Integrity evaluation of Alloy 600 RV head penetration tubes in Korean PWR plants

International Nuclear Information System (INIS)

Kang, Young Hwan; Park, Sung Ho; Hong, Sung Yull; Choi, Kwang Hee

1995-01-01

The structural integrity assessment of Alloy 600 RV head penetration tubes has been an important issue for the economical and reliable operation of power plants. In this paper, an overview of the integrity evaluation program for the RV head penetration tubes in Korean nuclear power plants is presented. Since the crack growth mechanism of the penetration tube is due to the primary water stress corrosion cracking (PWSCC) which is mainly related to the stress at the tube, the present paper consists of three primary activities: the stress evaluation, the flaw evaluation, and data generation through material and mechanical tests. (author). 5 refs, 2 figs, 1 tab

Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

Energy Technology Data Exchange (ETDEWEB)

Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

2011-10-15

Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

Estimation of the number of physical flaws from periodic ISI data of SG tubes using effective POD

International Nuclear Information System (INIS)

Lee, Jae Bong; Park, Jai Hak; Kim, Hong Deok; Chung, Han Sub

2008-01-01

It is necessary to know the number of flaws and their size distribution in order to calculate the probability of failure or to estimate the amount of leakage through the tube wall of steam generators. But In-Service Inspection (ISI) flaw data is different from the physical flaw data. In case of a single inspection, it is easy to estimate the number of physical flaws using the POD curve. However, we may be faced with some difficulties in obtaining the number of physical flaws from the periodic in-service inspection data. In this study a method for estimating the number of physical flaws from periodic in-service inspection data is proposed. In order to calculate the number of physical flaws with periodic ISI data, both probabilities of detecting and missing flaws should be considered. And flaw initiation and growth history must be known also. The flaw initiation and growth history can be inferred from appropriate probabilistic flaw growth rate. Two inference methods are proposed and compared. One is Monte Carlo simulation method and the other is transition (stochastic) matrix method. The effective POD, the total possibility of detection considering both probabilities of detecting and missing flaws for each flaw size, can be calculated using above two inference methods. And two methods are compared and the usefulness and convenience are evaluated from several applications

Bounding the conservatism in flaw-related variables for pressure vessel integrity analyses

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.

1993-01-01

The fracture mechanics-based integrity analysis of a pressure vessel, whether performed deterministically or probabilistically, requires use of one or more flaw-related input variables, such as flaw size, number of flaws, flaw location, and flaw type. The specific values of these variables are generally selected with the intent to ensure conservative predictions of vessel integrity. These selected values, however, are largely independent of vessel-specific inspection results, or are, at best, deduced by ''conservative'' interpretation of vessel-specific inspection results without adequate consideration of the pertinent inspection system performance (reliability). In either case, the conservatism associated with the flaw-related variables chosen for analysis remains examination (NDE) technology and the recently formulated ASME Code procedures for qualifying NDE system capability and performance (as applied to selected nuclear power plant components) now provides a systematic means of bounding the conservatism in flaw-related input variables for pressure vessel integrity analyses. This is essentially achieved by establishing probabilistic (risk)-based limits on the assigned variable values, dependent upon the vessel inspection results and on the inspection system unreliability. Described herein is this probabilistic method and its potential application to: (i) defining a vessel-specific ''reference'' flaw for calculating pressure-temperature limit curves in the deterministic evaluation of pressurized water reactor (PWR) reactor vessels, and (ii) limiting the flaw distribution input to a PWR reactor vessel-specific, probabilistic integrity analysis for pressurized thermal shock loads

Methodology for failure assessment of SMART SG tube with once-through helical-coiled type

International Nuclear Information System (INIS)

Kim, Young Jin; Choi, Shin Beom; Cho, Doo Ho; Chang, Yoon Suk

2010-09-01

In this research project, existing integrity evaluation method for SMART steam generator tube with crack-like flaw was reviewed to determine subject analysis model and investigate possibility of failure under crack closure behavior. Furthermore, failure pressure estimation was proposed for SMART steam generator tubes containing wear-type defects. For each subject, the following issues are addressed: 1. Determination of subject analysis model for SMART SG tube contaning crack-like flaw 2. Applicability review on existing integrity evaluation method and investigation of failure possibility for SMART SG tube containing crack-like flaw 3. Development of failure pressure estimation model for SMART SG tube with wear type defect It is anticipated that if the technologies developed in this study are applied, structural integrity can be estimated accurately

Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

International Nuclear Information System (INIS)

Kim, Jong Sung; Oh, Young Jin

2014-01-01

If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

Eddy Current Flaw Characterization Using Neural Networks

International Nuclear Information System (INIS)

Song, S. J.; Park, H. J.; Shin, Y. K.

1998-01-01

Determination of location, shape and size of a flaw from its eddy current testing signal is one of the fundamental issues in eddy current nondestructive evaluation of steam generator tubes. Here, we propose an approach to this problem; an inversion of eddy current flaw signal using neural networks trained by finite element model-based synthetic signatures. Total 216 eddy current signals from four different types of axisymmetric flaws in tubes are generated by finite element models of which the accuracy is experimentally validated. From each simulated signature, total 24 eddy current features are extracted and among them 13 features are finally selected for flaw characterization. Based on these features, probabilistic neural networks discriminate flaws into four different types according to the location and the shape, and successively back propagation neural networks determine the size parameters of the discriminated flaw

Development of a 3D electromagnetic model for eddy current tubing inspection application to steam generator tubing

Energy Technology Data Exchange (ETDEWEB)

Maillot, V. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Pichenot, G.; Premel, D.; Sollier, T. [CEA Saclay, DRT/DECS, 91 - Gif-sur-Yvette (France)

2003-10-01

In nuclear plants, the inspection of heat exchanger tubes is usually carried out by using eddy current nondestructive testing. A numerical model, based on a volume integral approach using the Green's dyadic formalism, has been developed, with support from the French Institute for Radiological Protection and Nuclear Safety, to predict the response of an eddy current bobbin coil to 3D flaws located in the tube's wall. With an aim of integrating this model into the NDE multi techniques platform CIVA, it has been validated with experimental data for 2D and 3D flaws. (authors)

Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

Energy Technology Data Exchange (ETDEWEB)

Bahn, Chi Bum, E-mail: bahn@pusan.ac.kr [Pusan National University, 2 Busandaehak-ro 63 beon-gil, Geumjeong-gu, Busan 609-735 (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering & Construction Co. Inc., Seongnam 463-870 (Korea, Republic of); Majumdar, Saurin [Argonne National Laboratory, Lemont, IL 60439 (United States)

2015-11-15

Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

International Nuclear Information System (INIS)

Bahn, Chi Bum; Oh, Young-Jin; Majumdar, Saurin

2015-01-01

Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

Examination of parameters affecting overload fracture behavior of flaw-tip hydrides in Zr-2.5Nb pressure tubes in Candu reactors

International Nuclear Information System (INIS)

Cui, J.; Shek, G.K.; Wang, Z.R.

2007-01-01

Service-induced flaws in Zr-2.5Nb alloy pressure tubes in Candu (Canada Deuterium Uranium Reactors) nuclear reactors are susceptible to a crack initiation and growth mechanism known as Delayed Hydride Cracking (DHC), which is a repetitive process that involves hydrogen diffusion, hydride precipitation, growth and fracture of a hydride region at the flaw-tip under a constant load. Crack initiation may also occur under another loading condition when the hydride region is subjected to an overload. An overload occurs when the hydride region at the flaw tip is loaded to a stress higher than that at which this region is formed such as when the reactor experiences a transient pressure higher than the normal operating pressure where the hydride region is formed. Flaw disposition requires justification that the hydride region overload will not fracture the hydride region, and initiate DHC. In this work, monotonically increasing load experiments were performed on unirradiated Zr-2.5Nb pressure tube specimens containing simulated debris frets (V-notch) and bearing pad frets (BPF, U-shape notch) to examine overload fracture behavior of flaw-tip hydrides formed under hydride ratcheting conditions. Hydride cracking in the overload tests was detected by the acoustic emission technique and confirmed by post-test metallurgical examination. Test results indicate that the resistance to overload fracture is affected by a number of parameters including hydride formation stress, flaw shape (V-notch vs. BPF) and flaw radius (0.015 mm vs. 0.1 mm). The notch-tip hydride morphologies were examined by optical microscopy and scanning electron microscopy (SEM) which show that they are affected by the hydride formation conditions, resulting in different overload fracture resistance. Finite element stress analyses were also performed to obtain flaw-tip stress distributions for interpretation of the test results. (authors)

Flaw distributions and use of ISI data in RPV integrity evaluations

International Nuclear Information System (INIS)

Dimitrijevic, V.; Ammirato, F.

1993-01-01

A probabilistic method for developing post-inspection flaw distributions has been developed that explicitly accounts for the capability of the inspection procedure to detect and size flaws. This methodology has been used to develop flaw distributions for calculating reactor vessel failure probability under postulated pressurized thermal shock (PTS) conditions. Realistic flaw distributions are important because plant-specific PTS safety assessments are very sensitive to assumptions made about major flaw parameters such as density, size, shape, and location. PTS analysis made in the past do not consider ISI. Two main reasons are (1) lack of a general and approved methodology which provides directions for involvement of ISI results in developing new flaw parameters and (2) lack of confidence in the capability of ISI procedures to detect critical flaws that may be present near the clad-to-base metal interface of the vessel, the location of most concern for PTS conditions. Recent developments in ISI practice, however, have led to substantial improvement in ISI capability and provide a basis for using ISI data to develop plant-specific post-inspection flaw distributions for vessel integrity evaluations. The key components of this evaluation are (1) the generic (preinspection) flaw distribution, (2) a probabilistic flaw detection model, and (3) Bayesian updating of the prior flaw distribution with the detection model to develop a post-inspection flaw distribution. Destructive analysis of RPV weld material was performed to develop data to support the pre-inspection flaw distributions. Since the probability of detection (POD) plays such an important role in the analysis and a high POD is needed to make significant reductions in probability of failure, a procedure was developed to achieve and demonstrate POD greater than 0.9 by using a combination of independent inspection techniques

Evaluation of J-integral estimation scheme for flawed throughwall pipes

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.

1987-02-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses.

Evaluation of J-integral estimation scheme for flawed throughwall pipes

International Nuclear Information System (INIS)

Zahoor, A.

1987-01-01

The accuracy of the EPRI J-integral estimation scheme for pipes with throughwall cracks and subjected to pure bending was assessed using available experimental data on circumferentially flawed throughwall pipes. The evaluations were performed using elastic plastic J-integral (J) and tearing modulus (T) analysis methods. The results indicated that the EPRI J estimation scheme solutions are unnecessarily conservative compared to results from pipe experiments. As a result of these evaluations an improved J estimation scheme is developed, which is shown to have improved accuracy compared to the original EPRI J estimation scheme. These results imply that the flaw evaluation procedures in the ASME Code on austenitic piping welds are conservative. These results also have applications to the leak before break fracture mechanics analyses. (orig.)

Experimental study on flaw detectability of remote field eddy current testing

International Nuclear Information System (INIS)

Kamimura, T.; Fukui, S.; Iwahashi, Y.; Yamada, H.

1988-01-01

For the purpose of comprehending the effect in practical use of the remote field eddy current (RFEC) testing that becomes noticeable for the ISI technique of steel tubes, its flaw detectability was clarified through a model test. This study used straight and bending tubes of 3.8 mm in wall thickness and 31.8 mm in outside diameter. These tubes were inspected from their inside. After relations among the pickup coil output, coil distance, testing frequency, etc. were measured, a probe of the practical use type was manufactured to investigate its flaw detectability by means of simulated flaws. The authors discuss how it has been found that light local wall thinning on the outside surfaces can be detected by this technique and its effect in practical use can be expected with small influences due to magnetic permeability variations of tube materials, bending of tubes, etc

Sample summary report for ARG 1 pressure tube sample

International Nuclear Information System (INIS)

Belinco, C.

2006-01-01

The ARG 1 sample is made from an un-irradiated Zr-2.5% Nb pressure tube. The sample has 103.4 mm ID, 112 mm OD and approximately 500 mm length. A punch mark was made very close to one end of the sample. The punch mark indicates the 12 O'clock position and also identifies the face of the tube for making all the measurements. ARG 1 sample contains flaws on ID and OD surface. There was no intentional flaw within the wall of the pressure tube sample. Once the flaws are machined the pressure tube sample was covered from outside to hide the OD flaws. Approximately 50 mm length of pressure tube was left open at both the ends to facilitate the holding of sample in the fixtures for inspection. No flaw was machined in this zone of 50 mm on either end of the pressure tube sample. A total of 20 flaws were machined in ARG 1 sample. Out of these, 16 flaws were on the OD surface and the remaining 4 on the ID surface of the pressure tube. The flaws were characterized in to various groups like axial flaws, circumferential flaws, etc

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

International Nuclear Information System (INIS)

2006-05-01

Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

Development of technology on natural flaw fabrication and precise diagnosis for the major components in NPPs

International Nuclear Information System (INIS)

Han, Jung Ho; Choi, Myung Sik; Lee, Doek Hyun; Hur, Do Haeng

2002-01-01

The objective of this research is to develop a fabrication technology of natural flaw specimen of major components in NPPs and a technology of precise diagnosis for failure and degradation of components using natural flaw specimen. 1) Successful development of the natural flaw fabrication technology of SG tube 2) Evaluation of ECT signal and development of precise diagnosis using natural flaws. - Determination of length, depth, width, and multiplicity of fabricated natural flaws. - Informations about detectability and accuracy of ECT evaluation on various kinds of defects are collected when the combination of probe and frequency is changed. - An advanced technology for precise ECT evaluation is established. 3) Application of precise ECT diagnosis to failure analysis of SG tube in operation. - Fretting wear of KSNP SG. - ODSCC at tube expanded region of KSNP SG. - Determination of through/non-through wall of axial crack

Methodology for inferring initial flaw distribution

International Nuclear Information System (INIS)

Jouris, G.M.; Shaffer, D.H.

1980-01-01

It has been common practice in both deterministic and probabilistic assessment of the integrity of a pressure vessel to assume the presence of a rather large flaw (usually 1/4 the thickness of the vessel wall) in the belt-line region. Although it is highly unlikely that such a large flaw would be present, the assumption is adopted in order to be conservative. A more realistic approach, which can be incorporated in the probabilistic analysis of integrity, is to characterize the depth of a flaw as a random variable and thus allow the probabilities associated with the presence of various size flaws to be reflected in the final estimated probability of vessel failure. This is precisely the motivation for developing the methodology to obtain the distribution of initial flaw depth, which is presented in this paper. It should be mentioned that the methodology developed here is not an end in itself but rather provides an input distribution to be used in a comprehensive integrity assessment. (orig.)

Evaluation of ECT reliability for axial ODSCC in steam generator tubes

International Nuclear Information System (INIS)

Lee, Jae Bong; Park, Jai Hak; Kim, Hong Deok; Chung, Han Sub

2010-01-01

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 ).

Technical basis for the CANDU steam generator tube fitness-for-service guidelines

International Nuclear Information System (INIS)

Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

2002-01-01

Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

Nuclear reactor pressure vessel-specific flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.

1992-01-01

Vessel integrity predictions performed through fracture mechanics analysis of a pressurized thermal shock event have been shown to be significantly sensitive to the overall flaw distribution input. It has also been shown that modem vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. Throughout the program, new insight was obtained into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. For example, the potential application of a vessel-specific flaw distribution now provides at least one method by which a vessel-specific reference flaw size applicable to pressure-temperature limit curves determination can be estimated. This paper will discuss the development and application of the methodology and the impact to future vessel integrity analyses

Structural integrity assessments of steam generator tubes using the FAD methodology

Energy Technology Data Exchange (ETDEWEB)

Bergant, Marcos A., E-mail: marcos.bergant@cab.cnea.gov.ar [Gerencia CAREM, Centro Atómico Bariloche (CNEA), Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Yawny, Alejandro A., E-mail: yawny@cab.cnea.gov.ar [División Física de Metales, Centro Atómico Bariloche (CNEA)/CONICET, Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Perez Ipiña, Juan E., E-mail: juan.perezipina@fain.uncoma.edu.ar [Grupo Mecánica de Fractura, Universidad Nacional del Comahue/CONICET, Buenos Aires 1400, Neuquén 8300 (Argentina)

2015-12-15

Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

Structural integrity assessments of steam generator tubes using the FAD methodology

International Nuclear Information System (INIS)

Bergant, Marcos A.; Yawny, Alejandro A.; Perez Ipiña, Juan E.

2015-01-01

Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

Flaw evaluation charts

International Nuclear Information System (INIS)

Korosec, D.; Vojvodic Tuma, J.

1999-01-01

The structural integrity of the primary components in pressurized water reactor nuclear power plant is very important in the respect of safe and efficient operation. These components have to be subjected to periodic controls. In the light of fracture mechanics concept, the acceptance criteria for defects (flaws) are developed. Flaw evaluation procedure is necessary, to evaluate the defects regarding their acceptability for further operation. The objective of the flaw evaluation charts is to provide a series of simple graphs as decision maps. that immediate decision may be taken regarding the acceptability of a detected defects, on the basis of ASME Code XI criteria.(author)

Nuclear reactor pressure vessel flaw distribution development

International Nuclear Information System (INIS)

Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

1991-12-01

Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

Steam generator tube integrity program

International Nuclear Information System (INIS)

Dierks, D.R.; Shack, W.J.; Muscara, J.

1996-01-01

A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

3-D fracture analysis using a partial-reduced integration scheme

International Nuclear Information System (INIS)

Leitch, B.W.

1987-01-01

This paper presents details of 3-D elastic-plastic analyses of axially orientated external surface flaw in an internally pressurized thin-walled cylinder and discusses the variation of the J-integral values around the crack tip. A partial-reduced-integration-penalty method is introduced to minimize this variation of the J-integral near the crack tip. Utilizing 3-D symmetry, an eighth segment of a tube containing an elliptically shaped external surface flaw is modelled using 20-noded isoparametric elements. The crack-tip elements are collapsed to form a 1/r stress singularity about the curved crack front. The finite element model is subjected to internal pressure and axial pressure-generated loads. The virtual crack extension method is used to determine linear elastic stress intensity factors from the J-integral results at various points around the crack front. Despite the different material constants and the thinner wall thickness in this analysis, the elastic results compare favourably with those obtained by other researchers. The nonlinear stress-strain behaviour of the tube material is modelled using an incremental theory of plasticity. Variations of the J-integral values around the curved crack front of the 3-D flaw were seen. These variations could not be resolved by neglecting the immediate crack-tip elements J-integral results in favour of the more remote contour paths or else smoothed out when all the path results are averaged. Numerical incompatabilities in the 20-noded 3-D finite elements used to model the surface flaw were found. A partial-reduced integration scheme, using a combination of full and reduced integration elements, is proposed to determine J-integral results for 3-D fracture analyses. This procedure is applied to the analysis of an external semicircular surface flaw projecting halfway into the tube wall thickness. Examples of the J-integral values, before and after the partial-reduced integration method is employed, are given around the

Instruments for non-destructive evaluation of advanced test reactor inpile tubes

International Nuclear Information System (INIS)

Livingston, R.A.; Beller, L.S.; Edgett, S.M.

1986-01-01

The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

Flaw analysis in steam generator tube

International Nuclear Information System (INIS)

Hutin, J.P.; Billon, F.

1985-08-01

Operating more than 30 PWR units, Electricite de France has to face several steam generator tube problems. One of the most serious difficulties is the stress corrosion cracking due to primary fluid, just above the tube sheet, in the roll transition region. With regard to availability it is, of course, a major concern; with regard to safety, the point is that tube rupture should be preceded by a significant primary-to-secondary leak during normal operation so that the reactor should be shut down before failure occurs. The demonstration of this assessment asks for experimental and analytical evidences. In 1981, Elecricite de France started a comprehensive program on that subject. A general description of this program and the main results are to be presented during the SMIRT-8 Conference. The purpose of the present paper is to develop in greater detail the analytical part of the work

Fabrication Flaw Density and Distribution in Weld Repairs

International Nuclear Information System (INIS)

Doctor, Steven R.

2009-01-01

The Pacific Northwest National Laboratory (PNNL) is developing a generalized flaw distribution for the population of nuclear reactor pressure vessels and for piping welds in the U. S. operating reactors. The purpose of the generalized flaw distribution is to predict component-specific flaw densities. The estimates of fabrication flaws are intended for use in fracture mechanics structural integrity assessments. Structural integrity assessments, such as estimating the frequency of loss-of-coolant accidents, are performed by computer codes that require, as input, accurate estimates of flaw densities. Welds from four different cancelled reactor pressure vessels and a collection of archived pipes have been studied to develop empirical estimates of fabrication flaw densities. This paper describes the fabrication flaw distribution and characterization in the repair weld metal of vessels and piping. This work indicates that large flaws occur in these repairs which are complex in composition and sometimes include cracks on the ends of the repair cavities. Parametric analysis using an exponential fit is performed on the data. Construction records where available were reviewed. It is difficult to make conclusions due to the limited number of construction records reviewed. However, the records reviewed to date show a significant change in repair frequency over the years when the components in this study were fabricated. A description of repair flaw morphology is provided with a discussion of fracture mechanics significance.

Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1] [Sample summary reports of pressure tube samples from Argentina, India, Canada, Republic of Korea, and Romania

International Nuclear Information System (INIS)

2006-05-01

Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

Real time flaw detection and characterization in tube through partial least squares and SVR: Application to eddy current testing

Science.gov (United States)

Ahmed, Shamim; Miorelli, Roberto; Calmon, Pierre; Anselmi, Nicola; Salucci, Marco

2018-04-01

This paper describes Learning-By-Examples (LBE) technique for performing quasi real time flaw localization and characterization within a conductive tube based on Eddy Current Testing (ECT) signals. Within the framework of LBE, the combination of full-factorial (i.e., GRID) sampling and Partial Least Squares (PLS) feature extraction (i.e., GRID-PLS) techniques are applied for generating a suitable training set in offine phase. Support Vector Regression (SVR) is utilized for model development and inversion during offine and online phases, respectively. The performance and robustness of the proposed GIRD-PLS/SVR strategy on noisy test set is evaluated and compared with standard GRID/SVR approach.

Development of an integrated database management system to evaluate integrity of flawed components of nuclear power plant

International Nuclear Information System (INIS)

Mun, H. L.; Choi, S. N.; Jang, K. S.; Hong, S. Y.; Choi, J. B.; Kim, Y. J.

2001-01-01

The object of this paper is to develop an NPP-IDBMS(Integrated DataBase Management System for Nuclear Power Plants) for evaluating the integrity of components of nuclear power plant using relational data model. This paper describes the relational data model, structure and development strategy for the proposed NPP-IDBMS. The NPP-IDBMS consists of database, database management system and interface part. The database part consists of plant, shape, operating condition, material properties and stress database, which are required for the integrity evaluation of each component in nuclear power plants. For the development of stress database, an extensive finite element analysis was performed for various components considering operational transients. The developed NPP-IDBMS will provide efficient and accurate way to evaluate the integrity of flawed components

Advanced NDE (ANDE) and its application for pressure tube inspections in OPG reactors

Energy Technology Data Exchange (ETDEWEB)

Jarron, D.; Trelinski, M.; Kretz, S. [Ontario Power Generation, Ajax, Ontario (Canada)]. E-mail: don.jarron@opg.com; mike.trelinski@opg.com; steve.kretz@opg.com

2006-07-01

Periodic and in-service inspections of CANDU fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections (Universal Delivery Machine) has driven new methods for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) system has been designed and field implemented as a high speed data acquisition system to meet the requirements of the CSA N285.4 code. It was built from the solid foundation of CIGAR experience and uses cutting edge hardware and software to attain high speed data collection enabling relatively quick inspection of a large number of fuel channels. The capabilities of the ANDE inspection system include: Surface and volumetric inspection of pressure tube by ultrasonics; Flaw characterization by ultrasonics; Pressure tube diameter measurements; Pressure tube thickness measurements; Garter Spring location by Eddy Current; Garter Spring location by ultrasonics; Pressure tube sag measurement. In addition to the above, selected flaws/areas of a pressure tube can be replicated using a two plate ANDE replica tool. At the heart of the inspection system is a set of twelve ultrasonic probes positioned in such a way that the inspected areas are examined from various angles and directions and by various ultrasonic wave modes (shear and longitudinal). High frequency ultrasound used for the examinations allows for reliable detection of small flaws. Separate sensors have been installed on the inspection head for Garter Spring location and sag measurements. (author)

Integrity Assessment of GOH Heater Tube

Energy Technology Data Exchange (ETDEWEB)

Lee, Bong Sang; Hong, J. H.; Oh, Y. J.; Yoon, J. H.; Oh, J. M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1997-08-01

An assessment of structural integrity of ASTM A312-TP347 GOH heater tube was performed. The surface notches which had been produced during tube manufacturing process were analyzed microscopically. Chemical analysis, hardness tests, tensile tests, and J-Integral fracture resistance tests were carried out to compare the mechanical properties with those of the material specification and also with the other material of the same type. The test results showed the mechanical properties of the GOH tube material are within the specification range. An elastic-plastic fracture mechanics analysis based on the DPFAD method reveals the tube an appropriate safety margin for the normal operation. 13 refs., 5 tabs., 24 figs. (author)

An experimental study of ECT for fin-type copper alloy tubes

International Nuclear Information System (INIS)

Lee, Hyung Joon; Lee, Jeong Soon; Sung, Je Joong; Park, Cheon Woong; Suh, Dong Man; Yu, Taek In

2002-01-01

Eddy current detecting probes with inner and encircling coils were designed for the fin-type tubes that have uneven outer and inner surface to enhance the efficiency of heat emission. As the uneven surface of them, it is difficult to detect flaws in the tubes by eddy current test. In this paper, standard and artificial specimens with flaws for the different types of the tubes were manufactured. Eddy current test was performed with the designed probes, which have inner and encircling coils, for the prepared specimens. From the signals of the eddy current detecting probes, the phase and amplitude variation were analyzed and the best conditions of the flaw detection for the tubes were found.

