Steam Generator Tube Integrity Research Articles

A Bayesian Probabilistic Approach to Improved Health Management of Steam Generator Tubes

Steam generator tube integrity is critical for the safety and operability of pressurized water reactors. Any degradation and rupture of tubes can have catastrophic consequences, e.g., release of radioactivity into the atmosphere. Given the risk significance of steam generator tube ruptures, it is necessary to periodically inspect the tubes using nondestructive evaluation methods to detect and characterize unknown existing defects. To make accurate estimates of defect size and density, it is essential that detection uncertainty and measurement errors associated with nondestructive evaluation methods are characterized properly and accounted for in the evaluation. In this paper we propose a Bayesian approach that updates prior knowledge of defect size and density with nondestructive evaluation data, accounting for detection uncertainty and measurement errors. An example application of the proposed approach is then demonstrated for estimating defect size and density in steam generator tubes using eddy current evaluation data. The proposed Bayesian probabilistic approach helps improve health management of steam generator tubes, thereby enhancing the overall safety and operability of pressurized water reactors.

Predicting growth rate of wall thinning for severely degraded SG tubes using a statistical methodology

Based on the requirements of TSTF-449 or NEI 97-06, operational assessment (OA) should be performed to guarantee the steam generator (SG) tube integrity. OA is a forward looking evaluation of the SG tube conditions. One of main evaluations for the OA is to estimate the growth rate of tube degradation prior to the next SG tube inspection. Therefore, the majority of this paper is to predict the growth rate of wall thinning for the SG tubes by way of a statistical methodology. The wall thinning of degraded SG tubes predicted by the present model agrees well with the plant measured one. The relative errors between the predictions and measurements are less than 10%. In addition, the present model would over-predict the wall thinning in most cases, revealing that this methodology could provide a useful and conservative tool for the PWR plant staff to execute the OA for SG tubes.

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.

Structural Integrity Estimates of Steam Generator Tubes Containing Wear-Type Defects

It is requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. This critical defect size has been determined based on a concept of plastic instability, however, which is known to be too conservative for some locations and types of defects. The application of this concept may even cause premature retirement of steam generator tubes. In reality, a reliable structural integrity estimation for steam generator tubes containing a defect has received increasing attention. Although several guidelines have been developed and used for assessing defect containing tubes, most of these guidelines are focused on stress corrosion cracking or wall-thinning phenomena. Because some of steam generator tubes fail due to fretting and so on, specific integrity estimation schemes for relevant defects are required. In this paper, more than a hundred three-dimensional finite element analyses of steam generator tubes under internal pressure condition are carried out to simulate the failure behavior of steam generator tubes with specific defect configurations: elliptical wear-type, tapered wedge-type, and flat wear-type defects. After investigating the effect of key parameters such as defect depth, defect length, and wrap or tapered angle on equivalent stress across the ligament thickness, burst pressure estimation equations are proposed in relation to material strengths. Predicted burst pressures agreeded well with the corresponding experimental data, so the proposed equations can be used to assess the structural integrity of steam generator tubes with wear-type defects.

OPTIMIZED DESIGN PARAMETERS OF EDDY CURRENT BOBBIN PROBE FOR STEAM GENERATOR TUBE IN NPPS

The bobbin probe is now widely accepted as a basic and important ECT technique among various ECT techniques for steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe is also one of the essential components which consist of the whole ECT examination system, and provides us with decisive data for the evaluation of tube integrity in accordance with acceptance criteria described in specific procedures. Accordingly, the design of ECT probe is especially important to examination results because the quality of acquired ECT data is determined by the optimized probe design characteristics, such as coil geometry, electrical properties, operation frequency, and so on. In this study, the relationship between electric characteristic changes and differential coil gap variation for the optimization of the ECT probe signal was investigated. As a result, for the examination of volumetric flaws fabricated on the O.D. side of ASME type calibration standard, we propose that the best Lissajous signal for the examination will be obtained when the coil gap of differential bobbin probe is approximately 1.2 ~ 1.6 mm .

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.

Statistical Assessment of Integrity in Steam Generator Tubes Considering Uncertainty of NDE

A statistical assessment model for structural integrity of steam generator tubes with axial cracks at the top of the tubesheet was proposed using Monte Carlo method. In the model, a method for estimating the number of "real cracks" from in-situ inspection (ISI) data was used. Based on the estimated "real cracks", the number of newly detected cracks and growth of cracks during arbitrary operating period were simulated using the Monte Carlo method. The flaw growth rate used in the simulation was statistically calculated from the periodic in-service non-destructive inspection data. The number of cracks, the probabilistic distribution of crack sizes at the end of next operating interval and the probability of burst during operation were calculated from numerously repeated simulations using the proposed model.

표면 마모결함을 고려한 증기발생기 세관의 구조건전성 평가

During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective steam generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement.

Burst Pressure Estimation of Steam Generator Tubes Based on Fracture Mechanics Analyses

During the past couple of decades, several limit load solutions have been proposed to resolve steam generator (SG) tube integrity issue. However, for estimation of specific load carrying capacity under different conditions, these solutions have to be modified by using lots of experimental data. The objective of this paper is to introduce a new burst pressure estimation scheme based on fracture mechanics analyses for SG tubes with a crack. In this context, closed-form equations were derived to get relevant parameters from three dimensional elastic-plastic finite element analyses and, then, a series of structural integrity analyses were carried out using the predicted J-integrals from the equations. Finally, in comparison with the experimental data as well as corresponding estimation results from conventional limit load solutions, it was proven that the proposed estimation scheme can be used as an efficient tool for integrity evaluation of cracked SG tubes.

