs is problematic due to the absence of a reliable reference to compare plant measurements with during transients. The distributed application of control systems operating independently to keep the plant operating within the safe operating boundaries complicates the problem since the control systems would not only operate to reduce the effect of transient disturbances but fault disturbances as well.This paper provides a method to adapt the plant measurements that isolates the control actions on the fault and re-introduces it into the measurement data, thereby improving plant diagnostic performance.

ICONE19-43676 Assessment of Choking Flow Models for Subcooled Flashing Flow through Steam Generator Tube Cracks

The state-of-the-art modeling of leak rates through individual pipe stress corrosion cracks is based on regression analysis of experimentally determined leak rates and defects size. This has relatively large uncertainties. Accurate leak rate predictions require two key parameters, crack area and choking flow rate. The thermalhydraulics models employed in the leak rate calculations are generally based on the Henry-Fauske model of two-phase flow through long channels. For small length to diameter ratio (L/D) channels such as in steam generator tube cracks both mechanical and thermal nonequilibrium must be accounted. Another very important problem in verifying the model of the critical flow through cracks is the limited data on the crack geometry. Recently there is some database available on choking flow through cracks relevant to steam generator (SG) tubes to model the critical flow. These data are used in assessing the key choking flow models. Based on this assessment a mechanistic choking model is developed. The model is used to predict the choking flow rates for various experimental conditions for subcooled flashing flow through narrow slits with L/D varying from small values (〜5) to large values (100). Results are presented on the effects of thermal and mechanical non-equilibrium on the choking flow for small L/D channels.

Evaluation of ECT reliability for axial ODSCC in steam generator tubes

The integrity of steam generator tubes is usually evaluated based on eddy current test (ECT) results. Because detection capacity of the ECT is not perfect, all of the physical flaws, which actually exist in steam generator tubes, cannot be detected by ECT inspection. Therefore it is very important to analyze ECT reliability in the integrity assessment of steam generators. The reliability of an ECT inspection system is divided into reliability of inspection technique and reliability of quality of analyst. And the reliability of ECT results is also divided into reliability of size and reliability of detection. The reliability of ECT sizing is often characterized as a linear regression model relating true flaw size data to measured flaw size data. The reliability of detection is characterized in terms of probability of detection (POD), which is expressed as a function of flaw size. In this paper the reliability of an ECT inspection system is analyzed quantitatively. POD of the ECT inspection system for axial outside diameter stress corrosion cracks (ODSCC) in steam generator tubes is evaluated. Using a log-logistic regression model, POD is evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive inspections of cracked tubes. Crack length and crack depth are considered as variables in multivariate log-logistic regression and their effects on detection capacity are assessed using two-dimensional POD (2-D POD) surface. The reliability of detection is also analyzed using POD for inspection technique (POD T) and POD for analyst (POD A).

Assessment of Steam Generator Tube Flaw Size and Leak Rate Models

The Canadian Nuclear Safety Commission recently developed the CANTIA (CANDUTM Tube Inspection Assessment) methodology for probabilistic assessment of inspection strategies for steam generator (SG) tubes as a direct effect on the early detection and prevention of tube failure and primary-to-secondary leak of reactor coolant. In an effort to improve CANTIA, an SG tube integrity assessment code, a relevant survey of the literature on the discharge of subcooled water from cracks and critical flow models, SG tube cracks, leakage, and probabilistic assessment methodologies was carried out. The original CANTIA and ANL/CANTIA code models for the flaw opening area and flow leakage rate were reviewed. The predictions from the crack opening area and the leakage flow rate models were compared with experimental measured data from cracked SG tubes.

Influence of flow stress choice on the plastic collapse estimation of axially cracked steam generator tubes

The influence of the choice of flow stress on the plastic collapse estimation of axially cracked steam generator (SG) tubes is considered. The plastic limit and collapse loads of thick-walled tubes with external axial semi-elliptical surface cracks are investigated by three-dimensional non-linear finite element (FE) analyses. The limit pressure solution as a function of the crack depth, length and tube geometry has been developed on the basis of extensive FE limit load analyses employing the elastic–perfectly plastic material behaviour and small strain theory. Unlike the existing solutions, the newly developed analytical approximation of the plastic limit pressure for thick-walled tubes is applicable to a wide range of crack dimensions. Further, the plastic collapse analysis with a real strain-hardening material model and a large deformation theory is performed and an analytical approximation for the estimation of the flow stress is proposed. Numerical results show that the flow stress, defined by some failure assessment diagram (FAD) methods, depends not only on the tube material, but also on the crack geometry. It is shown that the plastic collapse pressure results, in the case of deeper cracks obtained by using the flow stress as the average of the yield stress and the ultimate tensile strength, can become unsafe.