Evaluation of canister weld flaw depth for concrete storage cask

Energy Technology Data Exchange (ETDEWEB)

Moon, Tae Chul; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of); Jung, Sung Hun; Lee, Young Oh; Jung, In Su [Korea Nuclear Engineering and Service Corp, Daejeon (Korea, Republic of)

2017-03-15

Domestically developed concrete storage casks include an internal canister to maintain the confinement integrity of radioactive materials. In this study, we analyzed the depth of flaws caused by loads that propagate canister weld cracks under normal, off-normal and accident conditions, and evaluated the maximum allowable weld flaw depth needed to secure the structural integrity of the canister weld and to reduce the welding time of the internal canister lid of the concrete storage cask. Structural analyses for normal, off-normal and accident conditions were performed using the general-purpose finite element analysis program ABAQUS; the allowable flaw depth was assessed according to ASME B and PV Code Section XI. Evaluation results revealed an allowable canister weld flaw depth of 18.75 mm for the concrete storage cask, which satisfies the critical flaw depth recommended in NUREG-1536.

Development and application of an LWR reactor pressure vessel-specific flaw distribution

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.

1991-01-01

Previous efforts by the US Department of Energy have shown that the PWR reactor vessel integrity predictions performed through probabilistic fracture mechanics analysis for a pressurized thermal shock event are significantly sensitive to the overall flaw distribution input. It has also been shown that modern vessel in-service inspection (ISI) results can be used for development of vessel flaw distribution(s) that are more representative of US vessels. This paper describes the development and application of a methodology to analyze ISI data for the purpose of flaw distribution determination. The resultant methodology considers detection reliability, flaw sizing accuracy, and flaw detection threshold in its application. Application of the methodology was then demonstrated using four recently acquired US PWR vessel inspection data sets. The methodology helped provide original insight into several key inspection performance and vessel integrity prediction practice issues that will impact future vessel integrity evaluation. This paper briefly discusses the development and application of the methodology and the impact to future vessel integrity analyses

Flaw distribution development from vessel ISI data

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.; Basin, S.L.; Rosinski, S.T.

1991-01-01

Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed

Methods to establish flaw tolerances

International Nuclear Information System (INIS)

Varga, T.

1978-01-01

Three conventional methods used to establish flaw tolerances are compared with new approaches using fracture mechanics. The conventional methods are those based on (a) non-destructive testing methods; (b) fabrication and quality assurance experience; and (c) service and damage experience. Pre-requisites of fracture mechanics methods are outlined, and summaries given of linear elastic mechanics (LEFM) and elastoplastic fracture mechanics (EPFM). The latter includes discussion of C.O.D.(crack opening displacement), the J-integral and equivalent energy. Proposals are made for establishing flaw tolerances. (U.K.)

Insensitivity to Flaws Leads to Damage Tolerance in Brittle Architected Meta-Materials

Science.gov (United States)

Montemayor, L. C.; Wong, W. H.; Zhang, Y.-W.; Greer, J. R.

2016-02-01

Cellular solids are instrumental in creating lightweight, strong, and damage-tolerant engineering materials. By extending feature size down to the nanoscale, we simultaneously exploit the architecture and material size effects to substantially enhance structural integrity of architected meta-materials. We discovered that hollow-tube alumina nanolattices with 3D kagome geometry that contained pre-fabricated flaws always failed at the same load as the pristine specimens when the ratio of notch length (a) to sample width (w) is no greater than 1/3, with no correlation between failure occurring at or away from the notch. Samples with (a/w) > 0.3, and notch length-to-unit cell size ratios of (a/l) > 5.2, failed at a lower peak loads because of the higher sample compliance when fewer unit cells span the intact region. Finite element simulations show that the failure is governed by purely tensile loading for (a/w) meta-materials may give rise to their damage tolerance and insensitivity of failure to the presence of flaws even when made entirely of intrinsically brittle materials.

PWR vessel flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

1990-01-01

This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

Working session 3: Tubing integrity

International Nuclear Information System (INIS)

Cueto-Felgueroso, C.; Strosnider, J.

1997-01-01

Twenty-three individuals representing nine countries (Belgium, Canada, the Czech Republic, France, Japan, the Slovak Republic, Spain, the UK, and the US) participated in the session on tube integrity. These individuals represented utilities, vendors, consultants and regulatory authorities. The major subjects discussed by the group included overall objectives of managing steam generator tube degradation, necessary elements of a steam generator degradation management program, the concept of degradation specific management, structural integrity evaluations, leakage evaluations, and specific degradation mechanisms. The group's discussions on these subjects, including conclusions and recommendations, are summarized in this article

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D. R.; Majumdar, S.; Kupperman, D. S.; Bakhtiari, S.; Shack, W. J.

2001-01-01

Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators

Ultrasonic inspection experience of steam generator tubes at Ontario Hydro and the TRUSTIE inspection system

International Nuclear Information System (INIS)

Choi, E.I.; Jansen, D.

1998-01-01

Ontario Hydro have been using ultrasonic test (UT) technique to inspect steam generator (SG) tubes since 1993. The UT technique has higher sensitivity in detecting flaws in SG tubes and can characterize the flaws with higher accuracy. Although an outside contractor was used initially, Ontario Hydro has been using a self-developed system since 1995. The TRUSTIE system (Tiny Rotating UltraSonic Tube Inspection Equipment) was developed by Ontario Hydro Technologies specifically for 12.7 mm outside diameter (OD) tubes, and later expanded to larger tubes. To date TRUSTIE has been used in all of Ontario Hydro's nuclear generating stations inspecting for flaws such as pitting, denting, and cracks at top-of-tubesheet to the U-bend region. (author)

A review of recent advances in the role of leak-before-break concept in assessments of flaws detected in CANDU pressure tubes

International Nuclear Information System (INIS)

Crespi, J.C.

1994-01-01

If a crack develops in a pressure tube, the leak is detected by monitoring the moisture in the gas annulus and the reactor shutdown before it becomes unstable. Because the delayed hydride cracking has been associated to date with all pressure tube failures at a rolled joints, the delayed hydride cracking is considered to be the dominant mecanism by which the flaws can grow to a size which exceeds the critical crack length. For the delayed hydride cracking failure mode leak-before-break is used as defense in depth against unstable rupture. The methodology depends on showing than the time available to detect a delayed hydride crack is much greater that the time required to detect it in the gas annulus. The time available is estimated from measurements of: (a) axial delayed hydride crack growth rates, (b) crack lengths at penetrations of the tube wall when leakage first occurs and (c) critical crack lengths at instability when a crack is growing by the delayed hydride cracking mechanism. A review of recent advances in the experimental data used in leak-before-break assessment are presented and discussed. (author). 17 refs, 6 figs, 2 tabs

Is tube feeding futile in advanced dementia?

Science.gov (United States)

Lynch, Matthew C.

2016-01-01

It is controversial whether tube feeding in people with dementia improves nutritional status or prolongs survival. Guidelines published by several professional societies cite observational studies that have shown no benefit and conclude that tube feeding in patients with advanced dementia should be avoided. However, all studies on tube feeding in dementia have major methodological flaws that invalidate their findings. The present evidence is not sufficient to justify general guidelines. Patients with advanced dementia represent a very heterogeneous group, and evidence demonstrates that some patients with dementia benefit from tube feeding. However, presently available guidelines make a single recommendation against tube feeding for all patients. Clinicians, patients, and surrogates should be aware that the guidelines and prior commentary on this topic tend both to overestimate the strength of evidence for futility and to exaggerate the burdens of tube feeding. Shared decision making requires accurate information tailored to the individual patient's particular situation, not blanket guidelines based on flawed data. Lay Summary: Many doctors believe that tube feeding does not help people with advanced dementia. Scientific studies suggest that people with dementia who have feeding tubes do not live longer or gain weight compared with those who are carefully hand fed. However, these studies are not very helpful because of flaws in design, which are discussed in this article. Guidelines from professional societies make a blanket recommendation against feeding tubes for anyone with dementia, but an individual approach that takes each person's situation into account seems more appropriate. Patients and surrogates should be aware that the guidelines on this topic tend both to underestimate the benefit and exaggerate the burdens of tube feeding. PMID:27833208

Systematization of simplified J-integral evaluation method for flaw evaluation at high temperature

International Nuclear Information System (INIS)

Miura, Naoki; Takahashi, Yukio; Nakayama, Yasunari; Shimakawa, Takashi

2000-01-01

J-integral is an effective inelastic fracture parameter for the flaw evaluation of cracked components at high temperature. The evaluation of J-integral for an arbitrary crack configuration and an arbitrary loading condition can be generally accomplished by detailed numerical analysis such as finite element analysis, however, it is time-consuming and requires a high degree of expertise for its implementation. Therefore, it is important to develop simplified J-integral estimation techniques from the viewpoint of industrial requirements. In this study, a simplified J-integral evaluation method is proposed to estimate two types of J-integral parameters. One is the fatigue J-integral range to describe fatigue crack propagation behavior, and the other is the creep J-integral to describe creep crack propagation behavior. This paper presents the systematization of the simplified J-integral evaluation method incorporated with the reference stress method and the concept of elastic follow-up, and proposes a comprehensive evaluation procedure. The verification of the proposed method is presented in Part II of this paper. (author)

Improved in-service inspection program for management of degradation in steam generator tubing

International Nuclear Information System (INIS)

Kurtz, R.; Heasler, P.; Muscara, J.

1992-01-01

This paper presents an overview of significant results from NRC-sponsored research on steam generator tube integrity and inspection. Burst test results are described along with empirical models to relate flaw geometry and size to tube burst pressure. Results of round robin examinations of a retired-from-service steam generator to determine eddy current inspection reliability are presented. An evaluation and comparison of various sampling plans for in-service inspection of steam generators is discussed. Finally, performance demonstration qualification efforts for eddy current inspection systems are described

Steam generator tube rupture simulation using extended finite element method

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

2016-08-15

Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

Steam generator tube rupture simulation using extended finite element method

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

2016-01-01

Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

Characteristics Testing of the ECT Bobbin Probe for S/G Tube Inspection

Energy Technology Data Exchange (ETDEWEB)

Nam, Min Woo; Lee, Hee Jong; Cho, Chan Hee; Yoo, Hyun Joo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2010-05-15

The bobbin probe technique is basically one of the important ECT methods for the steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe also is the essential component which consists of the whole ECT examination system, and provides a decisive data for the evaluation of tube integrity in compliance with acceptance criteria described in specific procedures. The selection of probe is especially important because the quality of acquired ECT data is determined by the probe design characteristics, such as geometry and operation frequency, and has an important effect on examination results. The Electric Power Research Institute (EPRI) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. Flawed tube removed from operation, or artificial flaw is required for the original qualification of technique combined with related flaw mechanism. In case where the original qualification has been completed, the concept of equivalency may be used to extend the original qualification to similar probe designs. The qualified acquisition technique may be modified to substitute or replace instruments or probes without re-qualification provided that the range of essential variables defined in the examination technique specification sheet are met. In this case, both the original and replaced instrument or probe shall be characterized utilizing EPRI Guideline supplement 'H1'. This study is the result of the comparative performance evaluation of bobbin coil eddy current probes manufactured by KEPRI and a foreign manufacturer. As a result of this study, although there were minor differences between the two probe types, it was evaluated that the two probes were almost identical in the significant performance characteristics described in the EPRI guideline

Burst pressure and leak rate from fretted SG tubes

International Nuclear Information System (INIS)

Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

2005-01-01

Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

Status of steam generator tubing integrity at Jaslovske Bohunice NPP

International Nuclear Information System (INIS)

Cepcek, S.

1997-01-01

Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented

Stochastic modeling of inspection uncertainties and applications to pitting flaws in steam generator tubes

International Nuclear Information System (INIS)

Mao, D.; Yuan, X.-X.; Pandey, M.D.

2009-01-01

Steam generators (SG) are a major pressure retaining component of great safety significance in nuclear power plants. Due to various manufacturing, operation and maintenance activities, as well as material interaction with the surrounding chemical environment, the SG tubes have been subject to a number of degradation modes. Among them, the under-deposit pitting corrosion at outside surfaces of the SG tubes just on top of the tubesheet support plates has had a serious impact on the integrity of the SG tubes. This paper presents an advanced probabilistic model of pitting corrosion characterizing the inherent randomness of the pitting process and measurement uncertainties of the in-service inspection (ISI) data obtained from eddy current (EC) inspections. A Bayesian method based on Markov Chain Monte Carlo (MCMC) simulation is developed for estimating the model parameters. The proposed model is able to predict the actual pit number, the actual pit depth as well as the maximum pit depth, which is the main interest of the pitting corrosion model. (author)

Radiation flaw detector for testing non-uniform surface bodies of revolution

International Nuclear Information System (INIS)

Valevich, M.I.

1984-01-01

Radiation flaw detector for testing bodies of revolution with non-uniform surface, welded joints, etc., based on spatial filtration and differentiation of ionizing radiation flux has been described. The calculation of the most important unit of flaw detector - integrators - is made. Experimental studies of the sensitivity have shown, that the radiation flaw detector can be used for rapid testing of products with the sensitivity comparable with the sensitivity of radiographic testing of steel

Design Concept of Array ECT Sensor for Steam Generator Tubing Inspection

Energy Technology Data Exchange (ETDEWEB)

Cho, Chan Hee; Lee, Tae Hun; Yoo, Hyun Ju [Korea Hydro and Nuclear Power Co. Ltd. CRI, Daejeon (Korea, Republic of)

2015-05-15

The eddy current testing, which is one of the nondestructive examination methods, is widely used for the inspection of heat exchangers including steam generator tubing in the nuclear power plant. It uses electromagnetic induction to detect flaws in conductive materials. Two types of eddy current probes are conventionally used for the inspection of steam generator tubing according to the main purpose. One is the bobbin probe technology and the other is the rotating probe. During the inspection, they have restrictions for the flaw detection or the inspection speed. An array probe can be alternative to the bobbin and rotating probes. The design concept of array coils with high sensitivity is described in this paper. It is expected that the eddy current testing using this type of array sensors may provide high detectability and resolution for flaws in steam generator tubing. Eddy current technology has some barriers for the inspection of steam generator tubing in the nuclear power plant. Bobbin probes offer poor circumferential crack detection and rotating probes are time and money consuming due to the mechanical rotation. Array probe inspection technique can replace bobbin and rotating probe techniques due to its sensitivity for flaw detection and inspection speed. In general, circular-shaped coils are considered in an array eddy current probe.

Implementation status of performance demonstration program for steam generator tubing analysts in Korea

International Nuclear Information System (INIS)

Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo; Hong, Sung Yull

2013-01-01

Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%

Implementation status of performance demonstration program for steam generator tubing analysts in Korea

Energy Technology Data Exchange (ETDEWEB)

Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo [KHNP Central Research Institute, Daejeon (Korea, Republic of); Hong, Sung Yull [Yeungnam Univ., Gyeongsan (Korea, Republic of)

2013-02-15

Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.

Probabilistic assessment of flaw evaluation procedures for pressure vessel integrity

International Nuclear Information System (INIS)

Shaffer, D.H.; Bamford, W.H.; Jouris, G.M.

1980-01-01

Prudent design procedures, in order to err in the direction of conservative over-strength rather than risky under-strength, have taken bounding values rather than best estimates for material parameters, and wherever possible, used conservative input for the calculations. The growing data base for this work is now beginning to allow an assessment of the conservatism that has been incorporated into the design procedure. Quantitative estimates of the variability associated with crack growth rates and fracture toughness have been generated in connection with other studies, and it would be useful to incorporate such information into an overall assessment of the design margins that are prescribed. In addition to getting an estimate of the conservatism in the current procedure, this study should provide a useful insight into the relative degree of margin that is introduced at each stage of the flaw evaluation process. Identification of the step by step margins should lead to more effective data collection programs from which information for adequately controlling the design conservatism can be obtained. The study will also provide valuable guidance in fixing revised design reference curves and safety factors so that adequate overall margins can be maintained without excess conservatism. This study is limited to vessel rupture in a brittle mode, and examples for illustration are particularly related to the beltline region of a reactor pressure vessel. The methodology, however, is applicable to all regions for which the required stress analyses, operating history, and material parameters are available. The work being carried out here is in consonance with ASME Section XI on Flaw Evaluation Procedures. It is concerned both with flaws under normal operating conditions and flaws under faulted conditions. (author)

Dynamic characteristics of steam generator U-tubes with defect

International Nuclear Information System (INIS)

Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

2005-01-01

This study investigates the fluid elastic instability characteristics of steam generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed in this study is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

Steam generator tube integrity program: Phase II, Final report

Energy Technology Data Exchange (ETDEWEB)

Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

1988-08-01

The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.

Steam generator tube integrity program: Phase II, Final report

International Nuclear Information System (INIS)

Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

1988-08-01

The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted

Integrating YouTube into the nursing curriculum.

Science.gov (United States)

Sharoff, Leighsa

2011-08-17

Nurse educators need to be innovative, stimulating, and engaging as they teach future nursing professionals. The use of YouTube in nursing education classes provides an easy, innovative, and user-friendly way to engage today's nursing students. YouTube presentations can be easily adapted into nursing courses at any level, be it a fundamentals course for undergraduate students or a theoretical foundations course for graduate students. In this article I will provide information to help educators effectively integrate YouTube into their course offerings. I will start by reviewing the phenomenon of social networking. Next I will discuss challenges and strategies related to YouTube learning experiences, after which I will share some of the legal considerations in using YouTube. I will conclude by describing how to engage students via YouTube and current research related to YouTube.

Currency flaw severity. [Banknotes

Energy Technology Data Exchange (ETDEWEB)

Johnson, C.; Burnett, M.; Goodman, C.; Sherrod, R.; Schmoyer, R.; Harrison, C.; Uppuluri, R.

1986-01-01

A survey of currency flaw severity was carried out using 300 banknotes and 37 judges. Each judge assigned each note to one of five flaw severity categories. These categories correspond to severity grades of 1 to 5 with 1 equivalent to ''always accepted'' and 5 ''never accepted.'' An average flaw severity grade for each note was obtained by taking the mean of the severity grades assigned to that note by the 37 judges. Thus, each note has a single numerical real-number flaw grade between 1 and 5. Mathematical modeling of the currency flaw survey results is continuing with some very promising initial results. Our present model handles common excess ink and missing ink flaw types quite well. We plan to extend the model to ink level, mash, setoff and blanket impression flaw types.

Feasibility study on the guided wave technique for condenser tube in NPP

International Nuclear Information System (INIS)

Choi, Sung Nam; Kim, Young Ho; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, W. G.

2004-01-01

The condenser tube is examined by the eddy current test (ECT) method to identify the integrity of the nuclear power plant. Because ECT probe is moved through the tube inside to identify flaws, the ECT probe should be exchanged periodically due to the wear of probe surface in order to remove the noise form the ECT signal. Moreover, it is impossible to examine the tube by ECT method because the ECT probe can not move through the inside due to the deformation such as dent. Recently, the theory of guided wave was established and the equipment applying the theory has been actively developed so as to overcome the limitation of ECT method for the tube inspection of heater exchanger in nuclear power plant. The object of this study is to know the feasibility of applying the guided wave technique to condenser tube in NPP

Component flaw evaluations

Energy Technology Data Exchange (ETDEWEB)

Yoon, K [Babcock and Wilcox Co., Lynchburg, VA (United States). Nuclear Power Div.

1988-12-31

This document deals with flaw evaluation during in-service inspection. These flaws can be divided into two groups: defects originating from the manufacturing fabrication stage or service-induced flaws. These are mainly caused by high cycle thermal fatigue and are influenced by the presence of stress corrosion cracking mechanisms such as nozzles or pump shaft. (TEC).

Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

2012-12-15

For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term

Evaluation of steam generator tube integrity during earthquakes

Energy Technology Data Exchange (ETDEWEB)

Kusakabe, Takaya; Kodama, Toshio [Mitsubishi Heavy Industries Ltd., Kobe (Japan). Kobe Shipyard and Machinery Works; Takamatsu, Hiroshi; Matsunaga, Tomoya

1999-07-01

This report shows an experimental study on the strength of PWR steam generator (SG) tubes with various defects under cyclic loads which simulate earthquakes. The tests were done using same SG tubing as actual plants with axial and circumferential defects with various length and depth. In the tests, straight tubes were loaded with cyclic bending moments to simulate earthquake waves and number of load cycles at which tube leak started or tube burst was counted. The test results showed that even tubes with very long crack made by EDM more than 80% depth could stand the maximum earthquake, and tubes with corrosion crack were far stronger than those. Thus the integrity of SG tubes with minute potential defects was demonstrated. (author)

Vibration and wear characteristics of steam generator tubes

International Nuclear Information System (INIS)

Choi, Young Hwan

2003-06-01

This study investigates the fluid elastic instability characteristics of Steam Generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages on Steam Generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions for determining the fluid elastic instability or fretting-wear parameters such as damping ratio, added mass and flow velocity are obtained from three-dimensional SG flow calculation using the ATHOS3 code. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

Evaluation of flawed-pipe experiments: Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.

1986-11-01

The purpose of this work was to perform elastic plastic fracture mechanics evaluations of experimental data that have become available from the NRC Degraded Pipe Program, Phase II (DPII) and other NRC and EPRI sponsored programs. These evaluations were used to assess flaw evaluation procedures for austenitic and ferritic steel piping. The results also have application to leak before break fracture mechanics analysis. An improved relationship was developed for computing the J-Integral for pipes containing throughwall flaws and loaded in pure bending. The results from several DPII experiments were compared to predictions based on new J estimation scheme solutions for circumferential, finite length part-throughwall flaws in pipes with bending loading. Comparisons of experimental maximum loads with those predicted using procedures in Paragraph IWB-3640, Section XI of the ASME Code indicate that the Code flaw evaluation procedures and allowables for austenitic steel pipe are appropriate and conservative. However, the comparisons also indicate that the base metal Code allowable loads may be about 15 to 20% high for small diameter piping (less than 8-inch diameter) at allowable a/t larger than about 0.5. The work further indicates that there is justification for reducing the conservatism in IWB-3640 allowable flaw sizes and loads for austenitic steel pipe with submerged or shielded metal arc welds.

Ultrasonic simulation studies for sizing of planar flaws in thick carbon steel welds

International Nuclear Information System (INIS)

Prakash, Alok

2015-01-01

Ultrasonic non-destructive testing typically involves detection of flaws that may affect the integrity of component under test. Once detected, the flaw is sized for its critical dimensions and its nature. The detection of flaw in the component by ultrasonic test is based on the principle of echo or reflection. Once the echo from a flaw is received, there are several approaches for analyzing the signal so that more and accurate information is obtained on the size of the flaw and its nature. The 6dB drop method is commonly used for sizing of flaws. This technique is based on determining the end points where the ultrasonic signal amplitude from the flaw drops to half of the peak amplitude. Though this method works well for large flaws whose size is larger than the beam width, it has a tendency to oversize the flaw which is smaller than the beam dimensions. In addition to beam divergence, flaw sizing also depends upon the orientation of the flaw with respect to incident sound beam. The paper describes the results of simulation studies on ultrasonic response from planar flaws of various orientations, their imaging and the methodology to be adopted for their accurate depth sizing. The paper also describes the experimental results to validate the flaw sizing approach

Probabilistic methodology for assessing steam generator tube inspection - Phase II: CANTIA - a probabilistic method for assessing steam generator tube inspections

International Nuclear Information System (INIS)

Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

1997-03-01

The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the documentation and verification of the code is provided in this volume. The user's manual for CANTIA is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

Measuring weld heat to evaluate weld integrity

Energy Technology Data Exchange (ETDEWEB)

Schauder, V., E-mail: schauder@hks-prozesstechnik.de [HKS-Prozesstechnik GmbH, Halle (Germany)

2015-11-15

Eddy current and ultrasonic testing are suitable for tube and pipe mills and have been used for weld seam flaw detection for decades, but a new process, thermography, is an alternative. By measuring the heat signature of the weld seam as it cools, it provides information about weld integrity at and below the surface. The thermal processes used to join metals, such as plasma, induction, laser, and gas tungsten arc welding (GTAW), have improved since they were developed, and they get better with each passing year. However, no industrial process is perfect, so companies that conduct research in flaw detection likewise continue to develop and improve the technologies used to verify weld integrity: ultrasonic testing (UT), eddy current testing (ET), hydrostatic, X-ray, magnetic particle, and liquid penetrant are among the most common. Two of these are used for verifying the integrity of the continuous welds such as those used on pipe and tube mills: UT and ET. Each uses a transmitter to send waves of ultrasonic energy or electrical current through the material and a receiver (probe) to detect disturbances in the flow. The two processes often are combined to capitalize on the strengths of each. While ET is good at detecting flaws at or near the surface, UT penetrates the material, detecting subsurface flaws. One drawback is that sound waves and electrical current waves have a specific direction of travel, or an alignment. A linear defect that runs parallel to the direction of travel of the ultrasonic sound wave or a flaw that is parallel to the coil winding direction of the ET probe can go undetected. A second drawback is that they don't detect cold welds. An alternative process, thermography, works in a different fashion: It monitors the heat of the material as the weld cools. Although it measures the heat at the surface, the heat signature provides clues about cooling activity deep in the material, resulting in a thorough assessment of the weld's integrity It

Application of the Guided Wave Technique to the Heat Exchanger Tube in NPP

International Nuclear Information System (INIS)

Yang, Dong Soon; Kim, Hyung Nam; Yoo, Hyun Joo

2005-01-01

The heat exchanger tube is examined by the method of eddy current test(ECT) to identify the integrity of the nuclear power plant. Because ECT probe is moved through the tube inside to identify flaws, the ECT probe should be exchanged periodically due to the wear of probe surface in order to remove the noise form the ECT signal. Moreover, it is impossible to examine the tube by ECT method because the ECT probe can not move through the inside due to the deformation such as dent. Recently, the theory of guided wave was established and the equipment applying the theory has been actively developed so as to overcome the limitation of ECT method for the tube inspection of heater exchanger in nuclear power plant. The object of this study is to know the application of the guided wave technique to heat exchanger tube in NPP

CANDU fuel sheath integrity and oxide layer thickness determination by Eddy current technique

International Nuclear Information System (INIS)

Gheorghe, Gabriela; Man, Ion; Parvan, Marcel; Valeca, Serban

2010-01-01

This paper presents results concerning the integrity assessment of the fuel elements cladding and measurements of the oxide layer on sheaths, using the eddy current technique. Flaw detection using eddy current provides information about the integrity of fuel element sheath or presence of defects in the sheath produced by irradiation. The control equipment consists of a flaw detector with eddy currents, operable in the frequency range 10 Hz to 10 MHz, and a differential probe. The calibration of the flaw detector is done using artificial defects (longitudinal, transversal, external and internal notches, bored and unbored holes) obtained on Zircaloy-4 tubes identical to those out of which the sheath of the CANDU fuel element is manufactured (having a diameter of 13.08 mm and a wall thickness of 0.4 mm). When analyzing the behavior of the fuel elements' cladding facing the corrosion is important to know the thickness of the zirconium oxide layer. The calibration of the device measuring the thickness of the oxide layer is done using a Zircaloy-4 tube identical to that which the cladding of the CANDU fuel element is manufactured of, and calibration foils, as well. (authors)

Statistical flaw detection: Application to flaws below curved surfaces

International Nuclear Information System (INIS)

Elsley, R.K.; Fertig, K.W.; Linebarger, R.S.; Richardson, J.M.