증기발생기 세관의 파괴저항 특성 측정에 관한 연구

The structural and leakage integrity of steam generator tubes should be sustained against all postulated loads even if a crack is present. During the past three decades, most of the efforts with respect to integrity evaluation of steam generator tubes have been focused on limit load solutions but, recently, the applicability of elastic-plastic fracture mechanics was examined cautiously due to its effectiveness. The purpose of this paper is to introduce a testing method to estimate fracture resistance characteristics of steam generator tubes with a through-wall crack. Due to limited thickness and diameter, inevitably, the steam generator tubes themselves were tested instead of standard specimen or alternative ones. Also, a series of three dimensional elastic-plastic finite element analyses were carried out to derive closed-form estimation equations with respect to J-integral and crack extension for direct current potential drop method. Since the effectiveness of JSUBIC/SUB as well as J-R curves was proven through comparison with those of standard specimens taken from pipes, it is believed that the proposed scheme can be utilized as an efficient tool for integrity evaluation of cracked steam generator tubes.

관통균열 세관의 파열압력 예측을 위한 탄소성 파괴역학 해석

The structural and leakage integrity of steam generator tubes should be sustained all postulated loads with appropriate margin even if a crack is present. During the past three decades, for effective integrity evaluation, several limit load solutions have been used world-widely. However, to predict accurately load carrying capacities of specific components under different conditions, the solutions have to be modified by using lots of experimental data. The purpose of this paper is to propose a new burst pressure estimation scheme based on fracture mechanics analyses for steam generator tube with an axial or circumferential through-wall crack. A series of three dimensional elastic-plastic finite element analyses were carried out and, then, closed-form estimation equations with respect to both J-integral and crack opening displacement were derived through reference stress method. The developed engineering equations were utilized for structural integrity evaluation and the resulting data were compared to the corresponding ones from experiments as well as limit load solutions. Thereafter, since the effectiveness was proven by promising results, it is believed that the proposed estimation scheme can be used as an efficient tool for integrity evaluation of cracked steam generator tubes.

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.

Prediction of structural integrity of steam generator tubes under severe accident conditions

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating and design-basis accident conditions are reviewed. These rate-independent flow stress models are inadequate for predicting failure of steam generator tubes under severe accident conditions because the temperature of the tubes during such accidents can reach as high as 800°C where creep effects become important. Therefore, a creep rupture model for predicting failure was developed and validated by tests on unflawed and flawed specimens containing axial and circumferential flaws and loaded by constant as well as ramped temperature and pressure loadings. Finally, tests were conducted using pressure and temperature histories that are calculated to occur during postulated severe accidents. In all cases, the creep rupture model predicted the failure temperature and time more accurately than the flow stress models.

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.

Steam generator tube integrity program

The degradation of steam generator tubes in pressurized water nuclear reactors continues to be a serious problem, and the US Nuclear Regulatory Commission (NRC) is developing a performance-based rule and regulatory guide for steam generator tube integrity. To support the evaluation of industry-proposed implementation of these performance-based criteria, the NRC is sponsoring a new research program at Argonne National Laboratory on steam generator tubing degradation. The objective of the new program is to provide the necessary experimental data and predictive correlations and models that will permit the NRC to independently evaluate the integrity of steam generator tubes. The technical work in the program is divided into four tasks, (1) assessment of inspection reliability, (2) research on in-service inspection technology, (3) research on degradation modes and integrity, (4) development of methodology and technical assessments for current and emerging regulatory issues. The objectives of and planned research activities under each of these four tasks are described here.

Integrity of VVER steam generator tubes

The eddy-current inspection is applied to Vitkovice to non-destructive testing of the steam generator tubes. The steam generator tube specimens, 16 mm by 1.5 mm, are made of AISI 321 steel with stress corrosion cracks. They were tested by the single-frequency eddy current inspection method. Samples with known flaws were metallographically sectioned to determine the actual extent of corrosion degradation. The relation between measurement of the tube wall corrosion damage and actual extent of corrosion attack or stress corrosion cracking (SCC) was determined. Pressurized specimens with machined notches of various geometries were tested up to failure. The effect of flaw depth and length on the burst pressure were evaluated.

Recent Concerns with Reactor Coolant Chemistry Technology in Pressurized Water Reactors

Three concerns have recently arisen in reactor coolant chemistry technology: (a) the projected shortage of highly enriched, 99.99% 7Li for the control of reactor coolant pH, (b) the availability of anion resins for minimizing chloride elution problems, and (c) the additional inservice inspection requirements for the integrity of steam generator tubes, which subject plant personnel to radiation exposure. The normal consumption of 7Li can be reduced with modifications to plant operations, particularly those associated with purification demineralizers. The ion-exchange industry is having some difficulty in supplying the anion resins with low residual chlorides that are needed for the chloride elution problem, but they are actively working to provide them in the near future. Recognizing the importance of radiation levels associated with steam generator inservice inspection, Babcock & Wilcox is conducting a study program to determine the characteristics of corrosion products, to develop mathematical models that predict corrosion product behavior, and to develop methods for removing or minimizing the buildup of corrosion products in the reactor coolant system.