Limit Loads for Circumferential Cracks in the Steam Generator Tube under Combined Pressure and Bending Loads

Steam generator in a nuclear power plant is huge heat exchanger that transfers heat from reactor to make steam to drive turbine-generator. Failure of the steam generator tubes can result in the release of fission products to the secondary side. Therefore, accurate integrity assessment of the cracked steam generator tubes is of great importance for maintaining the safety of the nuclear power plant. This paper provides limit loads for circumferential through-wall cracks in steam generator tubes under combined internal pressure and bending loads. Such limit loads are developed on the basis of three dimensional finite element analyses assuming elastic-perfectly plastic material behavior. As for the crack location, both the top of the tubesheet and U-bend regions are considered. The analysis results can be directly applied to the practical integrity assessment of cracked steam generator tubes, because the comparison between experimental data and FE results shows a very good agreement.

Determination of failure pressure for tubes with two non-aligned axial through-wall cracks

The 40% of wall thickness criterion which has been used as a plugging rule is applicable only to a single cracked steam generator tubes. In the previous studies performed by authors, several failure prediction models were introduced to estimate the plastic collapse pressures of steam generator tubes containing two adjacent collinear or parallel axial through-wall cracks. The objective of this study is to examine the failure prediction models and propose optimum ones for two non-aligned axial through-wall cracks in steam generator tubes. In order to determine the optimum ones, a series of plastic collapse tests and finite element analyses were carried out for steam generator tubes with two machined non-aligned axial through-wall cracks. Thereby, either the plastic zone contact model or COD based model was selected as the optimum one according to axial distance between two cracks. Finally, the optimum failure prediction model was used to demonstrate the conservatism of flaw characterization rules for multiple cracks having equal lengths according to ASME code.

Elastic‐plastic fracture mechanics assessment for steam generator tubes with through‐wall cracks

ABSTRACTSThe present work provides an elastic‐plastic fracture mechanics (EPFM) assessment scheme for a steam generator tube with a through‐wall crack under internal pressure. Noting that the geometry and material are rather uniform for steam generator tubes, and furthermore the only loading to be considered is internal pressure, an engineering EPFM analysis method is proposed to assess through‐wall cracks in steam generator tubes. Important outcomes of the present work are closed‐form approximations for J and crack opening displacement (COD). Sufficient confidence in the proposed J and COD estimates is gained from good agreements with the finite element results over a wide range of the crack length and pressure magnitude. Another important element of the EPFM assessment is to determine relevant J‐resistance curve for steam generator tubes. To improve the accuracy of predicting tube failure, the present paper also proposes a new method to determine fracture toughness using an actual tubular specimen instead of using a standard specimen, from which J‐resistance curves of steam generator tubes are obtained. Using the proposed J and toughness estimates, maximum pressures of steam generator tubes with through‐wall crack are estimated based on EPFM analysis, which is compared with experimental results and predicted ones based on limit load approach.

Coalescence evaluation of collinear axial through-wall cracks in steam generator tubes

To maintain the structural integrity of steam generator tubes, usually, 40% of wall thickness plugging criterion has been adopted. However, since the criterion is applicable only for the steam generator tube containing a single crack, the interaction effect of multiple cracks cannot be considered. In this paper, the coalescence pressure of tube with dual cracks is evaluated based on detailed three-dimensional elastic–plastic finite element analyses. In terms of the crack configuration, collinear axial through-wall cracks with various length, distance and ratio between individual cracks are selected. The applicability of failure pressure prediction models recently proposed by the authors was verified by comparing the finite element analyses results with corresponding experimental data for tubes with two identical cracks. Further, in order to quantify the effect of crack length ratio on failure behavior, the failure pressure prediction model was used expansively for tubes containing different-sized cracks and a coalescence evaluation diagram was developed.