1984-01-01

This chapter presents a practical approach to the optimum detection of flaws in the presence of noise signals. A decision theoretic approach is used to derive a detection algorithm which is adapted to the noise environment in which a particular measurement is being made. An automatic procedure for characterizing the noises and developing the optimum detection algorithm is presented. The proposed method makes use of an explicit knowledge of the noise processes in order to design a flaw detection algorithm which optimally detects flaws in the presence of such noise. It is concluded that this approach will provide a number of advantages in practical testing situations, including the detection of smaller flaws, faster scanning due to the use of less highly focussed transducers, and less need for operator optimization of the measurement process. The described algorithms were implemented on the Digital Ultrasonic Instrument (DUI), which is a high speed all-digital instrument for performing sophisticated calculations on ultrasonic signals

Experimental verification on limit load estimation method for pipes with an arbitrary shaped circumferential surface flaw

International Nuclear Information System (INIS)

Li, Yinsheng; Hasegawa, Kunio; Miura, Naoki; Hoshino, Katsuaki

2010-01-01

When a flaw is detected in stainless steel pipes during in-service inspection, the limit load criterion given in the codes such as JSME Rules on Fitness-for-Service for Nuclear Power Plants or ASME Boiler and Pressure Vessel Code Section XI can be applied to evaluate the integrity of the pipe. However, in these codes, the limit load criterion is only provided for pipes containing a flaw with uniform depth, although many flaws with complicated shape such as stress corrosion cracking have been actually detected in pipes. In order to evaluate the integrity of the flawed pipes for general case, a limit load estimation method has been proposed by authors considering a circumferential surface flaw with arbitrary shape. The plastic collapse bending moment and corresponding stress are obtained by dividing the surface flaw into several segmented sub-flaws. In this paper, the proposed method was verified by comparing with experimental results. Four-point bending experiments were carried out for full scale stainless steel pipes with a symmetrical or non-symmetrical circumferential flaw. Estimated failure bending moments by the proposed method were found to be in good agreement with the experimental results, and the proposed method was confirmed to be effective for evaluating bending failure of pipes with flaw. (author)

Integrity Analysis of Damaged Steam Generator Tubes

International Nuclear Information System (INIS)

Stanic, D.

1998-01-01

Variety of degradation mechanisms affecting steam generator tubes makes steam generators as one of the critical components in the nuclear power plants. Depending of their nature, degradation mechanisms cause different types of damages. It requires performance of extensive integrity analysis in order to access various conditions of crack behavior under operating and accidental conditions. Development and application of advanced eddy current techniques for steam generator examination provide good characterization of found damages. Damage characteristics (shape, orientation and dimensions) may be defined and used for further evaluation of damage influence on tube integrity. In comparison with experimental and analytical methods, numerical methods are also efficient tools for integrity assessment. Application of finite element methods provides relatively simple modeling of different type of damages and simulation of various operating conditions. The stress and strain analysis may be performed for elastic and elasto-plastic state with good ability for visual presentation of results. Furthermore, the fracture mechanics parameters may be calculated. Results obtained by numerical analysis supplemented with experimental results are the base for definition of alternative plugging criteria which may significantly reduce the number of plugged tubes. (author)

Evaluation of the eddy-current method of inspecting steam generator tubing

International Nuclear Information System (INIS)

Flora, J.H.; Brown, S.D.; Weeks, J.R.

1976-01-01

The objective of this project has been to evaluate the eddy-current method of inspecting steam generator tubing by conducting a series of laboratory experiments with conventional eddy-current equipment. The experiments have involved obtaining eddy-current measurements on samples of 7/8-inch OD Inconel-600 tubing provided by the Westinghouse Nuclear Energy Systems Division. A variety of machined defects and some chemically induced flaws, such as stress corrosion cracks were fabricated in the tubing. Statistical evaluation of the data was employed to estimate the error encountered in measuring corrosion defects of various depths. It appears that the eddy-current technique can provide a reasonable measure of defect depth under certain conditions. On the other hand, the evaluation indicates that it is difficult to determine the depth of certain types of flaws with reliability and precision. Furthermore, although some defects as shallow as 10 percent of the tube wall could be detected, it was not possible to detect other types of flaws that were less than 40 percent deep even when the tube supports were not near the defects. The difficulty in detecting small volume flaws is attributed to low signal-to-noise ratio. Noise is a result of unwanted signals from test variables, such as wobble and variations in tube properties. The error in measurement of certain types of larger defects is associatedin part with test variables and also with the effects that the geometry of the defect has on the eddy-current signal patterns. The distortions in signal patterns caused by gradual wastage type defects and the poor reproducibility of signal patterns obtained from notches that represent stress corrosion cracks are described. Some developments that will rectify these detection and depth measurement problems are discussed

Demonstration for the Applicability of the EPRI ETSS on the SG Tube Wear Defects Formed at the Tube Support Structure

International Nuclear Information System (INIS)

Shin, Ki Seok; Cheon, Keun Young; Nam, Min Woo; Min, Kyong Mahn

2013-01-01

In this paper, the authorized EPRI ETSS 27906.2 applied to the detection of tapered wear volumetric indications and depth sizing within the free span area, loose part not present was reviewed and applied to the site SG tubes for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation. The experiment to demonstrate the applicability of EPRI ETSS was performed by the employment of the newly prepared STD tube and resulted in ensuring the effectiveness and equivalency of the EPRI ETSS as well. The authorized EPRI ETSS 27906.2 for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation was reviewed and applied to the site SG tubes. The testing results were reviewed with the influences of SG tube material and the support structure. The impact of the tube materials was insignificant and that of the tube support structure showed somewhat conservative results. The testing resulted in successful demonstration of applicability of the EPRI ETSS on the SG tube wear defects at the tube support. One of the major flaw mechanisms detected in the currently operating domestic OPR-1000 pressurized water reactors(PWR's) steam generator(SG) tubes is wear defect. In general, wear defect has been constantly detected in the upper tube bundle imposed to the flow induced vibration interaction between tube and its support structure, and the quantity of the affected tubes has also shown the tendency to increase as plant operation life is added. In order to take appropriate measures and maintain the structural integrity for the SG tubes, wear defect is currently categorized as active damage mechanism and the tubes containing 40% or greater wear depth of the nominal tube wall thickness shall be plugged per SGMP(SG Management Program) Recently, a fairly large amplitude of wear defects on the Batwing(BW), one of the upper tube support structures in the SG tubes

Demonstration for the Applicability of the EPRI ETSS on the SG Tube Wear Defects Formed at the Tube Support Structure

Energy Technology Data Exchange (ETDEWEB)

Shin, Ki Seok; Cheon, Keun Young; Nam, Min Woo [Korea Hydro and Nuclear Power Co. Ltd, Daejeon (Korea, Republic of); Min, Kyong Mahn [Universal Monitoring and Inspection Inc., Daejeon (Korea, Republic of)

2013-10-15

In this paper, the authorized EPRI ETSS 27906.2 applied to the detection of tapered wear volumetric indications and depth sizing within the free span area, loose part not present was reviewed and applied to the site SG tubes for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation. The experiment to demonstrate the applicability of EPRI ETSS was performed by the employment of the newly prepared STD tube and resulted in ensuring the effectiveness and equivalency of the EPRI ETSS as well. The authorized EPRI ETSS 27906.2 for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation was reviewed and applied to the site SG tubes. The testing results were reviewed with the influences of SG tube material and the support structure. The impact of the tube materials was insignificant and that of the tube support structure showed somewhat conservative results. The testing resulted in successful demonstration of applicability of the EPRI ETSS on the SG tube wear defects at the tube support. One of the major flaw mechanisms detected in the currently operating domestic OPR-1000 pressurized water reactors(PWR's) steam generator(SG) tubes is wear defect. In general, wear defect has been constantly detected in the upper tube bundle imposed to the flow induced vibration interaction between tube and its support structure, and the quantity of the affected tubes has also shown the tendency to increase as plant operation life is added. In order to take appropriate measures and maintain the structural integrity for the SG tubes, wear defect is currently categorized as active damage mechanism and the tubes containing 40% or greater wear depth of the nominal tube wall thickness shall be plugged per SGMP(SG Management Program) Recently, a fairly large amplitude of wear defects on the Batwing(BW), one of the upper tube support structures in the SG

Flaw evaluation methodology for class 2, 3 components in light water reactors

International Nuclear Information System (INIS)

Miura, Naoki; Kashima, Koichi; Miyazaki, Katsumasa; Hasegawa, Kunio; Oritani, Naohiko

2006-01-01

It is quite important to validate the structural integrity of operating plant components as aged LWR plants are gradually increasing in Japan. The rules on fitness-for-service for nuclear power plants constituted by the JSME provides flaw evaluation methodology. They are mainly focused on Class 1 components, while flaw evaluation criteria for Class 2, 3 components are not consolidated. As such, they also required from the viewpoints of in-service inspection request, reduction of operating cost and systematization of consistent code/standard. In this study, basic concept of flaw evaluation for Class 2, 3 piping was considered, and it is concluded that the same evaluation procedure as Class 1 piping in the current rules is applicable. Some technical issues on practical flaw evaluation for Class 2, 3 piping were listed up, and a countermeasure for each issue was devised. Especially, both allowable flaw sizes in acceptance standards and critical flaw sizes in acceptance criteria have to be determined in consideration of degraded fracture toughness. (author)

Automatic integrated testing bench for tubes in translation

International Nuclear Information System (INIS)

Dufayet, J.P.; Perdijon, J.

1976-01-01

All the nondestructive tests required for receiving the cladding tubes intended for fast nuclear reactor are integrated on this bench: quality control by eddy currents and ultra-sounds, thickness and (inner and outer) diameter measurement. The linear displacement of the tube allows very high rates to be attained [fr

Development of the double-wall-tube steam generator. Evaluation of inner tube leak detection system

International Nuclear Information System (INIS)

Teraoku, Takuji; Kisohara, Naoyuki

1995-01-01

A double-wall-tube steam generator (DWT-SG) is considered to have possibility of eliminating a secondary heat transport system to realize a reliable and simplified FBR plant. Thus, basic tests for inner/outer tube leak detection and prototypical leak tests by use of the 1MWt DWT-SG model have been performed to evaluate the feasibility of DWT-SG. Their results demonstrated that the inner leak detection system can definitely detect a steam leak from an inner tube flaw. Analyses of the inner tube leak and detection behavior obtained in the 1MWt DWT-SG test enabled to estimate the performance of the inner tube detection system of the commercial DWT-SG system. (author)

WWER steam generator tube structural and leakage integrity

International Nuclear Information System (INIS)

Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

1998-01-01

The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

Testing external surface of fuel element tubes for power nuclear reactors

International Nuclear Information System (INIS)

Naugol'nykh, O.G.; Nelyubin, Yu.V.

1987-01-01

Optical methods are regarded perspective for discovery and detection of flaws of external surfaces of fuel element tubes. The TV method has highest information content among them. Two mock-ups of facilities based on the TV method using a ''dissector'' type TV device and a TV tube with charge accumulation (vidikon) have been developed. It is concluded that complex testing - combination of ultrasonic, photoelectric and TV methods in a facility is necessary for discovery and analysis of the whole variety of flaws, though sensitivity of the TV method is enough for disclosure of all the main defects

Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

International Nuclear Information System (INIS)

Majumdar, S.

1996-09-01

Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

International Nuclear Information System (INIS)

Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

1996-01-01

The rules that are currently under application to verify the acceptance of flaws in nuclear components rely on deterministic criteria supposed to ensure the safe operating of plants. The interest of having a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation to reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R6 procedure for assessing the integrity of the structure. (authors). 5 refs., 5 figs

Probabilistic calibration of safety coefficients for flawed components in nuclear engineering

International Nuclear Information System (INIS)

Ardillon, E.; Pitner, P.; Barthelet, B.; Remond, A.

1995-01-01

The current rules applied to verify the flaws acceptance in nuclear components rely on deterministic criteria supposed to ensure the plant safe operation. The interest in have a precise and reliable method to evaluate the safety margins and the integrity of components led Electricite de France to launch an approach to link directly safety coefficients with safety levels. This paper presents a probabilistic methodology to calibrate safety coefficients in relation do reliability target values. The proposed calibration procedure applies to the case of a ferritic flawed pipe using the R 6 procedure for assessing the structure integrity. (author). 5 refs., 5 figs., 1 tab

Modeling the X-ray Process, and X-ray Flaw Size Parameter for POD Studies

Science.gov (United States)

Koshti, Ajay M.

2014-01-01

Nondestructive evaluation (NDE) method reliability can be determined by a statistical flaw detection study called probability of detection (POD) study. In many instances, the NDE flaw detectability is given as a flaw size such as crack length. The flaw is either a crack or behaving like a crack in terms of affecting the structural integrity of the material. An alternate approach is to use a more complex flaw size parameter. The X-ray flaw size parameter, given here, takes into account many setup and geometric factors. The flaw size parameter relates to X-ray image contrast and is intended to have a monotonic correlation with the POD. Some factors such as set-up parameters, including X-ray energy, exposure, detector sensitivity, and material type that are not accounted for in the flaw size parameter may be accounted for in the technique calibration and controlled to meet certain quality requirements. The proposed flaw size parameter and the computer application described here give an alternate approach to conduct the POD studies. Results of the POD study can be applied to reliably detect small flaws through better assessment of effect of interaction between various geometric parameters on the flaw detectability. Moreover, a contrast simulation algorithm for a simple part-source-detector geometry using calibration data is also provided for the POD estimation.

Structural integrity assessment of steam generator tubes deteriorated through primary water stress corrosion cracking in transition region of tube expansion

International Nuclear Information System (INIS)

Silveira, Helvecio Carlos Klinke da

2002-01-01

In PWR plants, steam generator tube degradation has been one of the most important economical concerns, besides causing operational safety problems. In this work, a survey of steam generator tube degradation modes is done. Degradation mechanisms and influence factors are introduced and discussed. The importance of stress corrosion cracking, especially in transition region of tube expansion zone, is underlined. The actual steam generator tube plugging criteria are conservative. Proposed alternative criteria are introduced and discussed. Distinction is done to structural integrity assessment of defective tubes. Real data of tube defect indications of axial cracks in expansion transition zone due to primary water stress corrosion cracking are used in analysis. Results allow discussing application aspects of deterministic and probabilistic criteria on structural integrity assessment of tubes with defect indications. Applied models are specifics, but the application of concept may be extended to other steam generator tube degradation modes. (author)

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D.R.; Majumdar, S.; Kupperman, D.S.; Bakhtiari, S.; Shack, W.J.

2002-01-01

Industry efforts have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, stress corrosions cracking (SCC) and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding, there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by 'plug or repair on detection' because current NDE techniques for characterization of flaws and the knowledge of SCC crack growth rates are not accurate enough to permit continued operation. Replacement generators with improved designs and materials have performed well to date, but previous experience with the appearance of some types of SCC in Alloy 600 after 10 years or more of operation and laboratory results suggest additional understanding of corrosion performance of these materials is needed. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators. (author)

Steam generator tube integrity requirements and operating experience in the United States

International Nuclear Information System (INIS)

Karwoski, K.J.

2009-01-01

Steam generator tube integrity is important to the safe operation of pressurized-water reactors. For ensuring tube integrity, the U.S. Nuclear Regulatory Commission uses a regulatory framework that is largely performance based. This performance-based framework is supplemented with some prescriptive requirements. The framework recognizes that there are three combinations of tube materials and heat treatments currently used in the United States and that the operating experience depends, in part, on the type of material used. This paper summarizes the regulatory framework for ensuring steam generator tube integrity, it highlights the current status of steam generators, and it highlights some of the steam generator issues and challenges that exist in the United States. (author)

Evaluating empirical/analytical techniques to predict structural integrity of pipe containing surface flaws

International Nuclear Information System (INIS)

Reuter, W.G.; Server, W.L.

1982-01-01

Data from flat-plate specimens containing either triangular-, ellipsoidal- or rectangular-shaped surface flaws were evaluated by several potential analytical techniques. These techniques were modified as needed to predict conditions for initiation of subcritical crack growth, for the defect to penetrate the 6.4 mm (0.25 in.) wall thickness, and for instability (plastic or unstable). The modified analytical techniques developed from the plate specimens were then used to make predictions which are compared with test results obtained from pipe specimens containing triangular-shaped surface flaws

Variable flaw shape analysis for a reactor vessel under pressurized thermal shock loading

International Nuclear Information System (INIS)

Yang, C.Y.; Bamford, W.H.

1984-01-01

A study has been conducted to characterize the response of semi-elliptic surface flaws to thermal shock conditions which can result from safety injection actuation in nuclear reactor vessels. A methodology was developed to predict the behavior of a flaw during sample pressurized thermal shock events. The effects of a number of key variables on the flaw propagation were studied, including fracture toughness of the material and its gradient through the thickness, irradiation effects, effects of warm prestressing, and effects of the stainless steel cladding. The results of these studies show that under thermal shock loading conditions the flaw always tends to elongate along the vessel inside surface from the initial aspect ratio. However, the flaw shape always remains finite rather than becoming continuously long, as has often been assumed in earlier analyses. The final shape and size of the flaws were found to be rather strongly dependent on the effects of warm prestressing and the distribution of neutron flux. The improved methodology results in a more accurate and more realistic treatment of flaw shape changes during thermal shock events and provides the potential for quantifying additional margins for reactor vessel integrity analyses

Development of an intelligent system for ultrasonic flaw classification in weldments

International Nuclear Information System (INIS)

Song, Sung-Jin; Kim, Hak-Joon; Cho, Hyeon

2002-01-01

Even though ultrasonic pattern recognition is considered as the most effective and promising approach to flaw classification in weldments, its application to the realistic field inspection is still very limited due to the crucial barriers in cost, time and reliability. To reduce such barriers, previously we have proposed an intelligent system approach that consisted of the following four ingredients: (1) a PC-based ultrasonic testing (PC-UT) system; (2) an effective invariant ultrasonic flaw classification algorithm; (3) an intelligent flaw classification software; and (4) a database with abundant experimental flaw signals. In the present work, for performing the ultrasonic flaw classification in weldments in a real-time fashion in many real word situations, we develop an intelligent system, which is called the 'Intelligent Ultrasonic Evaluation System (IUES)' by the integration of the above four ingredients into a single, unified system. In addition, for the improvement of classification accuracy of flaws, especially slag inclusions, we expand the feature set by adding new informative features, and demonstrate the enhanced performance of the IUES with flaw signals in the database constructed previously. And then, to take care of the increased redundancy in the feature set due to the addition of features, we also propose two efficient schemes for feature selection: the forward selection with trial and error, and the forward selection with criteria of the error probability and the linear correlation coefficients of individual features

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

International Nuclear Information System (INIS)

Upadhyaya, Belle R.; Wesley Hines, J.

2004-01-01

The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers.

Energy Technology Data Exchange (ETDEWEB)

Belle R. Upadhyaya; J. Wesley Hines

2004-09-27

The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

Flaw Detection using Electromagneticc Acourstic Resonance for a Tube of NPP

Energy Technology Data Exchange (ETDEWEB)

Lee, Dong Min; Kim, Yong Kwon; Choi, Sang Hoon [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

We present preliminary experimental verification of the proposed method and show that the defects on steel tubes can be detected based on changes in resonant frequencies. The result of this study showed that the electromagnetic acoustic resonance technique can detect notch defects on the tube and the changes in resonant frequencies are approximately proportional to the width of defect.

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers, Volumes 1, 2

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle R. [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Lu, Baofu [Univ. of Tennessee, Knoxville, TN (United States)

2005-06-03

wavelet transforms and image processing techniques for isolating flaw types. Development and implementation of a new nonlinear and non-stationary signal processing method, called the Hilbert-Huang transform (HHT), for flaw detection and location. This is a more robust and adaptive approach compared to the wavelet transform.Implementation of a moving-window technique in the time domain for detecting and quantifying flaw types in tubular structures. A window zooming technique was also developed for flaw location in tubes. Theoretical study of elastic wave propagation (longitudinal and shear waves) in metallic flat plates and tubing with and without flaws. Simulation of the Lamb wave propagation using the finite-element code ABAQUS. This enabled the verification of the experimental results. The research tasks included both analytical research and experimental studies. The experimental results helped to enhance the robustness of fault monitoring methods and to provide a systematic verification of the analytical results. The results of this research were disseminated in scientific meetings. The journal manuscript titled, "Structural Integrity Monitoring of Steam generator Tubing Using Transient Acoustic Signal Analysis," was published in IEEE Trasactions on Nuclear Science, Vol. 52, No. 1, February 2005. The new findings of this research have potential applications in aerospace and civil structures. The report contains a complete bibliography that was developed during the course of the project.

Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

Methodology of structural integrity assesment of CANDU-6 NPP fuel channels

International Nuclear Information System (INIS)

Radu, Vasile

2004-01-01

The paper describes the methodology of assessing the structural integrity in the CANDU-6 fuel channels making use of alternative methods of evaluation. An evaluation of the structural integrity of a CANDU-6 pressure tube made of Zr-2,5%Nb presenting both sharp and blunted defects is done. These analyses are made in compliance with the Canadian guide 'Pressure Tubes Fitness-for-service', and other two Recognizing procedures internationally adopted: the British procedure R6/Rev.4 and the American procedure API 579. Previously, the data base containing the data on materials property as well as the heat and dynamical loads in normal operation in CANDU-6 pressure tubes was established. Obtaining complete diagrams for structural integrity of pressure tubes with sharp and blunted defects in conditions of normal operation and long-term irradiation is the next step of the methodology to be developed. Modelling sharp and blunted defects on the inner side of pressure tubes, which can occur in normal reactor operating conditions is achieved by means of capabilities of pre-processing of two finite element analysis codes namely FEA-Crack and FEA-Flaw. The second part of the work deals with the analyses by finite element method of the fracture mechanics by means of the FEA-Crack code and with the evaluation of sharp and blunted defects by FAD diagrams in compliance with the British procedure R6/Rev.4. For typical models of blunted defects and thermo-mechanical loads specific to normal operation finite element analyses by FEA-Flaw codes were performed. Then FAD-iDHC diagrams were constructed to evaluate the initiation of the slow cracking under hydrogen/deuterium absorption, the known phenomenon of delayed hydride cracking (DHC)

Finite-element analysis of flawed and unflawed pipe tests

International Nuclear Information System (INIS)

James, R.J.; Nickell, R.E.; Sullaway, M.F.

1989-12-01

Contemporary versions of the general purpose, nonlinear finite element program ABAQUS have been used in structural response verification exercises on flawed and unflawed austenitic stainless steel and ferritic steel piping. Among the topics examined, through comparison between ABAQUS calculations and test results, were: (1) the effect of using variations in the stress-strain relationship from the test article material on the calculated response; (2) the convergence properties of various finite element representations of the pipe geometry, using shell, beam and continuum models; (3) the effect of test system compliance; and (4) the validity of ABAQUS J-integral routines for flawed pipe evaluations. The study was culminated by the development and demonstration of a ''macroelement'' representation for the flawed pipe section. The macroelement can be inserted into an existing piping system model, in order to accurately treat the crack-opening and crack-closing static and dynamic response. 11 refs., 20 figs., 1 tab

Eddy current test of fin tubes for a heat exchanger

International Nuclear Information System (INIS)

KIm, Young Joo; Lee, Se Kyung; Chung, Min Hwa

1992-01-01

Eddy current probes were designed for the test of fin tubes. Fin tubes, often used for heat exchangers, have uneven outer and inner surfaces to enhance the heat emission. The surface roughness make it difficult to detect flaws employing eddy current test(ECT). In order to overcome the difficulties we performed two types of works, one is the delopment of ECT probes, and the other is the signal processing including fast Fourier transform and digital filtering. In the development of ECT probes, we adopted empirical design method. Our ECT probes for fin tubes are inside diameter type. And we are specially concerned about geometric features such as the widths of the coils composing an ECT probe. We fabricated four probes with various coil widths. Eddy current test was performed using those ECT probes on specimens with artificial flaws. After analyzing the output signals, we found that, in order for the effective testing, the width of a coil should be determined considering the pitch of the fins of a tube. And we also learned that the frequency filtering could improve the s/n ratio.

Determination of Flaw Size from Thermographic Data

Science.gov (United States)

Winfree, William P.; Howell, Patricia A.; Zalameda, Joseph N.

2014-01-01

Conventional methods for reducing the pulsed thermographic responses of delaminations tend to overestimate the size of the flaw. Since the heat diffuses in the plane parallel to the surface, the resulting temperature profile over the flaw is larger than the flaw. A variational method is presented for reducing the thermographic data to produce an estimated size for the flaw that is much closer to the true size of the flaw. The size is determined from the spatial thermal response of the exterior surface above the flaw and a constraint on the length of the contour surrounding the flaw. The technique is applied to experimental data acquired on a flat bottom hole composite specimen.

Steam generator tube integrity program. Phase I report

International Nuclear Information System (INIS)

Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

1979-09-01

The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

Investigation of the radiation leakage from X ray flaw detectors and the improvement measures for the unqualified products

International Nuclear Information System (INIS)

Li Yiachun; Wu Yi; Pang Hu; Bai Bin

1997-01-01

The authors introduce investigation methods and results for radiation leakage from X ray flaw detectors, which are used in Beijing area. Total 21 sets of flaw detectors made in 8 factories in Beijing, Shanghai etc. have been tested, of which 16 sets made in Beijing, Dandong and Japan are gas cooling flaw detectors, and rest 5 sets made in Shanghai and Germany are water or oil cooling detectors. The air Kerma rate of leakage radiation at 1 m from the X ray tube target were measured by Type FJ-347A X, γ dosimeter. It can be seen from the results that, compared with the trade standard ZBY315-83, 5 sets of water or oil cooling flaw detectors are all qualified. However, only two sets of gas cooling detectors are qualified, and the radiation leakage of another 14 sets are over the values specified in the standard. The reason is analyzed, and some advices about the measures of improving radiation protection structure design and production technology for the unqualified products have been proposed

Experimental study of flow friction characteristics of integral pin-fin tubes

International Nuclear Information System (INIS)

Ding Ming; Yan Changqi; Sun Licheng

2007-01-01

Friction characteristics of integral pin-fin tubes, through which lubricating-oil flowed vertically, were studied experimentally. Effects of the pitch, the height of fins and the machining direction on friction coefficient were analyzed. The experimental results showed that the friction coefficient of the integral pin-fro tube was obviously lager than that of smooth tube. Compared with other influential factors, the effect of the height of fins was dominant. Because the three-dimensional pin fin could disturb and destroy the boundary layer, when the Reynolds Number reached 200-300, the friction coefficient curve began to bend, that was, a turning point was appeared in the friction coefficient curve. (authors)

Flaw shape reconstruction – an experimental approach

Directory of Open Access Journals (Sweden)

Marilena STANCULESCU

2009-05-01

Full Text Available Flaws can be classified as acceptable and unacceptable flaws. As a result of nondestructive testing, one takes de decision Admit/Reject regarding the tested product related to some acceptability criteria. In order to take the right decision, one should know the shape and the dimension of the flaw. On the other hand, the flaws considered to be acceptable, develop in time, such that they can become unacceptable. In this case, the knowledge of the shape and dimension of the flaw allows determining the product time life. For interior flaw shape reconstruction the best procedure is the use of difference static magnetic field. We have a stationary magnetic field problem, but we face the problem given by the nonlinear media. This paper presents the results of the experimental work for control specimen with and without flaw.

Time-dependent crack growth in steam generator tube leakage

International Nuclear Information System (INIS)

Chung, H.D.; Lee, J.H.; Park, Y.W.; Choi, Y.H.

2006-01-01

In general, cracks found in steam generator tubes have semi-elliptical shapes and it is assumed to be rectangular shape for conservatism after crack penetration. Hence, the leak and crack growth behavior has not been clearly understood after the elliptical crack penetrates the tube wall. Several experimental results performed by Argonne Nation Laboratory exhibited time-dependent crack growth behavior of rectangular flaws as well as trapezoidal flaws under constant pressure. The crack growth faster than expected was observed in both cases, which is likely attributed to time-dependent crack growth accompanied by fatigue sources such as the interaction between active jet and crack. The stress intensity factor, K 1 , is necessary for the prediction of the observed fatigue crack growth behavior. However, no K 1 solution is available for a trapezoidal flaw. The objective of this study is to develop the stress intensity factor which can be used for the fatigue analysis of a trapezoidal crack. To simplify the analysis, the crack is assumed to be a symmetric trapezoidal shape. A new K 1 formula for axial trapezoidal through-wall cracks was proposed based on the FEM results. (author)

Flaw identification using acoustic emission

International Nuclear Information System (INIS)

Woodward, B.; McDonald, N.R.

1975-01-01

Acoustic emission 'signatures' contain information about the fine structure of metallurgical source events and their interpretation may provide a means of assessing the severity of internal flaws as well as surface flaws. The ultimate aim of this research on signature analysis is to develop a real time non-destructive testing technique having the capability of flaw recognition as well as flaw location in nuclear reactor components and structures under stress. Thus the requisite, unlike that in most acoustic emission work to date, is for a technique which affords discrimination between acoustic emission from different types of flaws propagating simultaneously. The approach described here requires detailed analysis of the emission signatures in terms of a specific statistical parameter, energy spectral density. In order to realise the full inspection potential of acoustic emission monitoring data obtained from zirconium and steel testpieces have been correlated with metallurgical condition and mechanical behaviour, since the nature of emission signatures is strongly affected by the physical characteristics and internal structure of the material. (Auth.)

Fuel assembly and fuel cladding tube

International Nuclear Information System (INIS)

Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

1996-01-01

A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

Failure of fretted steam generator tubes under accident conditions

International Nuclear Information System (INIS)

Forrest, C.F.

1996-10-01

Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which would result in high cross-flow velocities. Fourteen specimen tubes were tested, each having one or two types of defect machined into the surface simulating fretting-wear type scars found in some operating steam generators. The tubes were tested at flow velocities sufficient to induce high fluid elastic-type vibrations. Seven of the tubes failed near the thinnest section of the defects during the one-hour tests, due to impacting and/or rubbing between the tube and the support. Strain gauges, displacement transducers, force gauges and an accelerometer were used on the target tube and/or the tube immediately downstream of it to measure their vibrational characteristics

Assembly and method for testing the integrity of stuffing tubes

Energy Technology Data Exchange (ETDEWEB)

Morrison, E.F.

1996-12-31

A stuffing tube integrity checking assembly includes first and second annular seals, with each seal adapted to be positioned about a stuffing tube penetration component. An annular inflation bladder is provided, the bladder having a slot extending longitudinally there along and including a separator for sealing the slot. A first valve is in fluid communication with the bladder for introducing pressurized fluid to the space defined by the bladder when mounted about the tube. First and second releasible clamps are provided. Each clamp assembly is positioned about the bladder for securing the bladder to one of the seals for thereby establishing a fluid-tight chamber about the tube.

Failures of fine tubes of steam generators and the essential defects

International Nuclear Information System (INIS)

Kawano, Shinji; Ebisawa, Toru; Sato, Susumu.

1976-01-01

Light water reactors were sold to Japan as their economy and safety have been established, but the average availability of 11 reactors in Japan during 7 year operation is only 53%, and it is being proved that there are questions in the safety and economy. In this report, the failures of fine tubes of steam generators are discussed from the standpoint of the corrosion of materials. First, the functions and construction of the fine tubes of steam generators in PWRs are explained. The failures of the fine tubes of steam generators became frequent since the beginning of 1970s as large capacity nuclear power stations have started the operation. When the fine tubes are pierced with holes during operation and the radioactivity in primary coolant leaks into secondary coolant, it is detected with radioactivity monitors. In order to find out the broken tubes, eddy current flaw detectors are used, and the tubes on which flaws were detected we plugged by explosion welding. In these works, many manual operations are included, and the radiation exposure of workers and the difficulties in the operations are the problems. The cases of the tube failures in Japan and foreign countries, the causes and the countermeasures are described. Chemical corrosion, thermal stress cycle, shaving off due to eddy flow, and stress corrosion are the probable causes. The safety of steam generators is essentially in extremely poor state. The seriousness of the tube failures in steam generators is emphasized. (Kako, I.)

Structural and leakage integrity assessment of WWER steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Splichal, K.; Otruba, J. [Nuclear Research Inst., Rez (Switzerland)

1997-12-31

The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

Structural and leakage integrity assessment of WWER steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Splichal, K; Otruba, J [Nuclear Research Inst., Rez (Switzerland)

1998-12-31

The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction. 10 refs.

Structural and leakage integrity assessment of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Otruba, J.

1997-01-01

The integrity of heat exchange tubes may influence the life-time of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirement is to assure a very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evolution and heat exchange tubes plugging based on eddy current test inspection. The stress corrosion cracking and pitting are the main corrosion damage of WWER heat exchange tubes and are initiated from the outer surface. They are influenced by water chemistry, temperature and tube wall stress level. They take place under crevice corrosion condition and are indicated especially (1) under the tube support plates, where up to 90-95 % of defects detected by the ECT method occur, and (2) on free spans under tube deposit layers. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through-wall cracks, oriented above all in the axial direction

Ultrasonic inspection of steam generator tubing for cracks, wall thinning and cross-sectional deformation

International Nuclear Information System (INIS)

Meyer, P.A.; Carodiskey, T.J.

1988-01-01

Periodic inspection of steam generator tubing is an important consideration in the efficient operation of a power generating facility. Since the operating life of these generators is finite, failures will occur. Due to the chemistry of the environment, thermal cycling, and other factors, flaws may develop that can cause rapid deterioration of the tubing while the overall performance of the unit may appear normal. In earlier presentation, the authors presented an ultrasonic bore-side array transducer which can be used with a conventional flaw detector instrument for the location of circumferential crack type defects on the outside tube surface. since that time, much additional experience has been gained on the performance of these probes. Probe performance has been characterized using fatigue crack samples and these results are reviewed. Probes have also been developed having 16 elements for use in larger diameter (25 mm) tubes. The bore-side array concept has been expanded to normal incidence tube well inspection allowing simultaneous wall thickness and eccentricity measurement which is very useful in the assessment of tube wastage and deformation. Preliminary data obtained in this area is presented

Heat Exchanger Tube Inspection of Nuclear Power Plants using IRIS Technique

International Nuclear Information System (INIS)

Yoon, Byung Sik; Yang, Seung Han; Song, Seok Yoon; Kim, Yong Sik; Lee, Hee Jong

2005-01-01

Inspection of heat exchange tubing include steam generator of nuclear power plant mostly performed with eddy current method. Recently, various inspection technique is available such as remote field eddy current, flux leakage and ultrasonic methods. Each of these techniques has its merits and limitations. Electromagnetic techniques are very useful to locate areas of concern but sizing is hard because of the difficult interpretation of an electric signature. On the other hand, ultrasonic methods are very accurate in measuring wall loss damage, and are reliable for detecting cracks. Additionally ultrasound methods is not affected by support plates or tube sheets and variation of electrical conductivity or permeability. Ultrasound data is also easier to analyze since the data displayed is generally the remaining wall thickness. It should be emphasized that ultrasound is an important tool for sizing defects in tubing. In addition, it can be used in situations where eddy current or remote field eddy current is not reliable, or as a flaw assessment tool to supplement the electromagnetic data. The need to develop specialized ultrasonic tools for tubing inspection was necessary considering the limitations of electromagnetic techniques to some common inspection problems. These problems the sizing of wall loss in carbon steel tubes near the tube sheet or support plate, sizing internal erosion damage, and crack detection. This paper will present an IRIS(Internal Rotating Inspection System) ultrasonic tube inspection technique for heat exchanger tubing in nuclear power plant and verify inspection reliability for artificial flaw embedded to condenser tube

Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Determination of hydrogen concentration and blister characterization

International Nuclear Information System (INIS)

2009-03-01

Heavy water reactors (HWRs) comprise significant numbers of today's operating nuclear power plants, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes, are an important factor in ensuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Intercomparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the framework of the IAEA's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of IAEA's project on advanced technologies for HWRs. The objective of the CRP was to compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP participants investigated the capability of different techniques to detect and characterize flaws. During the second phase participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in zirconium alloys. The intention was to identify the most effective pressure tube inspection and diagnostic methods and to identify further development needs. The organizations which participated in phase 2 of this CRP are: - Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL), Chalk River Laboratories (CRL), Canada; - Bhabha Atomic Research Centre (BARC), India; - Korea Atomic Energy Research Institute (KAERI), Republic of Korea; - National Institute for Research and Development for Technical Physics (NIRDTP), Romania; - Nuclear Non-Destructive Testing Research and Services (NNDT), Romania. IAEA-TECDOC-1499

Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

International Nuclear Information System (INIS)

2007-12-01

Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

Applied research for profilometric testing of the state of interior surfaces in heat exchanger tubes

International Nuclear Information System (INIS)

Gyongyosi, Tiberiu; Panaitescu, Valeriu Nicolae

2009-01-01

Generally, the surface flaws identified at heat exchangers tubing are characteristic for the heat secondary systems, located on the external surfaces of the heat exchanger tubes and are mostly the results of the ageing phenomena in systems operation. The tests performed, with the impressing replicating device confirmed the applicability of the technique, functionality of the device and resulted in replicas on metal support, these being the hard copy of the negative of the test tube surface, allowing the profile measurement. The visual inspection of the replicas on the metallic support gives information about the surface geometry replicated, pointing out the marks, which belong to the same area under observation. The minimum and maximum values for the depth of the channel worked out in the inner test tube wall have been determined by profile graphic measurement on the replicas. The paper presents the structural and functional description of the experimental devices. The first results and some conclusions are also included. Two patent applications were submitted at State Office for Inventions and Trademarks (OSIM) covering the original data to protect royalty: 'The local pit flaws, scratches, incipient micro-cracks replicating device on inner cylindrical surfaces', under no. A/00299/17.04.2008 and 'The annular local flaw, incipient micro-cracks replicating device on inner cylindrical surface' under no. A/00300/17.04.2008

Depth analysis of mechanically machined flaws on steam generator tubings using multi-parameter algorithm

International Nuclear Information System (INIS)

Nam Gung, Chan; Lee, Yoon Sang; Hwang, Seong Sik; Kim, Hong Pyo

2004-01-01

The eddy current testing (ECT) is a nondestructive technique. It is used for evaluation of material's integrity, especially, steam generator (SG) tubing in nuclear plants, due to their rapid inspection, safe and easy operation. For depth measurement of defects, we prepared Electro Discharge Machined (EDM) notches that have several of defects and applied multi-parameter (MP) algorithm. It is a crack shape estimation program developed in Argonne National Laboratory (ANL). To evaluate the MP algorithm, we compared defect profile with fractography of the defects. In the following sections, we described the basic structure of a computer-aided data analysis algorithm used as means of more accurate and efficient processing of ECT data, and explained the specification of a standard calibration. Finally, we discussed the accuracy of estimated depth profile compared with conventional ECT method

Application of non-destructive liner thickness measurement technique for manufacturing and inspection process of zirconium lined cladding tube

International Nuclear Information System (INIS)

Nakazawa, Norio; Fukuda, Akihiro; Fujii, Noritsugu; Inoue, Koichi

1986-01-01

Recently, in order to meet the difference of electric power demand owing to electric power situation, large scale load following operation has become necessary. Therefore, the development of the cladding tubes which withstand power variation has been carried out, as the result, zirconium-lined zircaloy 2 cladding tubes have been developed. In order to reduce the sensitivity to stress corrosion cracking, these zirconium-lined cladding tubes require uniform liner thickness over the whole surface and whole length. Kobe Steel Ltd. developed the nondestructive liner thickness measuring technique based on ultrasonic flaw detection technique and eddy current flaw detection technique. These equipments were applied to the manufacturing and inspection processes of the zirconium-lined cladding tubes, and have demonstrated superiority in the control and assurance of the liner thickness of products. Zirconium-lined cladding tubes, the development of the measuring technique for guaranteeing the uniform liner thickness and the liner thickness control in the manufacturing and inspection processes are described. (Kako, I.)

Nondestructive evaluation of the QT on the SG tubes affected by chemical cleaning

Energy Technology Data Exchange (ETDEWEB)

Shin, Ki Seok Shin; Cheon, Keun Young; Kim, Wang Bae [Central Research Institute, Daejeon (Korea, Republic of); Min, Kyong Mahn [UMI, Daejeon (Korea, Republic of)

2012-10-15

The major mechanisms of flaws detected on the currently operating steam generator(SG) tubes are wear and stress corrosion cracking(SCC) defects. Wear defect has continuously occurred in the upper tube bundle imposed to the flow induced vibration at the interaction between tube and its support structure. Meanwhile, SCC has been formed by a variety of mixed mode, such as the corrosion susceptible material, residual stress and secondary side chemical environment of the SG tubes. Recently, corrosion related defects were detected in the domestic OPR 1000 model SG tubes especially in the egg crate tube support plate(TSP), as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). Therefore, the need to take corrective measures against the corrosion defects is required and various studies have been conducted to clarify the main causes of the defects. In general, as a representing SG tube materials, Ni based alloy 600 tubes have been widely applied and also adversely shown weak properties on the corrosion cracking resistivity. According to the studies on the factors developing corrosion cracking, densely accumulated sludge pile on the secondary side of the SG tubes have been mainly attributed to the formation of the corrosion defects. Therefore, it is imperative to secure applicable and efficient sludge removal process. In this paper, the chemical cleaning processes to dissolve and remove the sludge, thus promote the integrity of the SG tubes were introduced and eddy current testing(ECT) results on the pre cracked SG tubes to determine the effectiveness of those processes were represented as well.

Flaws in Commercial Reading Materials.

Science.gov (United States)

Axelrod, Jerome

Three flaws found in commercial reading materials, such as workbooks and kits, are discussed in this paper, and examples of the flaws are taken from specific materials. The first problem noted is that illustrations frequently provide the information that the learner is supposed to supply through phonetic or structural analysis; the illustrations…

The detection of tightly closed flaws by nondestructive testing (NDT) methods. [fatigue crack formation in aluminum alloy test specimens

Science.gov (United States)

Rummel, W. D.; Rathke, R. A.; Todd, P. H., Jr.; Mullen, S. J.

1975-01-01

Liquid penetrant, ultrasonic, eddy current and X-radiographic techniques were optimized and applied to the evaluation of 2219-T87 aluminum alloy test specimens in integrally stiffened panel, and weld panel configurations. Fatigue cracks in integrally stiffened panels, lack-of-fusion in weld panels, and fatigue cracks in weld panels were the flaw types used for evaluation. A 2319 aluminum alloy weld filler rod was used for all welding to produce the test specimens. Forty seven integrally stiffened panels containing a total of 146 fatigue cracks, ninety three lack-of-penetration (LOP) specimens containing a total of 239 LOP flaws, and one-hundred seventeen welded specimens containing a total of 293 fatigue cracks were evaluated. Nondestructive test detection reliability enhancement was evaluated during separate inspection sequences in the specimens in the 'as-machined or as-welded', post etched and post proof loaded conditions. Results of the nondestructive test evaluations were compared to the actual flaw size obtained by measurement of the fracture specimens after completing all inspection sequences. Inspection data were then analyzed to provide a statistical basis for determining the flaw detection reliability.

An engineering approach for examining crack growth and stability in flawed structures

International Nuclear Information System (INIS)

Shih, C.F.; German, M.D.; Kumar, V.

1981-01-01

Progress made in two research programmes, sponsored by the Electric Power Research Institute (EPRI), to identify viable parameters for characterising crack initiation and continued extension are summarised. An engineering/design methodology, based on these parameters, for the assessment of crack growth and instability in engineering structures which are stressed beyond the regime of applicability of linear elastic fracture mechanics is developed. The ultimate goal in the development of such a methodology is to establish an improved basis for analysing the effect of flaws (postulated or detected) on the safety margins of pressure boundary components of light water-cooled type nuclear steam supply systems. The methodology can also be employed for structural integrity analyses of other engineering components. Extensive experimental and analytical investigations undertaken to evaluate potential criteria for crack initiation and growth and the selection of the final criteria for analysing crack growth and stability in flawed structures are summarised. The experimental and analytical results obtained to date suggest that parameters based on the J-integral and the crack tip opening displacement, delta, are the most promising. This is not surprising since, from a theoretical basis, the two approaches are similar if certain conditions are met. An engineering/design approach for the assessment of crack growth and instability in flawed structures is outlined. (author)

3D integrated HYDRA simulations of hohlraums including fill tubes

Science.gov (United States)

Marinak, M. M.; Milovich, J.; Hammel, B. A.; Macphee, A. G.; Smalyuk, V. A.; Kerbel, G. D.; Sepke, S.; Patel, M. V.

2017-10-01

Measurements of fill tube perturbations from hydro growth radiography (HGR) experiments on the National Ignition Facility show spoke perturbations in the ablator radiating from the base of the tube. These correspond to the shadow of the 10 μm diameter glass fill tube cast by hot spots at early time. We present 3D integrated HYDRA simulations of these experiments which include the fill tube. Meshing techniques are described which were employed to resolve the fill tube structure and associated perturbations in the simulations. We examine the extent to which the specific illumination geometry necessary to accommodate a backlighter in the HGR experiment contributes to the spoke pattern. Simulations presented include high resolution calculations run on the Trinity machine operated by the Alliance for Computing at Extreme Scale (ACES) partnership. This work was performed under the auspices of the Lawrence Livermore National Security, LLC, (LLNS) under Contract No. DE-AC52-07NA27344.

Improvement of ISI techniques by multi-frequency eddy current testing method for steam generator tube in PWR plant

International Nuclear Information System (INIS)

Endo, Takashi; Kamimura, Takeo; Nishihara, Masatoshi; Araki, Yasuo; Fukui, Shigetaka.

1982-05-01

Eddy current flaw detection techniques are applied to the in-service inspection (ISI) of steam generator tubes in pressurized water reactors (PWR) plant. To improve the reliability and operating efficiency of the plants, efforts are being made to develop eddy current testing methods of various kinds. Multi-frequency eddy current testing method, one of new method, has recently been applied to actual heat exchanger tubes, contributing to the improvement of the detectability and signal evaluation of the ISI. The outline of multi-frequency eddy current testing method and its effects on the improvement of flaw detecting and signal evaluation accuracy are described. (author)

Thermal-shock experiments with flawed clad cylinders

International Nuclear Information System (INIS)

Cheverton, R.D.; Bryson, J.W.; Alexander, D.J.

1989-01-01

The life expectancy of LWR pressure vessels is influenced by a reduction in fracture toughness that is the result of radiation damage. As the fracture toughness decreases, the probability of propagation of preexisting flaws (sharp, crack-like defects) in the wall of the vessel increases. The probability of propagation is also influenced by the type of loading condition and the type of flaws that might exist. A loading condition of particular concern is referred to as pressurized thermal shock (PTS), and a flaw of particular concern for PTS loading conditions is a shallow surface flaw. A sudden cooling (thermal shock) of the inner surface of the vessel results in relatively high tensile stresses and relatively low fracture toughness at the inner surface. In addition, the attenuation of the fast-neutron fluence also results in relatively low fracture toughness at the inner surface. Under some circumstances, this combination of high stress and low toughness at the inner surface makes it possible for very shallow surface flaws to propagate. The PTS issue has been under investigation for quite some time, but thus far possible beneficial effects, other than thermal resistance, of the cladding on the inner surface of the vessel have not been included in the analysis of flaw behavior. This document discusses this effect of cladding on surface flaws and crack propagation

Probabilistic analysis of flaw distribution on structure under cyclic load

International Nuclear Information System (INIS)

Kwak, Sang Log; Choi, Young Hwan; Kim, Hho Jung

2003-01-01

Flaw geometries, applied stress, and material properties are major input variables for the fracture mechanics analysis. Probabilistic approach can be applied for the consideration of uncertainties within these input variables. But probabilistic analysis requires many assumptions due to the lack of initial flaw distributions data. In this study correlations are examined between initial flaw distributions and in-service flaw distributions on structures under cyclic load. For the analysis, LEFM theories and Monte Carlo simulation are applied. Result shows that in-service flaw distributions are determined by initial flaw distributions rather than fatigue crack growth rate. So initial flaw distribution can be derived from in-service flaw distributions

Evaluation of flaws in carbon steel piping. Final report

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Gamble, R.M.; Mehta, H.S.; Yukawa, S.; Ranganath, S.

1986-10-01

The objective of this program was to develop flaw evaluation procedures and allowable flaw sizes for ferritic piping used in light water reactor (LWR) power generation facilities. The program results provide relevant ASME Code groups with the information necessary to define flaw evaluation procedures, allowable flaw sizes, and their associated bases for Section XI of the code. Because there are several possible flaw-related failure modes for ferritic piping over the LWR operating temperature range, three analysis methods were employed to develop the evaluation procedures. These include limit load analysis for plastic collapse, elastic plastic fracture mechanics (EPFM) analysis for ductile tearing, and linear elastic fracture mechanics (LEFM) analysis for non ductile crack extension. To ensure the appropriate analysis method is used in an evaluation, a step by step procedure also is provided to identify the relevant acceptance standard or procedure on a case by case basis. The tensile strength and toughness properties required to complete the flaw evaluation for any of the three analysis methods are included in the evaluation procedure. The flaw evaluation standards are provided in tabular form for the plastic collapse and ductile tearing modes, where the allowable part through flaw depth is defined as a function of load and flaw length. For non ductile crack extension, linear elastic fracture mechanics analysis methods, similar to those in Appendix A of Section XI, are defined. Evaluation flaw sizes and procedures are developed for both longitudinal and circumferential flaw orientations and normal/upset and emergency/faulted operating conditions. The tables are based on margins on load of 2.77 and 1.39 for circumferential flaws and 3.0 and 1.5 for longitudinal flaws for normal/upset and emergency/faulted conditions, respectively.

Evaluation of flaws in carbon steel piping. Final report

International Nuclear Information System (INIS)

Zahoor, A.; Gamble, R.M.; Mehta, H.S.; Yukawa, S.; Ranganath, S.

1986-10-01

The objective of this program was to develop flaw evaluation procedures and allowable flaw sizes for ferritic piping used in light water reactor (LWR) power generation facilities. The program results provide relevant ASME Code groups with the information necessary to define flaw evaluation procedures, allowable flaw sizes, and their associated bases for Section XI of the code. Because there are several possible flaw-related failure modes for ferritic piping over the LWR operating temperature range, three analysis methods were employed to develop the evaluation procedures. These include limit load analysis for plastic collapse, elastic plastic fracture mechanics (EPFM) analysis for ductile tearing, and linear elastic fracture mechanics (LEFM) analysis for non ductile crack extension. To ensure the appropriate analysis method is used in an evaluation, a step by step procedure also is provided to identify the relevant acceptance standard or procedure on a case by case basis. The tensile strength and toughness properties required to complete the flaw evaluation for any of the three analysis methods are included in the evaluation procedure. The flaw evaluation standards are provided in tabular form for the plastic collapse and ductile tearing modes, where the allowable part through flaw depth is defined as a function of load and flaw length. For non ductile crack extension, linear elastic fracture mechanics analysis methods, similar to those in Appendix A of Section XI, are defined. Evaluation flaw sizes and procedures are developed for both longitudinal and circumferential flaw orientations and normal/upset and emergency/faulted operating conditions. The tables are based on margins on load of 2.77 and 1.39 for circumferential flaws and 3.0 and 1.5 for longitudinal flaws for normal/upset and emergency/faulted conditions, respectively

The influence of long-range residual stress on plastic collapse of pressurised pipes with and without flaws

International Nuclear Information System (INIS)

Wu, Gui-Yi; Smith, David J.; Pavier, Martyn J.

2013-01-01

Structural integrity assessments of pressurised pipes include plastic collapse as a potential failure mode. This paper uses analytical and numerical models to explore the effect of the end conditions of the pipe on the collapse pressure. The pipe is open-ended and two bounding conditions are addressed: one where axial loading is applied to the ends of the pipe and the other where a fixed axial displacement is applied. The fixed axial displacement condition represents long-range or fit-up residual stress. It is common practice to treat long-range residual stress in the same way as axial loading, leading to the conclusion that such long-range residual stress reduces the collapse pressure. Pipes in a number of states are considered: pipes with no flaws, pipes with fully circumferential flaws and pipes with part circumferential flaws. The flaws consist of either a crack or a slot on the external surface of the pipe. For the axial load condition, the collapse pressure for a flawed pipe is reduced when higher magnitudes of tensile or compressive axial loads are applied. For the fixed displacement condition however, the magnitude of the displacement may have little or no effect on the collapse pressure. The results of the work indicate that substantially conservative assessments may be made of the collapse pressures of pipes containing flaws, when long-range residual stress is taken to be a form of axial loading. -- Highlights: • The effect of end conditions on the collapse pressure of a pipe has been explored. • Fixed displacement conditions represent long-range residual stress. • Long-range residual stress is commonly thought to contribute to plastic collapse. • We show long-range residual stress has no influence on collapse for flawed pipes. • It is therefore possible to reduce conservatism in structural integrity assessment

Supersonic flaw detection device for nozzle

International Nuclear Information System (INIS)

Hata, Moriki.

1996-01-01

In a supersonic flaw detection device to be attached to a body surface of a reactor pressure vessel for automatically detecting flaws of a welded portion of a horizontally connected nozzle by using supersonic waves, a running vehicle automatically running along a circumferential direction of the nozzle comprises a supersonic flaw detection means for detecting flaws of the welded portion of the nozzle by using supersonic waves, and an inclination angle sensor for detecting the inclination angle of the running vehicle relative to the central axis of the nozzle. The running distance of the vehicle running along the circumference of the nozzle, namely, the position of the running vehicle from a reference point of the nozzle can be detected accurately by dividing the distance around the nozzle by the inclination angle detected by the inclination angle sensor. Accordingly, disadvantages in the prior art, for example, that the detected values obtained by using an encoder are changed by slipping or idle running of the magnet wheels are eliminated, and accurate flaw detection can be conducted. In addition, an operation of visually adjusting the reference point for the device can be eliminated. An operator's exposure dose can be reduced. (N.H.)

Eddy-current inspection of ferromagnetic tubing using pulsed magnetic saturation

Energy Technology Data Exchange (ETDEWEB)

Dodd, C V; Deeds, W E

1986-07-01

A pulsed eddy-current system has been designed and developed for nondestructive evaluation of 2.25Cr-1Mo steam generator tubing from the bore side. Since the tubing is ferromagnetic, a large current pulse is sent through a driver coil to produce magnetic saturation all the way through the tube wall. A pickup coil produces an output pulse that is dependent upon the tube properties as well as the driving pulse. The output pulse heights at selected times are used as data that are computer-correlated with calibration data taken from machined standards. Performance data, circuit diagrams, and computer programs are given for the system, which has been demonstrated to detect small flaws located near the outside of a thick ferromagnetic tube.

Labor security in radiation flaw detection

International Nuclear Information System (INIS)

Margulis, U.Ya.; Chistov, E.D.; Partolin, O.F.; Pertsov, V.A.; Momzhiev, B.N.; Sprygaev, I.F.

1986-01-01

Problems of ensuring safe labour conditions in radiation flaw detection are considered. Methods for ionizing radiation protection are given calculating techniques for shielding flaw detectors and stationary structures are presented as well. Safe methods of nondestructive testing of items under field conditions, in a shop and special laboratories using gamma- and X-ray flaw detectors, betatrons, electron accelerators are described. Attention is paid to the principles of radiation factor stantardization as well as radiation monitoring. Analysis of accidents and recommendations on their prevention and liquidation of accidental consequences are given

Performance demonstration requirements for eddy current steam generator tube inspection

International Nuclear Information System (INIS)

Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

1992-10-01

This paper describes the methodology used for developing performance demonstration tests for steam generator tube eddy current (ET) inspection systems. The methodology is based on statistical design principles. Implementation of a performance demonstration test based on these design principles will help to ensure that field inspection systems have a high probability of detecting and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented. Probability of detection and flaw sizing tests are described

Risk assessment of severe accident-induced steam generator tube rupture

International Nuclear Information System (INIS)

1998-03-01

This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC's Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs

Probabilistic methodology for assessing steam generator tube inspection - Phase II: User's manual for CANTIA Version 1.1

International Nuclear Information System (INIS)

Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

1997-03-01

The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the user's manual is provided in this volume. The documentation and verification of the CANTIA code is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

Data analysis for steam generator tubing samples

International Nuclear Information System (INIS)

Dodd, C.V.

1996-07-01

The objective of the Improved Eddy-Current ISI for Steam Generators program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report provides a description of the application of advanced eddy-current neural network analysis methods for the detection and evaluation of common steam generator tubing flaws including axial and circumferential outer-diameter stress-corrosion cracking and intergranular attack. The report describes the training of the neural networks on tubing samples with known defects and the subsequent evaluation results for unknown samples. Evaluations were done in the presence of artifacts. Computer programs are given in the appendix

Midland reactor pressure vessel flaw distribution

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.; Rosinski, S.T.

1993-12-01

The results of laboratory nondestructive examination (NDE), and destructive cross-sectioning of selected weldment sections of the Midland reactor pressure vessel were analyzed per a previously developed methodology in order to develop a flaw distribution. The flaw distributions developed from the NDE results obtained by two different ultrasonic test (UT) inspections (Electric Power Research Institute NDE Center and Pacific Northwest Laboratories) were not statistically significantly different. However, the distribution developed from the NDE Center's (destructive) cross-sectioning-based data was found to be significantly different than those obtained through the UT inspections. A fracture mechanics-based comparison of the flaw distributions showed that the cross-sectioning-based data, conservatively interpreted (all defects considered as flaws), gave a significantly lower vessel failure probability when compared with the failure probability values obtained using the UT-based distributions. Given that the cross-sectioning data were reportedly biased toward larger, more significant-appearing (by UT) indications, it is concluded that the nondestructive examinations produced definitively conservative results. In addition to the Midland vessel inspection-related analyses, a set of twenty-seven numerical simulations, designed to provide a preliminary quantitative assessment of the accuracy of the flaw distribution method used here, were conducted. The calculations showed that, in more than half the cases, the analysis produced reasonably accurate predictions

Investigation of the integrity of u-bend tube bundles subjected to flow-induced vibrations

Energy Technology Data Exchange (ETDEWEB)

Hassan, M. [University of Guelph, Guelph, Ontario (Canada); Riznic, J. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2012-07-01

Maintaining the integrity of nuclear steam generator (SG) tubes in CANDU reactors is a major safety issue since they maintain the physical barrier between the primary and secondary coolants. The integrity of these tubes can be compromised due to flow-induced vibrations in the form of fatigue and fretting wear damage. Wear is a result of the tube impacting and sliding against its loose supports, and it becomes more severe as the tube/support clearance increases. The vibration is caused by fluid flow around these tubes through turbulence and fluidelastic instability mechanisms. Supports are installed to stiffen the structure and to ensure safe and stable operation. The U-bend region is the most critical part since it is subjected to high cross flow. Therefore, special attention is paid to properly supporting this region. However, in some situations, tube support plates (TSP) located on the straight part of the tube may deteriorate to the point where extremely large clearances, or even total wastage of the supports, may result. One possible cause for such a situation is corrosion and/or excessive fretting wear. This loss of TSP may affect the rate of wear in the U-bend portion of the tube due to the increased flexibility in this region. The integrity could be seriously breached as result of a potential support loss. This paper addresses the flow-induced vibrations (FIV) aspect, consequences, and suggested remedies for support degradation. This analysis will include fretting wear producing parameters, such as impact force and normal work rate. Turbulence and fluidelastic instability (FEI) are considered to be the main excitation mechanisms. The investigation is conducted through a numerical simulation of the full Ubend tube bundles including modelling the variable flow distribution, flow excitation, impact, and friction at the supports. (author)

The influence of finite-length flaw effects on PTS analyses

International Nuclear Information System (INIS)

Keeney-Walker, J.; Dickson, T.L.

1993-01-01

Current licensing issues within the nuclear industry dictate a need to investigate the effects of cladding on the extension of small finite-length cracks near the inside surface of a vessel. Because flaws having depths of the order of the combined clad and heat affected zone thickness dominate the frequency distribution of flaws, their initiation probabilities can govern calculated vessel failure probabilities. Current pressurized-thermal-shock (PTS) analysis computer programs recognize the influence of the inner-surface cladding layer in the heat transfer and stress analysis models, but assume the cladding fracture toughness is the same as that for the base material. The programs do not recognize the influence cladding may have in inhibiting crack initiation and propagation of shallow finite-length surface flaws. Limited experimental data and analyses indicate the cladding can inhibit the propagation of certain shallow flaws. This paper describes an analytical study which was carried out to determine (1) the minimum flaw depth for crack initiation under PTS loading for semicircular surface flaws in a clad reactor pressure vessel and (2) the impact, in terms of the conditional probability of vessel failure, of using a semicircular surface flaw as the initial flaw and assuming that the flaw cannot propagate in the cladding. The analytical results indicate that for initiation a much deeper critical crack depth is required for the finite-length flaw than for the infinite-length flaw, except for the least severe transient. The minimum flaw depths required for crack initiation from the finite-length flaw analyses were incorporated into a modified version of the OCA-P code. The modified code was applied to the analysis of selected PTS transients, and the results produced a substantial decrease in the conditional probability of failure. This initial study indicates a significant effect on probabilistic fracture analyses by incorporating finite-length flaw results

An interim report on shallow-flaw fracture technology development

International Nuclear Information System (INIS)

Pennell, W.E.; Bass, B.R.; Bryson, J.W.; McAfee, W.J.

1995-01-01

Shallow-flaw fracture technology is being developed for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVS) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) a strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness

Detection and mode identification of axial cracks in the steam generator tube of the nuclear power plant using ultrasonic guided wave

International Nuclear Information System (INIS)

Yoon, Byungsik; Yang, Seunghan; Lee, Heejong; Kim, Yongsik

2010-01-01

For those people who are involved in NDE, there is a growing concern regarding the significant traveling distance of a guided wave in a structure, which ensures the inspection of a large area of the structure from a single location. A significant number of studies on the guided wave have therefore been made to apply the foregoing to a nondestructive evaluation in many different industries and resulted in an increase in the efficiency of practical guided wave inspection. Unlike the previous studies based mainly on the detection of circumferential flaws, this study is focused on the axial flaw detection in the steam generator tubes of Korean standard nuclear power plants by generating the guided wave by changing frequency and selecting the applicable mode from the dispersion curve for the steam generator tube calculated in this study, where the dispersion-based short-time Fourier transform (D-STFT) algorithm is used to enhance mode identification. In conclusion, the L (0,1) mode at 2.25 MHz is found to be most sensitive in detecting axial flaws in a steam generator tube. (author)

Investigation of reliability of EC method for inspection of VVER steam generator tubes

International Nuclear Information System (INIS)

Corak, Z.

2004-01-01

Complete and accurate non-destructive examinations (NDE) data provides the basis for performing mitigating actions and corrective repairs. It is important that detection and characterization of flaws are done properly at an early stage. EPRI Document PWR Steam Generator Examination Guidelines recommends an approach that is intended to provide the following: Ensure accurate assessment of steam generator tube integrity; Extend the reliable, cost effective, operating life of the steam generators, and Maximize the availability of the unit. Steam Generator Eddy Current Data Analysis Performance Demonstration represents the culmination of the intense two-year industry effort in the development of a performance demonstration program for eddy current testing (ECT) of steam generator tubing. It is referred to as the Industry Database (IDB) and provides a capability for individual organizations to implement SG ECT performance demonstration programs in accordance with the requirements specified in Appendices G and H of the ISI Guidelines. The Appendix G of EPRI Document PWR Steam Generator Examination Guidelines specifies personnel training and qualification requirements for NDE personnel who analyze NDE data for PWR steam generator tubing. Its purpose is to insure a continuing uniform knowledge base and skill level for data analysis. The European methodology document is intended to provide a general framework for development of qualifications for the inspection of specific components to ensure they are developed in a consistent way throughout Europe while still allowing qualification to be tailored in detail to meet different nation requirements. In the European methodology document one will not find a detailed description of how the inspection of a specific component should be qualified. A recommended practice is a document produced by ENIQ to support the production of detailed qualification procedures by individual countries. VVER SG tubes are inspected by EC method but a

Reliably detectable flaw size for NDE methods that use calibration

Science.gov (United States)

Koshti, Ajay M.

2017-04-01

Probability of detection (POD) analysis is used in assessing reliably detectable flaw size in nondestructive evaluation (NDE). MIL-HDBK-1823 and associated mh18232 POD software gives most common methods of POD analysis. In this paper, POD analysis is applied to an NDE method, such as eddy current testing, where calibration is used. NDE calibration standards have known size artificial flaws such as electro-discharge machined (EDM) notches and flat bottom hole (FBH) reflectors which are used to set instrument sensitivity for detection of real flaws. Real flaws such as cracks and crack-like flaws are desired to be detected using these NDE methods. A reliably detectable crack size is required for safe life analysis of fracture critical parts. Therefore, it is important to correlate signal responses from real flaws with signal responses form artificial flaws used in calibration process to determine reliably detectable flaw size.

Program to develop acoustic emission: flaw relationship for inservice monitoring of nuclear pressure vessels. Progress report No. 1, July 1, 1976--February 1, 1977

International Nuclear Information System (INIS)

Hutton, P.H.; Schwenk, E.B.

1977-03-01

This is a laboratory research program to characterize acoustic emission (AE) from flaw growth and noise from innocuous sources in A533B Class 1 pressure vessel steel. The objectives are: characterize AE from a limited range of defects and material property conditions of concern to reactor pressure vessel integrity; characterize AE from innocuous sources (including defects); develop criteria for distinguishing significant flaws from innocuous sources; and develop an AE flaw damage model to serve as a basis for relating in-service AE to pressure vessel integrity. The purpose of the program is to build an experimental evaluation of the feasibility of detecting and analyzing flaw growth in reactor pressure boundaries by continuously monitoring for AE. A detailed program plan in the form of an analysis-before-test document has been prepared and approved

Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

International Nuclear Information System (INIS)

Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

2017-01-01

Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

Final report on development evaluation of Task Group 3 pressure tubes

International Nuclear Information System (INIS)

Fleck, R.G.; Price, E.G.; Cheadle, B.A.

1983-11-01

This report describes the production and evaluation of pressure tubes manufactured to the recommendations of Task Group 3 (TG3) of the Creep Engineering Design Plan. The Zr-2.5 wt percent Nb tubes were manufactured by modified production route to change their metallurgical structure and so reduce the in-service elongation rates. Three modified routes were investigated and a total of twenty-eight tubes produced. There were no difficulties in manufacture and the tubes satisfied the quality assurance and design specifications of reactor grade tubes. Metallurgical evaluation showed that the expected changes in microstructure had occurred but not to the extent anticipated. The TG3 tubes were found to have comparable properties to current tubes when tested for: tensile strength (irradiated and unirradiated); hydride cracking; stress to reorient hydrides; hydrogen diffusion; flaw tolerance; corrosion (irradiated and unirradiated); wear; rolled joint characteristics; irradiation creep and growth. Lower in-service elongation rates are expected for tubes produced by two of the modified routes

RID-41 gamma flaw detector

International Nuclear Information System (INIS)

Glebov, V.N.; Zubkov, V.S.; Majorov, A.N.; Murashev, A.I.; Firstov, V.G.; Yampol'skij, V.V.; Goncharov, V.I.; Sakhanov, A.S.

1978-01-01

The design is described and the main characteristics are given of a universal stationary hose-type gamma flow detector with a 60 Co source from 3O to 4g0 Ci for high-productive control of thick-walled products from steel and other materials. The principal units of the instrument are a radiation head, a control panel, and a charge-exchange container. The flaw detector may be used both in shield chambers and in shop or mounting conditions on complying with due requirements of radiation protection. The high activity of the source at relatively small dimensions of its active part ensures good detection of defects. The high radioscopy rate permits to use the flaw detector in conditions of increased background radiation, e.g. during routine repairs and inspections at nuclear power plants. The instrument may also be used in radiometric complexes, and produces a considerable economic effect. This flaw-detector corresponds to ISO and IAEA requirements and may be delivered for export

Rotating sensor technology for the inspection of steam generator tubing

International Nuclear Information System (INIS)

Glass, S.W.; Richards, T.A.

1986-01-01

A high-resolution profilometry system, has been developed to assess the dimensional condition of steam generator tubes and rapidly produce the data to evaluate the potential for developing in-service leaks. The probe has an electromechanical sensor in a rotating head. This technique has been demonstrated in the field at four U.S. plants and one plant owned by Electricite de France. The Indian Point-2 plant of Consolidated Edison has twice used this technology to save tubes that would have been plugged with the go-gauge criterion and identifying other high-risk candidates for plugging that might otherwise not have been removed from service. As an extension of the PROFIL-360 technology, a rotating eddy current system (EDDY-360) has also been developed. The system provides improved sensitivity, resolution, and characterization of small-volume flaws and complete circumferential coverage as compared to conventional (bobbin and 8 x 1) eddy current techniques. Enhanced eddy current data processing provides on-line data analysis and real-time imaging of detected flaws. (author)

Improved criteria for the repair of fabrication flaws

International Nuclear Information System (INIS)

Doctor, S.R.; Schuster, G.J.; Simonen, F.A.

2003-01-01

Section III of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for nuclear power plant components requires radiographic examinations (RT) of welds and requires repairs for RT indications that exceed code acceptable sizes. This paper describes research that has generated data on welding flaws, which indicated that the largest flaws occur in repaired welds. The fabrication flaws were detected in material removed from cancelled nuclear power plants using high sensitivity Nondestructive Examination (NDE) and validated by complementary NDE and destructive testing. Evidence suggests that repairs are often for small and benign RT indications at locations buried within the vessel or pipe wall. Probabilistic fracture mechanics calculations are described in this paper to predict the increases in vessel failure probabilities caused by the repair-induced flaws. Calculations address failures of embrittled vessel welds for pressurized thermal shock (PTS) transients. In this case small flaws, which are relatively common, can cause brittle fracture, such that the rarely encountered repair flaws of large sizes gave only modestly increased failure probabilities. The paper recommends the use of more discriminating ultrasonic examinations in place of RT examinations along with repair criteria based on a fitness-for-purpose approach that minimize the number of unjustified repairs. (author)

Evaluation of flaws in ferritic piping: ASME Code Appendix J, Deformation Plasticity Failure Assessment Diagram (DPFAD)

International Nuclear Information System (INIS)

Bloom, J.M.

1991-08-01

This report summarizes the methods and bases used by an ASME Code procedure for the evaluation of flaws in ferritic piping. The procedure is currently under consideration by the ASME Boiler and Pressure Vessel Code Committee of Section 11. The procedure was initially proposed in 1985 for the evaluation of the acceptability of flaws detected in piping during in-service inspection for certain materials, identified in Article IWB-3640 of the ASME Boiler and Pressure Vessel Code Section 11 ''Rules for In-service Inspection of Nuclear Power Plant Components.'' for which the fracture toughness is not sufficiently high to justify acceptance based solely on the plastic limit load evaluation methodology of Appendix C and IWB-3641. The procedure, referred to as Appendix J, originally included two approaches: a J-integral based tearing instability (J-T) analysis and the deformation plasticity failure assessment diagram (DPFAD) methodology. In Appendix J, a general DPFAD approach was simplified for application to part-through wall flows in ferritic piping through the use of a single DPFAD curve for circumferential flaws. Axial flaws are handled using two DPFAD curves where the ratio of flaw depth to wall thickness is used to determine the appropriate DPFAD curve. Flaws are evaluated in Appendix J by comparing the actual pipe applied stress with the allowable stress with the appropriate safety factors for the flaw size at the end of the evaluation period. Assessment points for circumferential and axial flaws are plotted on the appropriate failure assessment diagram. In addition, this report summarizes the experimental test predictions of the results of the Battelle Columbus Laboratory experiments, the Eiber experiments, and the JAERI tests using the Appendix J DPFAD methodology. Lastly, this report also provides guidelines for handling residual stresses in the evaluation procedure. 22 refs., 13 figs., 5 tabs

A charged-particle manipulator utilizing a co-axial tube electrodynamic trap with an integrated camera

International Nuclear Information System (INIS)

Jiang, L; Pau, S; Whitten, W B

2011-01-01

A charged-particle manipulator was designed and fabricated with an integrated imaging camera allowing real-time in-situ monitoring of trapped particle motion even when the trap device is under motion or rotation. The trap device was made of two co-axial electrically conductive tubes with diameters of 5.5 mm and 7 mm for the inner tube and outer tube, respectively; the imaging camera with its optical fiber bundle was integrated within the tubular trap device to realize a single instrument functioning as a manipulator. Motion of suspended microparticles of 3 μm to 50 μm in diameter can be monitored using the integrated camera regardless of the trap device orientations. This manipulator provides capability of controlled manipulation of trapped particles by tuning the operating conditions while monitoring the feedback of real-time particle motion. Imaging of suspended particles was not interrupted while the manipulator was translated and/or rotated. This integrated manipulator can be used for charged particle transport and repositioning.

Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

International Nuclear Information System (INIS)

Choi, Jong-Ho

2008-01-01

Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes for Heavy Water Reactors (HWRs), are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Project (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. The objective of the CRP was to inter-compare inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of the CRP, participants investigated the capability of different techniques to detect and characterize flaws. During the second phase, participants collaborated to determine the hydrogen concentration and to detect and characterize hydride blisters in zirconium alloy pressure tubes. Eight organizations from six countries, which operate HWRs, have participated in this CRP, Most of the techniques examined are well established and many of them are regularly used during in-service inspection of pressure tubes. The inter-comparison of these techniques provides a platform for identifying a particular technique (or a set of techniques), which is more accurate and reliable as compared to others for a specified task. The CRP also witnessed some new methodologies, which can be implemented on in-service inspection tools. These new techniques could complement the existing ones to overcome their limitations, thereby improving the reliability and accuracy of in-service inspection. This CRP also identified future areas of research and development. (author)

Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

International Nuclear Information System (INIS)

Diercks, D.R.

1997-02-01

The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Diercks, D.R. [Argonne National Lab., IL (United States)

1997-02-01

The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

The behavior of shallow flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Rolfe, S.T.

1991-11-01

Both analytical and experimental studies have shown that the effect of crack length, a, on the elastic-plastic toughness of structural steels is significant. The objective of this report is to recommend those research investigations that are necessary to understand the phenomenon of shallow behavior as it affects fracture toughness so that the results can be used properly in the structural margin assessment of reactor pressure vessels (RPVs) with flaws. Preliminary test results of A 533 B steel show an elevated crack-tip-opening displacement (CTOD) toughness similar to that observed for structural steels tested at the University of Kansas. Thus, the inherent resistance to fracture initiation of A 533 B steel with shallow flaws appears to be higher than that used in the current American Society of Mechanical Engineers (ASME) design curves based on testing fracture mechanics specimens with deep flaws. If this higher toughness of laboratory specimens with shallow flaws can be transferred to a higher resistance to failure in RPV design or analysis, then the actual margin of safety in nuclear vessels with shallow flaws would be greater than is currently assumed on the basis of deep-flaw test results. This elevation in toughness and greater resistance to fracture would be a very desirable situation, particularly for the pressurized-thermal shock (PTS) analysis in which shallow flaws are assumed to exist. Before any advantage can be taken of this possible increase in initiation toughness, numerous factors must be analyzed to ensure the transferability of the data. This report reviews those factors and makes recommendations of studies that are needed to assess the transferability of shallow-flaw toughness test results to the structural margin assessment of RPV with shallow flaws. 14 refs., 8 figs

Axial strain localization of CuCrZr tubes during manufacturing of ITER-like mono-block W/Cu components using HIP

International Nuclear Information System (INIS)

Zhao, S.X.; Peng, L.J.; Li, Q.; Wang, W.J.; Wei, R.; Qin, S.G.; Shi, Y.L.; Chang, S.P.; Xu, Y.; Liu, G.H.; Wang, T.J.; Luo, G.-N.

2014-01-01

Highlights: • Axial cracking and denting of CuCrZr tubes were observed. • Annealing the as-received tubes can alleviate cracking. • Denting results in the formation bonding flaws at the Cu/CuCrZr interfaces. - Abstract: Two forms of axial strain localization of CuCrZr tubes, i.e., cracking and denting, were observed during the manufacturing of ITER-like mono-block W/Cu components for EAST employing hot isostatic pressing (HIP). Microscopic investigations indicate that the occurrence of axial strain localization correlates to the heavily deformed Cu grains and elongated Cr-rich precipitates as well as highly anisotropic microstructures, which impair the circumferential ductility. Annealing the as-received tubes at 600 °C alleviates cracking due to partial recrystallization of Cu grains. However, the annealed tubes are still sensitive to wall thinning (caused by non-uniform polishing or tube bending), which results in denting. Denting may cause bonding flaws at CuCrZr/Cu interfaces and the underlying mechanisms are discussed. To some extent, denting seems do not affect the high heat flux performance of components. In this paper, we demonstrate that testing only the axial mechanical properties is not enough for manufacturers who use HIP or hot radial pressing technologies, especially for those anisotropic tubes

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

International Nuclear Information System (INIS)

Gutkin, L.; Scarth, D.A.

2014-01-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Gutkin, L.; Scarth, D.A. [Kinectrics Inc., Toronto, ON (Canada)

2014-07-01

Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

Flaw evolution monitoring by acoustic emission technique

International Nuclear Information System (INIS)

Ghia, S.; Sala, A.; Lucia, A.

1986-01-01

Flaw evolution monitoring during mechanical fatigue test has been performed by acoustic emission (AE) technique. Testing on 1:5 reduced scale vessel containing fabrication defects was carried out in the frame of an European program for pressure component residual life evaluation. Characteristics of AE signals associated to flaw evolution are discussed

Statistics applied to the testing of cladding tubes

International Nuclear Information System (INIS)

Perdijon, J.

1987-01-01

Cladding tubes, either steel or zircaloy, are generally given a 100 % inspection through ultrasonic non-destructive testing. This inspection may be completed beneficially with an eddy current test, as this is not sensitive to the same defects as those typically traced by ultrasonic testing. Unfortunately, the two methods (as with other non-destructive tests) exhibit poor precision; this means that a flaw, whose size is close to that denoted as rejection limit, may be accepted or rejected. Currently, rejection, i.e. the measurement above which a tube is rejected, is generally determined through measuring a calibration tube at regular time intervals, and the signal of a given tube is compared to that of the most recently completed calibration. This measurement is thus subject to variations which can be attributed to an actual shift of adjustments as well as to poor precision. For this reason, monitoring instrument adjustments using the so-called control chart method are proposed

A Novel Approach for an Integrated Straw Tube-Microstrip Detector

Science.gov (United States)

Basile, E.; Bellucci, F.; Benussi, L.; Bertani, M.; Bianco, S.; Caponero, M. A.; Colonna, D.; Di Falco, F.; Fabbri, F. L.; Felli, F.; Giardoni, M.; La Monaca, A.; Mensitieri, G.; Ortenzi, B.; Pallotta, M.; Paolozzi, A.; Passamonti, L.; Pierluigi, D.; Pucci, C.; Russo, A.; Saviano, G.; Casali, F.; Bettuzzi, M.; Bianconi, D.; Baruffaldi, F.; Perilli, E.; Massa, F.

2006-06-01

We report on a novel concept of silicon microstrips and straw tubes detector, where integration is accomplished by a straw module with straws not subjected to mechanical tension in a Rohacell/spl reg/ lattice and carbon fiber reinforced plastic shell. Results on mechanical and test beam performances are reported as well.

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

Energy Technology Data Exchange (ETDEWEB)

Belle R. Upadhyaya; J. Wesley Hines

2004-09-27

Integrity monitoring and flaw diagnostics of flat beams and tubular structures was investigated in this research task using guided acoustic signals. A piezo-sensor suite was deployed to activate and collect Lamb wave signals that propagate along metallic specimens. The dispersion curves of Lamb waves along plate and tubular structures are generated through numerical analysis. Several advanced techniques were explored to extract representative features from acoustic time series. Among them, the Hilbert-Huang transform (HHT) is a recently developed technique for the analysis of non-linear and transient signals. A moving window method was introduced to generate the local peak characters from acoustic time series, and a zooming window technique was developed to localize the structural flaws. The time-frequency analysis and pattern recognition techniques were combined for classifying structural defects in brass tubes. Several types of flaws in brass tubes were tested, both in the air and in water. The techniques also proved to be effective under background/process noise. A detailed theoretical analysis of Lamb wave propagation was performed and simulations were carried out using the finite element software system ABAQUS. This analytical study confirmed the behavior of the acoustic signals acquired from the experimental studies. The report presents the background the analysis of acoustic signals acquired from piezo-electric transducers for structural defect monitoring. A comparison of the use of time-frequency techniques, including the Hilbert-Huang transform, is presented. The report presents the theoretical study of Lamb wave propagation in flat beams and tubular structures, and the need for mode separation in order to effectively perform defect diagnosis. The results of an extensive experimental study of detection, location, and isolation of structural defects in flat aluminum beams and brass tubes are presented. The results of this research show the feasibility of on

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

International Nuclear Information System (INIS)

Upadhyaya, Belle R.; Hines, J. Wesley

2004-01-01

Integrity monitoring and flaw diagnostics of flat beams and tubular structures was investigated in this research task using guided acoustic signals. A piezo-sensor suite was deployed to activate and collect Lamb wave signals that propagate along metallic specimens. The dispersion curves of Lamb waves along plate and tubular structures are generated through numerical analysis. Several advanced techniques were explored to extract representative features from acoustic time series. Among them, the Hilbert-Huang transform (HHT) is a recently developed technique for the analysis of non-linear and transient signals. A moving window method was introduced to generate the local peak characters from acoustic time series, and a zooming window technique was developed to localize the structural flaws. The time-frequency analysis and pattern recognition techniques were combined for classifying structural defects in brass tubes. Several types of flaws in brass tubes were tested, both in the air and in water. The techniques also proved to be effective under background/process noise. A detailed theoretical analysis of Lamb wave propagation was performed and simulations were carried out using the finite element software system ABAQUS. This analytical study confirmed the behavior of the acoustic signals acquired from the experimental studies. The report presents the background the analysis of acoustic signals acquired from piezo-electric transducers for structural defect monitoring. A comparison of the use of time-frequency techniques, including the Hilbert-Huang transform, is presented. The report presents the theoretical study of Lamb wave propagation in flat beams and tubular structures, and the need for mode separation in order to effectively perform defect diagnosis. The results of an extensive experimental study of detection, location, and isolation of structural defects in flat aluminum beams and brass tubes are presented. The results of this research show the feasibility of on

Ultrasonic inspection of inpile tubes

International Nuclear Information System (INIS)

Boyd, D.M.; Bossi, H.

1985-01-01

The in-service inspection (ISI) of inpile tubes can be performed accurately and safely with a semiautomatic ultrasonic inspection system. The ultrasonic technique uses a set of multiple transducers to detect and size cracks, voids, and laminations radially and circumferentially. Welds are also inspected for defects. The system is designed to inspect stainless steel and Inconel tubes ranging from 53.8 mm (2.12 in.) to 101.6 mm (4 in.) inner diameter with wall thickness on the order of 5 mm. The inspection head contains seven transducers mounted in a surface-following device. Six angle-beam transducers generate shear waves in the tubes. Two of the six are oriented to detect circumferential cracks, and two detect axial cracks. Although each of these four transducers is used in the pulse-echo mode, they are oriented in aligned sets so pitch-catch operation is possible if desired. The remaining angle-beam transducers are angulated to detect flaws that are off axial or circumferential orientation. The seventh transducer is used for longitudinal inspection and detects and sizes laminar-type defects

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

Pandey, M.D.; Radford, D.D.

2008-01-01

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach

Ultrasonic flaw detection in a monorail box beam

Science.gov (United States)

Zheng, Peng; Greve, David W.; Oppenheim, Irving J.

2009-03-01

A steel box beam in a monorail application is constructed with an epoxy grout wearing surface, precluding visual inspection of its top flange. This paper describes a sequence of experimental research tasks to develop an ultrasonic system to detect flaws (such as fatigue cracks) in that flange, and the results of a field test to demonstrate system performance. The problem is constrained by the fact that the flange is exposed only along its longitudinal edges, and by the fact that permanent installation of transducers at close spacing was deemed to be impractical. The system chosen for development, after experimental comparison of alternate technologies, features angle-beam ultrasonic transducers with fluid coupling to the flange edge; the emitting transducers create transverse waves that travel diagonally across the width of the flange, where an array of receiving transducers detect flaw reflections and flaw shadows. The system rolls along the box beam, surveying (screening) the top flange for the presence of flaws. In a first research task, conducted on a full-size beam specimen, we compared waves generated from different transducer locations, either the flange edge or the web face, and at different frequency ranges. At relatively low frequencies, such as 100 kHz, we observed Lamb wave modes, and at higher frequency, in the MHz range, we observed nearlylongitudinal waves with trailing pulses. In all cases we observed little attenuation by the wearing surface and little influence of reflection at the web-flange joints. At the conclusion of this task we made the design decision to use edgemounted transducers at relatively high frequency, with correspondingly short wavelength, for best scattering from flaws. In a second research task we conducted experiments at 55% scale on a steel plate, with machined flaws of different size, and detected flaws of target size for the intended application. We then compared the performance of bonded transducers, fluid

A Novel Approach for an Integrated Straw tube-Microstrip Detector

OpenAIRE

Basile, E.; Bellucci, F.; Benussi, L.; Bertani, M.; Bianco, S.; Caponero, M. A.; Colonna, D.; Di Falco, F.; Fabbri, F. L.; Felli, F.; Giardoni, M.; La Monaca, A.; Mensitieri, G.; Ortenzi, B.; Pallotta, M.

2005-01-01

We report on a novel concept of silicon microstrips and straw tubes detector, where integration is accomplished by a straw module with straws not subjected to mechanical tension in a Rohacell $^{\\circledR}$ lattice and carbon fiber reinforced plastic shell. Results on mechanical and test beam performances are reported on as well.

Round robin tests of the PISC III programme on defective steam generators tubes

International Nuclear Information System (INIS)

Birac, C.; Herkenrath, H.; Crutzen, S.; Miyake, Y.; Maciga, G.

1991-11-01

The PISC III actions are intended to extend the results and methodologies of the previous PISC exercises, i.e. the assessment of the capabilities of the various examination techniques when used on real or realistic flaws in real components under real conditions of inspection. Being aware of the industrial problems that the degradation of steam generator tubes can create, the PISC III management board decided to include in the PISC III programme a special action on steam generator tubes testing (SGT). (author)

Nondestructive detection of surface flaws in materials by infrared thermography

International Nuclear Information System (INIS)

Ishii, Toshimitsu; Ooka, Norikazu; Eto, Motokuni; Hoshiya, Taiji; Okamoto, Yoshizo

1999-01-01

Infrared thermography is one of the useful remote sensing techniques applied to the nondestructive detection of surface flaws in materials. Radiation temperatures of the specimen surface and surrounding walls as well as the difference in them are crucial factors to detect surface flaws from thermal images, and it is essential that these factors be properly evaluated beforehand in order to detect the flaws by infrared thermography. In this study, the radiation temperature of nuclear graphite specimens heated uniformly was measured by infrared thermography to evaluate the radiation characteristics such as emissivity, radiosity coefficient and variation of radiation temperature. The influence of the temperature difference between the test specimen and its surroundings on the limit of detection of pinhole flaws was discussed on the basis of the thermal images of graphite specimen with surface flaws. It was found that the thermal image of a small flaw was clearly visible with increase in the temperature difference. (author)

A new repair criterion for steam generator tubes with axial cracks based on probabilistic integrity assessment

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Su; Oh, Chang-Kyun [KEPCO Engineering and Construction Company, Inc., 269, Hyeoksin-ro, Gimcheon, Gyeongsangbuk-do 39660 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, College of Engineering, Kyung Hee University, 1732 Deokyoungdaero, Giheung, Yongin, Gyeonggi 446-701 (Korea, Republic of)

2017-03-15

Highlights: • Probabilistic assessment was performed for axially cracked steam generator tubes. • The threshold crack sizes were determined based on burst pressures of the tubes. • A new repair criterion was suggested as a function of operation time. - Abstract: Steam generator is one of the major components in a nuclear power plant, and it consists of thousands of thin-walled tubes. The operating record of the steam generators has indicated that a number of axial cracks due to stress corrosion have been frequently detected in the tubes. Since the tubes are closely related to the safety and also the efficiency of a nuclear power plant, an establishment of the appropriate repair criterion for the defected tubes and its applications are necessary. The objective of this paper is to develop an accurate repair criterion for the tubes with axial cracks. To do this, a thorough review is performed on the key parameters affecting the tube integrity, and then the probabilistic integrity assessment is carried out by considering the various uncertainties. In addition, the sizes of critical crack are determined by comparing the burst pressure of the cracked tube with the required performance criterion. Based on this result, the new repair criterion for the axially cracked tubes is defined from the reasonably conservative value such that the required performance criterion in terms of the burst pressure is able to be met during the next operating period.

Flaw evaluation of pressure vessel in pressurized water reactor

International Nuclear Information System (INIS)

Park, Ki Sung; Kim, Min Geol; Jeon, Chae Hong; Rhim, Soon Hyung; Kim, Seung Tae

1999-01-01

Flaw evaluation should be performed to determine the acceptance of a surface or a subsurface flaw detected during the in-service inspection without any repair or replacement. In this paper, the evaluation methodology and procedure were established according to ASME code Sec. XI and the evaluation program was coded. Using this program, a field engineer who doesn't have enough knowledge on fracture mechanics may be able to perform prompt and accurate flaw evaluation on site and decide whether a detected flaw be allowable or not. Analysis results were compared with those obtained from Westinghouse program called KCAL and FCG. Both results made good agreement and accuracy of the program developed in this paper was verified.=20

Structural integrity evaluation of SG tube with surface wear-type defects

International Nuclear Information System (INIS)

Kim, Jong Min; Huh, Nam Su; Chang, Yoon Suk; Kim, Young Jin; Hwang, Seong Sik; Kim, Joung Soo

2006-01-01

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

Steam Generator tube integrity -- US Nuclear Regulatory Commission perspective

International Nuclear Information System (INIS)

Murphy, E.L.; Sullivan, E.J.

1997-01-01

In the US, the current regulatory framework was developed in the 1970s when general wall thinning was the dominant degradation mechanism; and, as a result of changes in the forms of degradation being observed and improvements in inspection and tube repair technology, the regulatory framework needs to be updated. Operating experience indicates that the current U.S. requirements should be more stringent in some areas, while in other areas they are overly conservative. To date, this situation has been dealt with on a plant-specific basis in the US. However, the NRC staff is now developing a proposed steam generator rule as a generic framework for ensuring that the steam generator tubes are capable of performing their intended safety functions. This paper discusses the current U.S. regulatory framework for assuring steam generator (SG) tube integrity, the need to update this regulatory framework, the objectives of the new proposed rule, the US Nuclear Regulatory Commission (NRC) regulatory guide (RG) that will accompany the rule, how risk considerations affect the development of the new rule, and some outstanding issues relating to the rule that the NRC is still dealing with

A Bayesian approach to modeling and predicting pitting flaws in steam generator tubes

International Nuclear Information System (INIS)

Yuan, X.-X.; Mao, D.; Pandey, M.D.

2009-01-01

Steam generators in nuclear power plants have experienced varying degrees of under-deposit pitting corrosion. A probabilistic model to accurately predict pitting damage is necessary for effective life-cycle management of steam generators. This paper presents an advanced probabilistic model of pitting corrosion characterizing the inherent randomness of the pitting process and measurement uncertainties of the in-service inspection (ISI) data obtained from eddy current (EC) inspections. A Markov chain Monte Carlo simulation-based Bayesian method, enhanced by a data augmentation technique, is developed for estimating the model parameters. The proposed model is able to predict the actual pit number, the actual pit depth as well as the maximum pit depth, which is the main interest of the pitting corrosion model. The study also reveals the significance of inspection uncertainties in the modeling of pitting flaws using the ISI data: Without considering the probability-of-detection issues and measurement errors, the leakage risk resulted from the pitting corrosion would be under-estimated, despite the fact that the actual pit depth would usually be over-estimated.

Flaw detection device

International Nuclear Information System (INIS)

Sasahara, Toshihiko

1998-01-01

The present invention provides a device for detecting welded portions of a reactor pressure vessel. Namely, the device of the present invention comprises (1) a casing to be disposed on the surface to be detected, (2) a probe driving means loaded to the casing, (3) a probe driven along the surface to be detected and (4) a pressure reduction means for keeping the hollow portion in the casing to an evacuated atmosphere. The casing comprises a flexible suction edge to be tightly in contact with the surface to be tested for maintaining the air tight state, (6) a guide wheel for moving the casing along the surface to be tested and (7) a handle for performing transferring operation. The flaw detection device thus constituted has following features. The working efficiency upon conducting detection is improved. The influence of the weight of the device on the detection is small. The device can be applied on the surface of a nonmagnetic material. The efficiency for the flaw detection can be improved. (I.S.)

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

International Nuclear Information System (INIS)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size? The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service

Advances in flaw evaluation procedures and acceptance criteria for reactor piping

Energy Technology Data Exchange (ETDEWEB)

Gamble, R.M.; Zahoor, A.; Norris, D.M.

1986-01-01

During the past several years, intergranular stress corrosion cracks (IGSCC) have been detected in stainless steel piping in boiling water reactors (BWRs) and have resulted in an increased number of flaw evaluations. To reduce the outage time associated with evaluating IGSCC, various research and ASME code groups have spent significant effort to provide utility personnel with efficient means to detect, classify, and size flaws, and to determine suitability for return to service for flawed stainless steel piping. One of the several nondestructive evaluation technologies that has received considerable attention is fracture mechanics, the discipline that considers the failure of flawed material. Fracture mechanics can be used to answer two key questions concerning return to service of flawed pipe: (a) what is the largest flaw size that can be returned to service and still maintain adequate safety margins at the applied loads, and (b) how much operating time remains before the crack reaches the largest allowable size. The purpose of this paper is to provide an overview of the recently developed ASME code Section XI flaw size evaluation procedure and acceptance criteria for stainless steel piping and their application by BWR owners to efficiently determine if flaws found by nondestructive examination are acceptable for continued service.

Determination of K-factors for arbitrarily shaped flaws at pressure vessel nozzle corners

International Nuclear Information System (INIS)

Bryson, J.W.

1979-01-01

Photoelastic and finite element studies are being conducted to determine Mode I stress intensity factor distributions along arbitrarily shaped flaw fronts at pressure vessel nozzle corners. Comparisons of results from NOZ-FLAW, BIGIF, and the photoelastic studies showed that (1) good agreement was obtained between NOZ-FLAW and the photoelastically determined K 1 's for the deep flaw in an ITV model, (2) good agreement was obtained between NOZ-FLAW BIGIF for shallow and moderately deep flaws in a BWR model, and (3) less satisfactory agreement was obtained between NOZ- FLAW and the photoelastic results for the BWR models, particularly for moderately deep to deep flaws. Attempts are presently being made at understanding and explaining the discrepancies between the two

Estimation of Back-Surface Flaw Depth by Laminated Piezoelectric Highpolymer Film

Directory of Open Access Journals (Sweden)

Akinobu YAMAMOTO

2009-08-01

Full Text Available Piezoelectric thin films have been used to visualize back surface flaws in plates. If the plate with a surface flaw is deformed, the strain distribution appears on the other surface reflecting the location and the shape of the flaw. Such surface strain distribution can be transformed into the electric potential distribution on the piezoelectric film mounted on the plate surface. This paper deals with a NDE technique to estimate the depth of a back-surface flaw from the electric potential distribution on a laminated piezoelectric thin film. It is experimentally verified that the flaw depth can be exactly estimated by the peak height of the electric potential distribution.

NRC integrated program for the resolution of Unresolved Safety Issues A-3, A-4 and A-5 regarding steam generator tube integrity: Final report

International Nuclear Information System (INIS)

1988-09-01

This report presents the results of the NRC integrated program for the resolution of Unresolved Safety Issues (USIs) A-3, A-4, and A-5 regarding steam generator tube integrity. A generic risk assessment is provided and indicates that risk from steam generator tube rupture (SGTR) events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR 50, Appendices A and B. This report also identifies a number of staff-recommended actions that the staff finds can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of pressurized water reactors (PWRs) to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such actions do not constitute NRC requirements. In addition, this report describes a number of ongoing staff actions and studies involving steam generator issues which are being pursued to provide added assurance that risk from SGTR events will continue to be small. 146 refs., 5 figs., 11 tabs

Vibro-impact responses of a tube with tube--baffle interaction

International Nuclear Information System (INIS)

Shin, Y.S.; Sass, D.E.; Jendrzejczyk, J.A.

1978-01-01

The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments

Crack growth of throughwall flaw in Alloy 600 tube during leak testing

International Nuclear Information System (INIS)

Bahn, Chi Bum; Majumdar, Saurin

2015-01-01

Graphical abstract: - Highlights: • A series of leak testing was conducted at a constant pressure and room temperature. • The time-dependent increase in the leak rate was observed. • The fractography revealed slip offsets and crystallographic facets. • Time-dependent plasticity at the crack tip caused the slip offsets. • Fatigue by jet/structure interaction caused the crystallographic facets. - Abstract: We examined the issue of whether crack growth in a full thickness material can occur in a leaking crack. A series of leak tests was conducted at a room temperature and constant pressure (17.3 MPa) with Alloy 600 tube specimens containing a tight rectangular throughwall axial fatigue crack. To exclude a potential pulsation effect by a high pressure pump, the test water was pressurized by using high pressure nitrogen gas. Fractography showed that crack growth in the full thickness material can occur in the leaking crack by two mechanisms: time-dependent plasticity at the crack tip and fatigue induced by jet/structure interaction. The threshold leak rate at which the jet/structure interaction was triggered was between 1.3 and 3.3 L/min for the specific heat of the Alloy 600 tube tested

Development and validation of a simulation tool dedicated to eddy current non destructive testing of tubes

International Nuclear Information System (INIS)

Reboud, Ch.

2006-09-01

Eddy current testing (ECT) technique is widely used in industrial fields such as iron and steel industry. Dedicated simulation tools provide a great assistance for the optimisation of ECT processes. CEA and the Vallourec Research Center have collaborated in order to develop a simulation tool of ECT of tubes. The volume integral method has been chosen for the resolution of Maxwell equations in a stratified medium, in order to get accurate results with a computation time short enough to carry out optimisation or inversion procedures. A fast model has been developed for the simulation of ECT of non magnetic tubes using specific external probes. New flaw geometries have been modelled: holes and notches with flat bottom. Validations of the developments, which have been integrated to the CIVA platform, have been carried out using experimental data recorded in laboratory conditions and in. industrial conditions, successively. The integral equations derived are solved using the Galerkin variant of the method of moments with pulse functions as projection functions. In order to overcome some memory limitations, other projection functions have been considered. A new discretization scheme based on non-uniform B-Splines of degree 1 or 2 has been implemented, which constitutes an original contribution to the existing literature. The decrease of the mesh size needed to get a given accuracy on the result may lead to the simulation of more complex ECT configurations. (author)

Ductile fracture of cylindrical vessels containing a large flaw

Science.gov (United States)

Erdogan, F.; Irwin, G. R.; Ratwani, M.

1976-01-01

The fracture process in pressurized cylindrical vessels containing a relatively large flaw is considered. The flaw is assumed to be a part-through or through meridional crack. The flaw geometry, the yield behavior of the material, and the internal pressure are assumed to be such that in the neighborhood of the flaw the cylinder wall undergoes large-scale plastic deformations. Thus, the problem falls outside the range of applicability of conventional brittle fracture theories. To study the problem, plasticity considerations are introduced into the shell theory through the assumptions of fully-yielded net ligaments using a plastic strip model. Then a ductile fracture criterion is developed which is based on the concept of net ligament plastic instability. A limited verification is attempted by comparing the theoretical predictions with some existing experimental results.

Development and validation of a simulation tool dedicated to eddy current non destructive testing of tubes; Developpement d'un modele electromagnetique 3D pour la simulation du controle par Courants de Foucault de tubes en fabrication

Energy Technology Data Exchange (ETDEWEB)

Reboud, Ch

2006-09-15

Eddy current testing (ECT) technique is widely used in industrial fields such as iron and steel industry. Dedicated simulation tools provide a great assistance for the optimisation of ECT processes. CEA and the Vallourec Research Center have collaborated in order to develop a simulation tool of ECT of tubes. The volume integral method has been chosen for the resolution of Maxwell equations in a stratified medium, in order to get accurate results with a computation time short enough to carry out optimisation or inversion procedures. A fast model has been developed for the simulation of ECT of non magnetic tubes using specific external probes. New flaw geometries have been modelled: holes and notches with flat bottom. Validations of the developments, which have been integrated to the CIVA platform, have been carried out using experimental data recorded in laboratory conditions and in. industrial conditions, successively. The integral equations derived are solved using the Galerkin variant of the method of moments with pulse functions as projection functions. In order to overcome some memory limitations, other projection functions have been considered. A new discretization scheme based on non-uniform B-Splines of degree 1 or 2 has been implemented, which constitutes an original contribution to the existing literature. The decrease of the mesh size needed to get a given accuracy on the result may lead to the simulation of more complex ECT configurations. (author)

A time-domain synthetic aperture ultrasound imaging method for material flaw quantification with validations on small-scale artificial and natural flaws.

Science.gov (United States)

Guan, Xuefei; He, Jingjing; Rasselkorde, El Mahjoub

2015-02-01

A direct time-domain reconstruction and sizing method of synthetic aperture focusing technique (SAFT) is developed to improve the spatial resolution and sizing accuracy for phased-array ultrasonic inspections. The basic idea of the reconstruction algorithm is to coherently superimpose multiple A-scan measurements, incorporating the phase information of the sampling points. The algorithm involves data mapping and in-phase summation according to time-of-flight (TOF). Data mapping refers to the process of placing each of the sampling points to a two-/three-dimensional grid that represents the geometry model of the object being inspected. The value for each of the cells of the grid is a summation of all sampling points mapped into the cell. A sizing method based on the concept of 6 dB-drop is proposed to characterize the flaw boundary. The extents, orientation and the shape of the flaw can then be inferred to provide more information for life assessment calculations. Lab experiments are performed using a 10 MHz phased-array ultrasonic transducer to collect data from a cylinder material block with closely spaced artificial flaws and from a material block with a natural flaw. The developed method is used to process the experimental data to characterize the flaws. Using the developed method, the improvement of spatial resolution is observed. Results indicate that four closely spaced 0.794 mm-diameter flat-bottomed holes are clearly identified, and the quantification of size and orientation of the natural flaw is very close to the actual measurement made from digital microscopy after cutting the testing piece apart. Copyright © 2014 Elsevier B.V. All rights reserved.

An intelligent software approach to ultrasonic flaw classification in weldments

International Nuclear Information System (INIS)

Song, Sung Jin; Kim, Hak Joon; Lee, Hyun

1997-01-01

Ultrasonic pattern recognition is the most effective approach to the problem of discriminating types of flaws in weldments based on ultrasonic flaw signals. In spite of significant progress on this methodology, it has not been widely used in practical ultrasonic inspection of weldments in industry. Hence, for the convenient application of this approach in many practical situations, we develop an intelligent ultrasonic signature classification software which can discriminate types of flaws in weldments using various tools in artificial intelligence such as neural networks. This software shows excellent performances in an experimental problem where flaws in weldments are classified into two categories of cracks and non-cracks.

Limits to the Recognizability of Flaws in Non-Destructive Testing Steam-Generator Tubes for Nuclear-Power Plants

International Nuclear Information System (INIS)

Kuhlmann, A.; Adamsky, F.-J.

1965-01-01

In the Federal Republic of Germany there are nuclear reactors under construction with steam generators inside the reactor pressure-vessel. As a result design repairs of steam- generator tubes are very difficult and cause large shut-down times of the nuclear-power plant. It is known that numerous troubles in operating conventional power plants are results of steam-generator tube damages. Because of the high total costs of these reactors it. is necessary to construct the steam generators especially in such a manner that the load factor of the power plant is as high as possible. The Technischer Überwachungs-Verein Rheinland was charged to supervise and to test fabrication and construction of the steam generators to see that this part of the plant was as free of defects as possible. The experience gained during this work is of interest for manufacture and construction of steam generators for nuclear-power plants in general. This paper deals with the efficiency limits of non-destructive testing steam-generator tubes. The following tests performed will be discussed in detail: (a) Automatic ultrasonic testing of the straight tubes in the production facility; (b) Combined ultrasonic and radiographic testing of the bent tubes and tube weldings; (c) Other non-destructive tests. (author) [fr

A risk-informed approach to the assessment of DHC initiation in pressure tubes

International Nuclear Information System (INIS)

Sahoo, A.K.; Pandey, M.D.

2009-01-01

The delayed hydride cracking (DHC) of pressure tubes is a serious form of degradation in the reactor core. Flaws in pressure tubes generated by fretting or any other mechanism are potential stress raisers that could become sites of DHC initiation under right circumstances. CSA standard N285.8 recommends deterministic and probabilistic procedures for the assessment of potential for DHC initiation from planar flaws. The deterministic method is simple, but it lacks a risk-informed basis for the assessment. A full probabilistic method based on simulations is tedious to implement. This paper presents an innovative, semi-probabilistic method that bridges the gap between a simple deterministic analysis and complex simulations. In the proposed method, the deterministic assessment criterion of CSA N285.8 standard is calibrated to specified target probabilities of DHC initiation using the concept of partial factors. The main advantage of the proposed approach is that it provides a practical, risk-informed basis for DHC initiation assessment while retaining the simplicity of the deterministic method. (author)

Statistical flaw strength distributions for glass fibres: Correlation between bundle test and AFM-derived flaw size density functions

International Nuclear Information System (INIS)

Foray, G.; Descamps-Mandine, A.; R’Mili, M.; Lamon, J.

2012-01-01

The present paper investigates glass fibre flaw size distributions. Two commercial fibre grades (HP and HD) mainly used in cement-based composite reinforcement were studied. Glass fibre fractography is a difficult and time consuming exercise, and thus is seldom carried out. An approach based on tensile tests on multifilament bundles and examination of the fibre surface by atomic force microscopy (AFM) was used. Bundles of more than 500 single filaments each were tested. Thus a statistically significant database of failure data was built up for the HP and HD glass fibres. Gaussian flaw distributions were derived from the filament tensile strength data or extracted from the AFM images. The two distributions were compared. Defect sizes computed from raw AFM images agreed reasonably well with those derived from tensile strength data. Finally, the pertinence of a Gaussian distribution was discussed. The alternative Pareto distribution provided a fair approximation when dealing with AFM flaw size.

Vibro-impact responses of a tube with tube--baffle interaction. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Shin, Y S; Sass, D E; Jendrzejczyk, J A

1978-01-01

The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments.

Miniaturised Prandtl tube with integrated pressure sensors for micro-thruster plume characterisation

NARCIS (Netherlands)

Dijkstra, Marcel; Ma, Kechun; de Boer, Meint J.; Groenesteijn, Jarno; Lötters, Joost Conrad; Wiegerink, Remco J.

2014-01-01

A miniaturised Prandtl-tube sensor incorporating a 6 mm long 40 μm diameter microchannel with integrated pressure sensors has been realised. The sensor has been designed for the characterisation of rarefied plume flow from a MEMS-based monopropellant propulsion system for high-accuracy attitude

Real time automatic discriminating of ultrasonic flaws

International Nuclear Information System (INIS)

Suhairy Sani; Mohd Hanif Md Saad; Marzuki Mustafa; Mohd Redzwan Rosli

2009-01-01

This paper is concerned with the real time automatic discriminating of flaws from two categories; i. cracks (planar defect) and ii. Non-cracks (volumetric defect such as cluster porosity and slag) using pulse-echo ultrasound. The raw ultrasonic flaws signal were collected from a computerized robotic plane scanning system over the whole of each reflector as the primary source of data. The signal is then filtered and the analysis in both time and frequency domain were executed to obtain the selected feature. The real time feature analysis techniques measured the number of peaks, maximum index, pulse duration, rise time and fall time. The obtained features could be used to distinguish between quantitatively classified flaws by using various tools in artificial intelligence such as neural networks. The proposed algorithm and complete system were implemented in a computer software developed using Microsoft Visual BASIC 6.0 (author)

STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

Directory of Open Access Journals (Sweden)

HEOK-SOON LIM

2014-02-01

Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

Energy Technology Data Exchange (ETDEWEB)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

2014-02-15

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

International Nuclear Information System (INIS)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

2014-01-01

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

Microstructure, flaw tolerance, and reliability of Ce-TZP and Y-TZP ceramics

International Nuclear Information System (INIS)

Readey, M.J.; McCallen, C.L.

1995-01-01

Ce-TZP and Y-TZP ceramics were heat-treated for various times and temperatures in order to vary the microstructure. Flaw tolerance was investigated using the indentation-strength test. Reliability was quantified using conventional two-parameter Weibull statistics. Some Ce-TZP specimens were indented at slightly elevated temperatures where no transformation was observed. Results indicated that the Ce-TZP specimens were extremely flaw tolerant, and showed a relatively high Weibull modulus that scaled with both R-curve behavior and flaw tolerance. Y-TZP, on the other hand, with very little if any R-curve behavior or flaw tolerance, had a low Weibull modulus. The results also show that flaw history, i.e., whether or not a transformation zone exists along the wake of the crack, has a significant influence on strength. Strength was much less dependent on initial crack size when the crack had an associated transformation zone, whereas strength was highly dependent on cracks typical of natural processing defects. It is argued that the improvement in reliability, flaw tolerance, and dependence on flaw history are all ramifications of pronounced R-curve behavior

Integrating a Traveling Wave Tube into an AECR-U ion source

Energy Technology Data Exchange (ETDEWEB)

Covo, Michel Kireeff; Benitez, Janilee Y.; Ratti, Alessandro; Vujic, Jasmina L.

2011-07-01

An RF system of 500W - 10.75 to 12.75 GHz was designed and integrated into the Advanced Electron Cyclotron Resonance - Upgrade (AECR-U) ion source of the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory. The AECR-U produces ion beams for the Cyclotron giving large flexibility of ion species and charge states. The broadband frequency of a Traveling Wave Tube (TWT) allows modifying the volume that couples and heats the plasma. The TWT system design and integration with the AECR-U ion source and results from commissioning are presented.

Irradiation effects and the duplication of detected flaws in service

International Nuclear Information System (INIS)

Mager, T.R.

1976-01-01

ASME Code procedure for evaluating the acceptability of flaws detected during in-service inspection is revised. Critical crack size for instability is proposed as criteria for detected flaws in operating plants

Detection of flaws below curved surfaces

International Nuclear Information System (INIS)

Elsley, R.K.; Addison, R.C.; Graham, L.J.

1983-01-01

A measurement model has been developed to describe ultrasonic measurements made with circular piston transducers in parts with flat or cylindrically curved surfaces. The model includes noise terms to describe electrical noise, scatterer noise and echo noise as well as effects of attenuation, diffraction and Fresnel loss. An experimental procedure for calibrating the noise terms of the model was developed. Experimental measurements were made on a set of known flaws located beneath a cylindrically curved surface. The model was verified by using it to correct the experimental measurements to obtain the absolute scattering amplitude of the flaws. For longitudinal wave propagation within the part, the derived scattering amplitudes were consistent with predictions at internal angles of less than 30 0 . At larger angles, focusing and aberrations caused a lack of agreement; the model needs further refinement in this case. For shear waves, it was found that the frequency for optimum flaw detection in the presence of material noise is lower than that for longitudinal waves; lower frequency measurements are currently in progress. The measurement model was then used to make preliminary predictions of the best experimental measurement technique for the detection of cracks located under cylindrically curved surfaces

Common methodological flaws in economic evaluations.

Science.gov (United States)

Drummond, Michael; Sculpher, Mark

2005-07-01

Economic evaluations are increasingly being used by those bodies such as government agencies and managed care groups that make decisions about the reimbursement of health technologies. However, several reviews of economic evaluations point to numerous deficiencies in the methodology of studies or the failure to follow published methodological guidelines. This article, written for healthcare decision-makers and other users of economic evaluations, outlines the common methodological flaws in studies, focussing on those issues that are likely to be most important when deciding on the reimbursement, or guidance for use, of health technologies. The main flaws discussed are: (i) omission of important costs or benefits; (ii) inappropriate selection of alternatives for comparison; (iii) problems in making indirect comparisons; (iv) inadequate representation of the effectiveness data; (v) inappropriate extrapolation beyond the period observed in clinical studies; (vi) excessive use of assumptions rather than data; (vii) inadequate characterization of uncertainty; (viii) problems in aggregation of results; (ix) reporting of average cost-effectiveness ratios; (x) lack of consideration of generalizability issues; and (xi) selective reporting of findings. In each case examples are given from the literature and guidance is offered on how to detect flaws in economic evaluations.

A study on the dimensioning of flaws by acoustical holography

International Nuclear Information System (INIS)

Yamamoto, Michio; Ando, Tomozumi; Enami, Koji; Yajima, Minoru; Fukui, Shigetaka.

1978-01-01

As a means of evaluating the safety of flawed pressure vessels and other structures against fracture, fracture mechanics has come to be applied. For the application of fracture mechanics it is necessary to get information concerning the sizes and shapes of flaws. The ultrasonic flaw detection method that is widely used as a nondestructive inspection method cannot measure the sizes and shapes of flaws accurately. Considering that acoustical holography is an useful means for the dimensioning of flaws, we performed basic tests on this method and obtained the following results: (1) The measured values of artificial flaws (flat bottom drilled holes: 5 - 36 mm) made on a steel plate of 150 mm thick showed a good linear relation with their actual sizes and scatter in the measured values was +-3 - 6 mm. (2) The measured values of fatigue cracks (length: 5 - 57 mm) introduced into a steel plate of 150 mm thick also showed a good linear relation with their actual sizes and scatter in the measured values was +-3 mm. (3) It was found that acoustical holography can also be applied to heavy section cast steels. (4) The method of correcting distortion caused by curved surface was investigated by computer-aided simulation and it was considered that such distortion can be corrected by radial scanning of a transducer. (author)

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

Analysis of portable gamma flaw detectors concerning radiation hygiene

International Nuclear Information System (INIS)

Makarova, T.V.

1982-01-01

Design and shields of gamma flaw detectors as one of the main factors responsible for personnel dose were studied. The analysis was conducted using the results of radiation hygienic surveys of gamma flaw detection laboratories functioning constantly in Estonia. It is shown that recently the replacement of GUP apparatuses by flaw detectors of RID and ''Gamma-RID'' (types which have design and shielding advantages is observed. However personnel doses have not reduced considerably for the last 10 years. This fact is attributed to design disadvantages of the RID and ''Gamma-RID'' apparatuses the removing of which will give the decreasing of annual personnel dose by 80 %

Modelling of thermal behaviour of iron oxide layers on boiler tubes

Science.gov (United States)

Angelo, J. D.; Bennecer, A.; Kaczmarczyk, S.; Picton, P.

2016-05-01

Slender boiler tubes are subject to localised swelling when they are expose to excessive heat. The latter is due to the formation of an oxide layer, which acts as an insulation barrier. This excessive heat can lead to microstructural changes in the material that would reduce the mechanical strength and would eventually lead to critical and catastrophic failure. Detecting such creep damage remains a formidable challenge for boiler operators. It involves a costly process of shutting down the plant, performing electromagnetic and ultrasonic non-destructive inspection, repairing or replacing damaged tubes and finally restarting the plant to resume its service. This research explores through a model developed using a finite element computer simulation platform the thermal behaviour of slender tubes under constant temperature exceeding 723 °K. Our simulation results demonstrate that hematite layers up to 15 μm thickness inside the tubes do not act as insulation. They clearly show the process of long term overheating on the outside of boiler tubes which in turn leads to initiation of flaws.

Comparison of three flaw-location methods for automated ultrasonic testing

International Nuclear Information System (INIS)

Seiger, H.

1982-01-01

Two well-known methods for locating flaws by measurement of the transit time of ultrasonic pulses are examined theoretically. It is shown that neither is sufficiently reliable for use in automated ultrasonic testing. A third method, which takes into account the shape of the sound field from the probe and the uncertainty in measurement of probe-flaw distance and probe position, is introduced. An experimental comparison of the three methods indicates that use of the advanced method results in more accurate location of flaws. (author)

Effect of combined loading on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Chung Yeonki

1999-01-01

Considering a rational maintenance rule of Light Water Reactor piping, reliable flaw evaluation criteria are essential to determine how a detected flaw is detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes to be considered for carbon steel piping, and can be analyzed by the elastic-plastic fracture mechanics. Currently the analytical results are provided as flaw evaluation criteria using load correction factors such like the Z-factor in ASME Code Section 6. The present correction factors were conventionally determined taken a conservatism and a simplicity into account, however, the effect of internal pressure which would be an important factor under an actual plant condition was not adequately considered. Recently, a J-estimation scheme, 'LBB.ENGC' for ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was newly developed to have a better prediction with more realistic manner. This method is explicitly incorporated the contribution of both bending and tension due to internal pressure by means of the scheme compatible with an arbitrary combined loading history. In this paper, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. A correction factor based on the new J-estimation scheme was compared with the present correction factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of internal pressure. (author)

Evaluation of Effect by Internal Flow on Ultrasonic Testing Flaw Sizing in Piping

International Nuclear Information System (INIS)

Lee, Jeong Seok; Yoon, Byung Sik; Kim, Yong Sik

2013-01-01

In this study, the ultrasonic amplitude difference between air filled and water filled piping in nuclear power plant is compared by modeling approach. In this study, ultrasonic amplitude differences between air and water filled pipe are evaluated by modeling approach. Consequently, we propose the following results. The ultrasonic amplitude difference between air and water filled condition is measured by lower than 1 dB in modeling calculation. The flaw length sizing error between air and water filled condition shows same results based on 12 dB drop method even thought the amplitude difference is 1 dB. Most of the piping welds in nuclear power plants are inspected periodically using ultrasonic techniques to detect service-induced flaws such as IGSCC cracking. The inspection results provide information such as location, maximum amplitude response, ultrasonic length, height and finally the nature or flaw pattern. The founded flaw in ultrasonic inspection is accepted or rejected based on these information. Specially, the amplitude of flaw response is very important to estimate the flaw size. Currently the ultrasonic inspections in nuclear power plant components are performed by specific inspection procedure which describing inspection technique include inspection system, calibration methodology and flaw characterizing methodology. To perform ultrasonic inspection during in-service inspection, reference gain should be established before starting ultrasonic inspection by requirement of ASME code. This reference gain used as basic criteria to evaluate flaw sizing. Sometimes, a little difference in establishing reference gain between calibration and field condition can lead to deviation in flaw sizing. Due to this difference, the inspection result may cause flaw sizing error

Regression analysis of pulsed eddy current signals for inspection of steam generator tube support structures

International Nuclear Information System (INIS)

Buck, J.; Underhill, P.R.; Mokros, S.G.; Morelli, J.; Krause, T.W.; Babbar, V.K.; Lepine, B.

2015-01-01

Nuclear steam generator (SG) support structure degradation and fouling can result in damage to SG tubes and loss of SG efficiency. Conventional eddy current technology is extensively used to detect cracks, frets at supports and other flaws, but has limited capabilities in the presence of multiple degradation modes or fouling. Pulsed eddy current (PEC) combined with principal components analysis (PCA) and multiple linear regression models was examined for the inspection of support structure degradation and SG tube off-centering with the goal of extending results to include additional degradation modes. (author)

An evaluation of the statistical variability in thermal expansion properties of steam generator tubesheet (SA-508) and tubing (Alloy-600TT)

International Nuclear Information System (INIS)

Riccardella, P.C.; Staples, J.F.; Kandra, J.T.

2009-01-01

Inspections of steam generator tubing are performed in U.S. PWRs as part of the Steam Generator Management Program. Westinghouse has recently completed a technical justification demonstrating that in steam generators with thermally treated Ni-Cr Alloy (Alloy 600TT) tubes that are hydraulically expanded into low alloy steel (SA-508) tubesheets, flaws in the region of the tubes below a certain distance from the top of the tubesheet, denoted H * , will not result in reactor coolant pressure boundary breach nor unacceptable primary-to-secondary leakage. This is because, even if a flaw in this region were to result in complete tube sever, if the length of undegraded tube in the tubesheet exceeds H*, neither operating nor accident loadings create sufficient pull-out forces to overcome the frictional forces between the tube and tubesheet. One key component of this technical justification is the differential thermal expansion between the tube and tubesheet, since a significant portion of the pullout strength of the hydraulically expanded tube-to-tubesheet joint is due to mechanical interference resulting from the larger expansion of the tubing relative to the tubesheet at a given temperature. To address this phenomenon, a detailed statistical evaluation of coefficient of thermal expansion (CTE) data for the tubesheet material (SA-508) and the tube material (thermally treated Alloy-600) was performed. Data used in the evaluation included existing test results obtained from a number of sources as well as extensive new laboratory data developed specifically for this purpose. The evaluation resulted in recommended statistical distributions of this property for the two materials including their means and probabilistic variability. In addition, it was determined that the CTE values reported in the ASME Code (Section II) represent reasonably conservative mean values for both the tubesheet and tubing material. (author)

Regulation No. 0-31 on handling of radiation flaw-detectors

International Nuclear Information System (INIS)

1975-01-01

The regulation contains mandatory design, commissioning, and operational requirements for laboratories using flaw-detectors emitting ionizing radiation; also, design, manufacturing, and operational requirements for the production of any type of X-ray or gamma-ray flaw-detectors. Laboratories carrying out non-destructive testing are either stationary or mobile. Conceptual and operating designs are elaborated, including the building and the laboratory lay-outs, the mains, water supply, and sewerage system technological lay-out, explanatory comments, and a lay-out of the shielding equipment. Approbated designs are implemented, and the laboratories commissioned to representatives of the State Sanitary Inspectorate. Licences are issued by the Ministry of Public Health (MPH) and the Committee on Peaceful Uses of Atomic Energy (CPUAE). Any flaw-detector has to conform to the Bulgarian State Standards and be coordinated with the MPH, the CPUAE, and the Central Laboratory for Nuclear Flaw-Detection (CLNFD). The laboratories are required to have operational instructions, an emergency plan, and to keep technological and dosimetric records. The latter are provided and processed by the relevant service at the Research Institute of Radiobiology and Radiation Hygiene. For operations involving of flaw-detectors, presence of at least two workers is required. (G.G.)

R&D of CuCrZr tubes for W/Cu monoblock components

Energy Technology Data Exchange (ETDEWEB)

Zhao, Sixiang, E-mail: sxzhao@impcas.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), P.O. Box 1126, Hefei 230031 (China); Ma, Linsheng [State Nuclear Bao Ti Zirconium Industry Company, 206 Hi-Tech Avenue, Baoji 721013 (China); Peng, Lingjian [Advanced Technology & Materials Co., Ltd. - AT& M, Beijing 100081 (China); Gao, Bo [State Nuclear Bao Ti Zirconium Industry Company, 206 Hi-Tech Avenue, Baoji 721013 (China); Li, Chun [Laboratory of Advanced materials, School of Materials Science & Engineering, Tsinghua University, Beijing 100084 (China); Li, Qiang; Wang, Wanjing; Wei, Ran; Xu, Yuping [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), P.O. Box 1126, Hefei 230031 (China); Pan, Ningjie; Qin, Sigui; Shi, Yingli; Liu, Guohui; Wang, Tiejun [Advanced Technology & Materials Co., Ltd. - AT& M, Beijing 100081 (China); Luo, Guang-Nan, E-mail: gnluo@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP), P.O. Box 1126, Hefei 230031 (China); Hefei Center for Physical Science and Technology, Hefei 230031 (China); Hefei Science Center of CAS, Hefei 230031 (China)

2016-11-15

Highlights: • CuCrZr tubes with excellent HIP performance and good resistance to grain growth have been developed. • A circumferential ductility testing manner for small-diameter tubes has been utilized in this study. • The evolution of microstructures has been revealed throughout the new tube forming processes. - Abstract: In order to avoid the occurrence of two types of longitudinal defects (strain localization and folding flaws), which were observed in the CuCrZr tubes of EAST W/Cu upper divertor components, in the future manufacturing of monoblock components using hot isostatic pressing (HIP), a new CuCrZr tube forming protocol is proposed. The evolution of Cu grains and Cr-rich particles is monitored by scanning electron microscopy throughout the new tube forming processes. The final microstructures of the newly developed tubes are totally different from those of the EAST project previously chosen tubes and the elongation of Cr-rich precipitates has been substantially suppressed by using the new tube forming protocol. The newly developed tubes show better HIP performance than the EAST previously chosen ones. Since circumferential mechanical properties, especially ductility, are of great importance, a circumferential ductility testing manner for small-diameter tubes, which might be a supplement to longitudinal tensile testing, has been utilized and the preliminary testing results are given. The recrystallization behavior of the newly developed tubes is also investigated.

A review on the conservatism embodied in the integrity assessment procedure of CANDU pressure tubes

International Nuclear Information System (INIS)

Chung, Han Sub; Kim, Hong Duk

2009-01-01

Excessive conservatism may be embodied in the integrity assessment procedure for PTs of CANDUs, as guided in the CSA N285.8-05 code. Four sources of conservatism are suggested as (1) the effect of thermal history on DHC, (2) the effect of manufacturing variables on properties of PTs, (3) poor definition or limited availability of databases regarding the aging degradation trends and probability density functions of important material properties, and (4) the simplified equations defining threshold condition for the initiation of DHC at the volumetric flaws. It is suggested that both extensive review of databases available and extensive calculation case studies are required to remove the potential excessive conservatism.

The PISC programme on defective steam generator tubes inspection. A status report

International Nuclear Information System (INIS)

Birac, C.; Comby, R.; Maciga, G.; Von Estorff, U.; Zanella, G.L.

1994-06-01

The general objective of the PISC Program (Programme for the Inspection of Steel Components) is to assess experimentally procedures and techniques in use for the in-service inspection of pressure components. The program is mainly a round robin test, the results of which are compared with real characteristics of the flaws obtained by destructive analysis. Materials tested are INCONEL 600 tubes, diameter 22.22 mm, wall thickness 1.27 mm. The technique applied is eddy current testing. The program of capability tests on loose tubes was started in 1990, the round robin tests ended in 1993. The preliminary results are presented. (R.P.). 8 refs., 9 figs., 4 tabs

Ultrasonic imaging of material flaws exploiting multipath information

Science.gov (United States)

Shen, Xizhong; Zhang, Yimin D.; Demirli, Ramazan; Amin, Moeness G.

2011-05-01

In this paper, we consider ultrasonic imaging for the visualization of flaws in a material. Ultrasonic imaging is a powerful nondestructive testing (NDT) tool which assesses material conditions via the detection, localization, and classification of flaws inside a structure. Multipath exploitations provide extended virtual array apertures and, in turn, enhance imaging capability beyond the limitation of traditional multisensor approaches. We utilize reflections of ultrasonic signals which occur when encountering different media and interior discontinuities. The waveforms observed at the physical as well as virtual sensors yield additional measurements corresponding to different aspect angles. Exploitation of multipath information addresses unique issues observed in ultrasonic imaging. (1) Utilization of physical and virtual sensors significantly extends the array aperture for image enhancement. (2) Multipath signals extend the angle of view of the narrow beamwidth of the ultrasound transducers, allowing improved visibility and array design flexibility. (3) Ultrasonic signals experience difficulty in penetrating a flaw, thus the aspect angle of the observation is limited unless access to other sides is available. The significant extension of the aperture makes it possible to yield flaw observation from multiple aspect angles. We show that data fusion of physical and virtual sensor data significantly improves the detection and localization performance. The effectiveness of the proposed multipath exploitation approach is demonstrated through experimental studies.

Improved flaw detection and characterization with difference thermography

Science.gov (United States)

Winfree, William P.; Zalameda, Joseph N.; Howell, Patricia A.

2011-05-01

Flaw detection and characterization with thermographic techniques in graphite polymer composites is often limited by localized variations in the thermographic response. Variations in properties such as acceptable porosity, variations in fiber volume content and surface polymer thickness result in variations in the thermal response that in general cause significant variations in the initial thermal response. These variations result in a noise floor that increases the difficulty of detecting and characterizing deeper flaws. The paper investigates comparing thermographic responses taken before and after a change in state in a composite to improve the detection of subsurface flaws. A method is presented for registration of the responses before finding the difference. A significant improvement in the detectability is achieved by comparing the differences in response. Examples of changes in state due to application of a load and impact are presented.

Development of a diagnostic expert system for eddy current data analysis using applied artificial intelligence methods

International Nuclear Information System (INIS)

Upadhyaya, B.R.; Yan, W.; Henry, G.

1999-01-01

A diagnostic expert system that integrates database management methods, artificial neural networks, and decision-making using fuzzy logic has been developed for the automation of steam generator eddy current test (ECT) data analysis. The new system, known as EDDYAI, considers the following key issues: (1) digital eddy current test data calibration, compression, and representation; (2) development of robust neural networks with low probability of misclassification for flaw depth estimation; (3) flaw detection using fuzzy logic; (4) development of an expert system for database management, compilation of a trained neural network library, and a decision module; and (5) evaluation of the integrated approach using eddy current data. The implementation to field test data includes the selection of proper feature vectors for ECT data analysis, development of a methodology for large eddy current database management, artificial neural networks for flaw depth estimation, and a fuzzy logic decision algorithm for flaw detection. A large eddy current inspection database from the Electric Power Research Institute NDE Center is being utilized in this research towards the development of an expert system for steam generator tube diagnosis. The integration of ECT data pre-processing as part of the data management, fuzzy logic flaw detection technique, and tube defect parameter estimation using artificial neural networks are the fundamental contributions of this research. (orig.)

Development of a diagnostic expert system for eddy current data analysis using applied artificial intelligence methods

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, B.R.; Yan, W. [Tennessee Univ., Knoxville, TN (United States). Dept. of Nuclear Engineering; Behravesh, M.M. [Electric Power Research Institute, Palo Alto, CA (United States); Henry, G. [EPRI NDE Center, Charlotte, NC (United States)

1999-09-01

A diagnostic expert system that integrates database management methods, artificial neural networks, and decision-making using fuzzy logic has been developed for the automation of steam generator eddy current test (ECT) data analysis. The new system, known as EDDYAI, considers the following key issues: (1) digital eddy current test data calibration, compression, and representation; (2) development of robust neural networks with low probability of misclassification for flaw depth estimation; (3) flaw detection using fuzzy logic; (4) development of an expert system for database management, compilation of a trained neural network library, and a decision module; and (5) evaluation of the integrated approach using eddy current data. The implementation to field test data includes the selection of proper feature vectors for ECT data analysis, development of a methodology for large eddy current database management, artificial neural networks for flaw depth estimation, and a fuzzy logic decision algorithm for flaw detection. A large eddy current inspection database from the Electric Power Research Institute NDE Center is being utilized in this research towards the development of an expert system for steam generator tube diagnosis. The integration of ECT data pre-processing as part of the data management, fuzzy logic flaw detection technique, and tube defect parameter estimation using artificial neural networks are the fundamental contributions of this research. (orig.)

Performance demonstration tests for eddy current inspection of steam generator tubing

International Nuclear Information System (INIS)

Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

1996-05-01

This report describes the methodology and results for development of performance demonstration tests for eddy current (ET) inspection of steam generator tubes. Statistical test design principles were used to develop the performance demonstration tests. Thresholds on ET system inspection performance were selected to ensure that field inspection systems would have a high probability of detecting and and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented in detail. Statistical test design calculations for probability of detection and flaw sizing tests are described. A recommended performance demonstration test based on the design calculations is presented. A computer program for grading the probability of detection portion of the performance demonstration test is given

Performance demonstration tests for eddy current inspection of steam generator tubing

Energy Technology Data Exchange (ETDEWEB)

Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

1996-05-01

This report describes the methodology and results for development of performance demonstration tests for eddy current (ET) inspection of steam generator tubes. Statistical test design principles were used to develop the performance demonstration tests. Thresholds on ET system inspection performance were selected to ensure that field inspection systems would have a high probability of detecting and and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented in detail. Statistical test design calculations for probability of detection and flaw sizing tests are described. A recommended performance demonstration test based on the design calculations is presented. A computer program for grading the probability of detection portion of the performance demonstration test is given.

Flaw-size measurement in a weld samples by ultrasonic frequency analysis

International Nuclear Information System (INIS)

Adler, L.; Cook, K.V.; Whaley, H.L. Jr.; McClung, R.W.

1975-01-01

An ultrasonic frequency-analysis technique was developed and applies to characterize flaws in an 8-in. (203-mm) thick heavy-section steel weld specimen. The technique applies a multitransducer system. The spectrum of the received broad-band signal is frequency analyzed at two different receivers for each of the flaws. From the two spectra, the size and orientation of the flaw are determined by the use of an analytic model proposed earlier. (auth)

A methodology for determining fabrication flaws in a reactor pressure vessel

International Nuclear Information System (INIS)

Schuster, G.J.; Doctor, S.R.; Simonen, F.A.

1996-01-01

The Pacific Northwest National Laboratory (PNNL) conducted a program with the major objective of estimating the rate of occurrence of fabrication flaws in US light-water reactor pressure vessels (RPVs). In this study, RPV mate4rial was examined using the Synthetic Aperture Focusing Technique for Ultrasonic Testing (SAFT-UT) to detect and characterize flaws created during fabrication. The inspection data obtained in this program has been analyzed to address the rates of flaw occurrence

Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

International Nuclear Information System (INIS)

Joo, Kyung Mun; Hong, Jun Hee

2015-01-01

Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Joo, Kyung Mun [Korea Hydro and Nuclear Power Company Ltd., Central Research Institute, Daejeon (Korea, Republic of); Hong, Jun Hee [Dept. of mechanical Engineering, Chungnam National University, Daejeon (Korea, Republic of)

2015-02-15

Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

Development and performance of inspection equipment for pressure tubes in Fugen

International Nuclear Information System (INIS)

Naruo, Kazuteru; Tanimoto, Ken-ichi; Ohta, Takeo; Nakamura, Takahisa; Imaizumi, Kiyoshi.

1984-01-01

The pressure tubes of Fugen are the important equipment as the many tubes compose the core, and since they are made of Zr-2.5% Nb alloy which has been used for the first time in Japan, they have become the object of monitoring (the follow-up investigation of the change of inside diameter, the presence of defects and so on) in addition to the in-service inspection. In this paper, on the inspection equipment for pressure tubes, that has been developed independently by the Power Reactor and Nuclear Fuel Development Corp. in order to carry out the ISI and monitoring, the course of development and the construction and the performance are reported, and the results of having used it for the fourth regular inspection of Fugen are described. The 10-year plan of the ISI and monitoring of pressure tubes is shown. The core of Fugen is composed of 224 pressure tubes, therefore, the inspection is carried out by sampling inspection. The monitoring is carried out on four tubes for the follow-up investigation and one tube that shows the severest operation history at the time of inspection. The equipment performs ultrasonic flaw detection, the measurement of inside diameter and the visual inspection of internal surface. (Kako, I.)

An investigation on compression strength analysis of commercial aluminium tube to aluminium 2025 tube plate by using TIG welding process

Energy Technology Data Exchange (ETDEWEB)

Kannan, S., E-mail: kannan.dgl201127@gmail.com [Department of Mechanical Engineering and Mining Machinery Engineering, Indian Institute of Technology (ISM), Dhanbad, Jharkhand, India, 826004 (India); Senthil Kumaran, S., E-mail: sskumaran@ymail.com [Research and Development Center, Department of Mechanical Engineering, RVS Educational Trust' s Group of Institutions, RVS School of Engineering and Technology, Dindigul, Tamilnadu, India, 624005 (India); Kumaraswamidhas, L.A., E-mail: lakdhas1978@gmail.com [Department of Mechanical Engineering and Mining Machinery Engineering, Indian School of Mines University, Dhanbad, Jharkhand, India, 826004 (India)

2016-05-05

In this present study, Tungsten inert gas (TIG) welding was applied to weld the dissimilar materials and authenticate the mechanical and metallurgical properties of tube to tube plate made up of commercial aluminium and Al 2025 respectively using an Zirconiated tungsten electrode along with filler material aluminium ER 2219. In total, twenty five pieces has been subjected to compression strength and hardness value to evaluate the optimal joint strength. The three optimization technique has been used in this experiment. Taguchi L{sub 25} orthogonal array is used to identify the most influencing process parameter which affects the joint strength. ANOVA method is measured for both compression strength and hardness to calculate the percentage of contribution for each process parameter. Genetic algorithm is used to validate the results obtained from the both experimental value and optimization value. The micro structural study is depicted the welding joints characterization in between tube to tube plate joints. The radiograph test is conducted to prove the welds are non-defective and no flaws are found during the welding process. The mechanical property of compression strength and hardness has been measured to obtain the optimal joint strength of the welded sample was about 174.846 MPa and 131.364 Hv respectively. - Highlights: • Commercial Al tube and Al 2025 tube plate successfully welded by TIG welding. • Compression strength and hardness value proves to obtain optimal joint strength. • The maximum compression and hardness was achieved in various input parameters.

An investigation on compression strength analysis of commercial aluminium tube to aluminium 2025 tube plate by using TIG welding process

International Nuclear Information System (INIS)

Kannan, S.; Senthil Kumaran, S.; Kumaraswamidhas, L.A.

2016-01-01

In this present study, Tungsten inert gas (TIG) welding was applied to weld the dissimilar materials and authenticate the mechanical and metallurgical properties of tube to tube plate made up of commercial aluminium and Al 2025 respectively using an Zirconiated tungsten electrode along with filler material aluminium ER 2219. In total, twenty five pieces has been subjected to compression strength and hardness value to evaluate the optimal joint strength. The three optimization technique has been used in this experiment. Taguchi L_2_5 orthogonal array is used to identify the most influencing process parameter which affects the joint strength. ANOVA method is measured for both compression strength and hardness to calculate the percentage of contribution for each process parameter. Genetic algorithm is used to validate the results obtained from the both experimental value and optimization value. The micro structural study is depicted the welding joints characterization in between tube to tube plate joints. The radiograph test is conducted to prove the welds are non-defective and no flaws are found during the welding process. The mechanical property of compression strength and hardness has been measured to obtain the optimal joint strength of the welded sample was about 174.846 MPa and 131.364 Hv respectively. - Highlights: • Commercial Al tube and Al 2025 tube plate successfully welded by TIG welding. • Compression strength and hardness value proves to obtain optimal joint strength. • The maximum compression and hardness was achieved in various input parameters.

Integrated function nonimaging concentrating collector tubes for solar thermal energy

Science.gov (United States)

Winston, R.; Ogallagher, J. J.

1982-09-01

A substantial improvement in optical efficiency over contemporary external reflector evacuated tube collectors has been achieved by integrating the reflector surface into the outer glass envelope. Described are the design fabrication and test results for a prototype collector based on this concept. A comprehensive test program to measure performance and operational characteristics of a 2 sq m panel (45 tubes) has been completed. Efficiencies above 50% relative to beam at 200 C have been repeatedly demonstrated. Both the instantaneous and long term average performance of this totally stationary solar collector are comparable to those for tracking line focus parabolic troughs. The yield, reliability and stability of performance achieved have been excellent. Subcomponent assemblies and fabrication procedures have been used which are expected to be compatible with high volume production. The collector has a wide variety of applications in the 100 to 300 C range including industrial progress heat, air conditioning and Rankine engine operation.

Magnetite Core-Shell Nanoparticles in Nondestructive Flaw Detection of Polymeric Materials.

Science.gov (United States)

Hetti, Mimi; Wei, Qiang; Pohl, Rainer; Casperson, Ralf; Bartusch, Matthias; Neu, Volker; Pospiech, Doris; Voit, Brigitte

2016-10-04

Nondestructive flaw detection in polymeric materials is important but difficult to achieve. In this research, the application of magnetite nanoparticles (MNPs) in nondestructive flaw detection is studied and realized, to the best of our knowledge, for the first time. Superparamagnetic and highly magnetic (up to 63 emu/g) magnetite core-shell nanoparticles are prepared by grafting bromo-end-group-functionalized poly(glycidyl methacrylate) (Br-PGMA) onto surface-modified Fe 3 O 4 NPs. These Fe 3 O 4 -PGMA NPs are blended into bisphenol A diglycidylether (BADGE)-based epoxy to form homogeneously distributed magnetic epoxy nanocomposites (MENCs) after curing. The core Fe 3 O 4 of the Fe 3 O 4 -PGMA NPs endows the MENCs with magnetic property, which is crucial for nondestructive flaw detection of the materials, while the shell PGMA promotes colloidal stability and prevents NP aggregation during curing. The eddy current testing (ET) technique is first applied to detect flaws in the MENCs. Through the brightness contrast of the ET image, surficial and subsurficial flaws in MENCs can be detected, even for MENCs with low content of Fe 3 O 4 -PGMA NPs (1 wt %). The incorporation of Fe 3 O 4 -PGMA NPs can be easily extended to other polymer and polymer-based composite systems and opens a new and very promising pathway toward MNP-based nondestructive flaw detection in polymeric materials.

A study on the measurement of flaw sizes by acoustical holography

International Nuclear Information System (INIS)

Yamamoto, M.; Ando, T.; Enami, K.; Yajima, M.; Fukui, S.

1978-01-01

As a means of evaluating the safety of flawed pressure vessels and other structures against fracture, fracture mechanics has come to be applied. For the application of fracture mechanics it is necessary to get information concerning the sizes and shapes of flaws. The ultrasonic flaw detection method which is widely used as a nondestructive inspection method cannot measure the sizes and shapes of flaws accurately. Considering that acoustical holography is an useful means for the measurement of flaws, we performed basic tests on this method and obtained the following results: (1) The measured values of artificial flaws (flat bottom drilled holes: 5 -- 36 mm) made on a steel plate with a thickness of 150 mm showed a good linear relation with their actual sizes and scatter in the measured values was +-3 -- 6 mm. (2) The measured values of fatigue cracks (length: 5 -- 57 mm) introduced into a steel plate with thickness of 150 mm also showed a good linear relation with their actual sizes and scatter in the measured values was +-3 mm. (3) It was found that acoustical holography can also be applied to heavy section cast steels. (4) The method of correcting distortion caused by curved surface was investigated by computer-aided simulation and it was considered that such distortion can be corrected by radial scanning of a transducer. (auth.)

Ultrasonic Transducer Design for the Axial Flaw Detection of Dissimilar Metal Weld

International Nuclear Information System (INIS)

Yoon, Byung Sik; Kim, Yong Sik; Yang, Seung Han

2011-01-01

Dissimilar metal welds in nuclear power plant are known as very susceptible to PWSCC flaws, and periodically inspected by the qualified inspector and qualified procedure during in-service inspection period. According to field survey data, the majority of their DMWs are located on tapered nozzle or adjacent to a tapered component. These types of configurations restrict examination access and also limit examination volume coverage. Additionally, circumferential scan for axially oriented flaw is very difficult to detect located on tapered surface because the transducer can't receive flaw response from reflector for miss-orientation. To overcome this miss-orientation, it is necessary adapt skewed ultrasonic transducer accommodate tapered surface. The skewed refracted longitudinal ultrasonic transducer designed by modeling and manufactured from the modelling result for axial flaw detection. Experimental results showed that the skewed refracted longitudinal ultrasonic transducer get higher flaw response than non-skewed refracted longitudinal ultrasonic transducer

High-temperature flaw assessment procedure

International Nuclear Information System (INIS)

Ruggles, M.B.; Takahashi, Y.; Ainsworth, R.A.

1991-08-01

Described is the background work performed jointly by the Electric Power Research Institute in the United States, the Central Research Institute of Electric Power Industry in Japan and Nuclear Electric plc in the United Kingdom with the purpose of developing a high-temperature flaw assessment procedure for reactor components. Existing creep-fatigue crack-growth models are reviewed, and the most promising methods are identified. Sources of material data are outlined, and results of the fundamental deformation and crack-growth tests are discussed. Results of subcritical crack-growth exploratory tests, creep-fatigue crack-growth tests under repeated thermal transient conditions, and exploratory failure tests are presented and contrasted with the analytical modeling. Crack-growth assessment methods are presented and applied to a typical liquid-metal reactor component. The research activities presented herein served as a foundation for the Flaw Assessment Guide for High-Temperature Reactor Components Subjected to Creep-Fatigue Loading published separately. 30 refs., 108 figs., 13 tabs

Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

Energy Technology Data Exchange (ETDEWEB)

Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E. [Oak Ridge National Lab., Heavy-Section Steel Technology Program, Oak Ridge, TN (United States)

2001-07-01

A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

International Nuclear Information System (INIS)

Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E.

2001-01-01

A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

International Nuclear Information System (INIS)

Langman, V.J.; Weaver, K.R.

1984-05-01

The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

Stress intensities in flawed pressure vessels

International Nuclear Information System (INIS)

Smith, C.W.; Jolles, M.; Peters, W.H.

1977-01-01

A technique for determining the stess intensity factor (SIF) near pressure vessel flaws or cracks experimentally from photoelastic data for use in two-dimensional problems was developed in the 1950's. This technique was modified and extended to a variety of two-dimensional problems. The technique has been refined further and what has evolved may be regarded as a hybrid technique which affects a marriage between ''frozen stress'' photoelastic results and a simple least-squares digital computer program for estimating SIF values in three-dimensional problems. This technique, in its original modified form, has been shown to be applicable to a study of surface flaws and the applicability of the method to complex crack body geometries of current technological importance are discussed. The analytical foundations of the method are reviewed

Potential change in flaw geometry of an initially shallow finite-length surface flaw during a pressurized-thermal-shock transient

International Nuclear Information System (INIS)

Shum, D.K.; Bryson, J.W.; Merkle, J.G.

1993-09-01

This study presents preliminary estimates on whether an shallow, axially oriented, inner-surface finite-length flaw in a PWR-RPV would tend to elongate in the axial direction and/or deepen into the wall of the vessel during a postulated PTS transient. Analysis results obtained based on the assumptions of (1) linear-elastic material response, and (2) cladding with same toughness as the base metal, indicate that a nearly semicircular flaw would likely propagate in the axial direction followed by propagation into the wall of the vessel. Note that these results correspond to initiation within the lower-shelf fracture toughness temperature range, and that their general validity within the lower-transition temperature range remains to be determined. The sensitivity of the numerical results aid conclusions to the following analysis assumptions are evaluated: (1) reference flaw geometry along the entire crack front and especially within the cladding region; (2) linear-elastic vs elastic-plastic description of material response; and (3) base-material-only vs bimaterial cladding-base vessel-model assumption. The sensitivity evaluation indicates that the analysis results are very sensitive to the above assumptions

40° image intensifier tubes in an integrated helmet system

Science.gov (United States)

Schreyer, Herbert; Boehm, Hans-Dieter V.; Svedevall, B.

1993-12-01

EUROCOPTER has been under contract to the French and German ministries of defence for five years to develop the TIGER, a second generation anti-tank helicopter. A piloting thermal imager has been installed on a steerable platform in the helicopter nose in order to achieve the possibility of flying round the clock. In addition to this sensor, which is sensitive at a wavelength of 10 micrometers , the German side has proposed using an Integrated Helmet System in the PAH 2. This helmet, manufactured by GEC-Marconi Avionics, incorporates two cathode ray tubes (CRT) and two image intensifier tubes which allow the pilot to use an additional sensor in the visible and near infrared spectrum. The electronic part will be built by Teldix. EUROCOPTER DEUTSCHLAND has received the first demonstrator of this helmet for testing in the EUROCOPTER Visionics Laboratory. Later, the C-prototype will be integrated into a BK 117 helicopter (AVT Avionik Versuchstrager). This new helmet has a field of view of 40 degree(s), and exit pupil of 15 mm and improved possibilities of adjusting the optical part. Laboratory tests have been carried out to test important parameters like optical resolution under low light level conditions, field of view, eye relief or exit pupil. The CRT channels have been tested for resolution, distortion, vignetting and homogeneity. The requirements and the properties of the helmet, test procedures and the results of these tests are presented in the paper.

Advanced Signal Processing for Thermal Flaw Detection; TOPICAL

International Nuclear Information System (INIS)

VALLEY, MICHAEL T.; HANSCHE, BRUCE D.; PAEZ, THOMAS L.; URBINA, ANGEL; ASHBAUGH, DENNIS M.

2001-01-01

Dynamic thermography is a promising technology for inspecting metallic and composite structures used in high-consequence industries. However, the reliability and inspection sensitivity of this technology has historically been limited by the need for extensive operator experience and the use of human judgment and visual acuity to detect flaws in the large volume of infrared image data collected. To overcome these limitations new automated data analysis algorithms and software is needed. The primary objectives of this research effort were to develop a data processing methodology that is tied to the underlying physics, which reduces or removes the data interpretation requirements, and which eliminates the need to look at significant numbers of data frames to determine if a flaw is present. Considering the strengths and weakness of previous research efforts, this research elected to couple both the temporal and spatial attributes of the surface temperature. Of the possible algorithms investigated, the best performing was a radiance weighted root mean square Laplacian metric that included a multiplicative surface effect correction factor and a novel spatio-temporal parametric model for data smoothing. This metric demonstrated the potential for detecting flaws smaller than 0.075 inch in inspection areas on the order of one square foot. Included in this report is the development of a thermal imaging model, a weighted least squares thermal data smoothing algorithm, simulation and experimental flaw detection results, and an overview of the ATAC (Automated Thermal Analysis Code) software that was developed to analyze thermal inspection data

Fracture evaluation of an in-service piping flaw caused by microbiologically induced corrosion

International Nuclear Information System (INIS)

Rudland, D.L.; Scott, P.M.; Wilkowski, G.M.; Rahman, S.

1996-01-01

A pipe fracture experiment was conducted on a section of 6-inch nominal diameter pipe which was degraded by microbiologically induced corrosion (MIC) at a circumferential girth weld. The pipe was a section of one of the service water piping systems to one of the emergency diesel generators at the Haddam Neck (Connecticut Yankee) plant. The experimental results will help validate future ASME Section XI pipe flaw evaluation criteria for other than Class 1 piping. A critical aspect of this experiment was an assessment of the degree of conservatism embodied in the ASME definition of flaw size. The ASME flaw size definition assumes a rectangular shaped, constant depth flaw with a depth equal to its maximum depth for its entire length. Since most service flaws are irregular in shape, this definition may be overly conservative. Results from several fracture prediction models are compared with the experimental results. These results show that, for this case, the ASME Appendix H criteria significantly underpredicted the experimental maximum moment, while other fracture prediction models provided good predictions when accurate pipe, weld and flaw dimensions were used

Cracking and Failure in Rock Specimen Containing Combined Flaw and Hole under Uniaxial Compression

Directory of Open Access Journals (Sweden)

Xiang Fan

2018-01-01

Full Text Available Flaw is a key factor influencing failure behavior of a fractured specimen. In the present study, rectangular-flawed specimens were prepared using sandstone to investigate the effect of flaw on failure behavior of rock. Open flaw and cylindrical hole were simultaneously precut within rock specimens using high-pressure water jet cutting technology. Five series of specimens including intact, single-hole-alone, two-hole-alone, single-hole and two-flaw, and two-hole and single-flaw blocks were prepared. Uniaxial compressive tests using a rigid servo control instrument were carried out to investigate the fracture processes of these flawed specimens. It is observed that during loading, internal stress always intensively distributed at both sidewalls of open hole, especially at midpoint of sidewalls, so rock crumb flaking was firstly observed among all sandstone specimens containing single hole or two holes. Cracking around open hole is associated with the flaw inclination angle which was observed in Series III and V. Crack easily initiated at the tips of flaw with inclination angles of 0°, 30°, and 60° but hard for 90° in Series III and V. Rock burst was the major failure mode among most tested specimens, which generally induced new cracks and finally created crater shape. Additionally, due to extrusion between blocks, new shear or tensile cracks were generated and the rock specimen surface spalled. Eventually, four typical failure processes including rock crumb flaking, crack initiation and propagation, rock burst, and second rupture, were summarized.

The Cause of an Eddy Current Signal Noise from a Steam Generator Tube and its Effect on the Detectability of a Crack

International Nuclear Information System (INIS)

Lee, Deok Hyun; Choi, Myung Sik; Hur, Do Haeng; Kim, Kyung Mo; Han, Jung Ho

2008-01-01

An eddy current inspection has been applied for a pre-service and in-service examination of a steam generator in nuclear power plants. The experience from the inspection of steam generators showed that many plants had an excessive number of tubes with eddy current noise signals over several hundreds, which originated from manufacturing anomalies. The plants in U.S suffered significant downstream inspection costs, history reviews, and diagnostic testing because some signals resembled flaws and others masked a flaw. These lessens learned resulted in issuing the guidelines for steam generator tubing specifications and repair, in order to reduce the number of anomalous signals in the tubes and also to provide the requirement of a signal to noise ratio by applying a field type examination with bobbin coil eddy current probes at a manufacturing process. Besides the noise signals of a bobbin coil eddy current probe from manufacturing anomalies, the excessive background noise of the rotating coil eddy current probe signal is frequently observed from a tube and it negatively affects the detection and sizing estimate of a defect. Since the inspection intervals are being extended up to 60 months for the more recent steam generator of corrosion resistant alloy 690TT tubing, the detection of an earlier crack and an accurate sizing are becoming more important in the activity of a non-destructive examination. In this study, the cause of an eddy current signal noise of a rotating coil probe from a steam generator tube was examined and its influence on the detectability of a crack was analyzed

Application of elastic and elastic-plastic fracture mechanics methods to surface flaws

Science.gov (United States)

McCabe, Donald E.; Ernst, Hugo A.; Newman, James C., Jr.

Fuel tanks that are a part of the External Tank assembly for the Space Shuttle are made of relatively thin 2219-T87 aluminum plate. These tanks contain about 917 m of fusion weld seam, all of which is nondestructively inspected for flaws and all those found are repaired. The tanks are subsequently proof-tested to a pressure that is sufficiently severe to cause weld metal yielding in a few local regions of the weld seam. The work undertaken in the present project was to develop a capability to predict flaw growth from undetected surface flaws that are assumed to be located in the highly stressed regions. The technical challenge was to develop R-curve prediction capability for surface cracks in specimens that contain the flaws of unusual sizes and shapes deemed to be of interest. The test techniques developed and the elastic-plastic analysis concepts adopted are presented. The flaws of interest were quite small surface cracks that were narrow-deep ellipses that served to exacerbate the technical difficulties involved.

Design flaw could delay collider

CERN Multimedia

Cho, Adrian

2007-01-01

"A magnet for the Large Hadron Collider (LHC) failed during a key test at the European particle physics laboratory CERN last week. Physicists and engineers will have to repair the damaged magnet and retrofit others to correct the underlynig design flaw, which could delay the start-up of the mammouth subterranean machine." (1,5 page)

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

International Nuclear Information System (INIS)

Zahoor, A.; Ghassemi, B.B.

1991-01-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.)

Tearing stability analysis of an axial surface flaw in thick-walled pressure vessels

Energy Technology Data Exchange (ETDEWEB)

Zahoor, A.; Ghassemi, B.B. (NOVETECH Corp., Rockville, MD (USA))

1991-04-01

This paper presents two fracture mechanics models for evaluation of an axial surface flaw in pressure vessels. The surface flaw is located on the outside surface of the vessel. The first model assumes yielding of the remaining ligament directly ahead of the flaw. The second model assumes contained yielding ahead of the flaw and uses a linear elastic fracture mechanics solution. The former model is suitable for cases where the combination of material toughness, flaw size, and load is such that initiation of flaw growth follows ligament yielding. The latter model is suitable for low-toughness materials where initiation of crack growth and potential tearing instability may occur prior to the yielding of the ligament. Both models are suitable for thick-walled vessels. The paper discusses the applicability regime for both models. The models are then applied to a test vessel and the predicted failure pressure is compared against the pressure attained in the test. Results show that both models can be applied successfully. In particular, the contained yielding model when used with the plane-stress assumption can give reasonable predictions even for cases that involve yielding of the ligament. (orig.).

Ultrasonographic Detection of Tooth Flaws

Science.gov (United States)

Bertoncini, C. A.; Hinders, M. K.; Ghorayeb, S. R.

2010-02-01

The goal of our work is to adapt pulse-echo ultrasound into a high resolution imaging modality for early detection of oral diseases and for monitoring treatment outcome. In this talk we discuss our preliminary results in the detection of: demineralization of the enamel and dentin, demineralization or caries under and around existing restorations, caries on occlusal and interproximal surfaces, cracks of enamel and dentin, calculus, and periapical lesions. In vitro immersion tank experiments are compared to results from a handpiece which uses a compliant delay line to couple the ultrasound to the tooth surface. Because the waveform echoes are complex, and in order to make clinical interpretation of ultrasonic waveform data in real time, it is necessary to automatically interpret the signals. We apply the dynamic wavelet fingerprint algorithms to identify and delineate echographic features that correspond to the flaws of interest in teeth. The resulting features show a clear distinction between flawed and unflawed waveforms collected with an ultrasonic handpiece on both phantom and human cadaver teeth.

Probabilistic assessment of critically flawed LMFBR PHTS piping elbows

International Nuclear Information System (INIS)

Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.

1982-01-01

One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results

Evaluation of sampling plans for in-service inspection of steam generator tubes

International Nuclear Information System (INIS)

Kurtz, R.J.; Heasler, P.G.; Baird, D.B.

1994-02-01

This report summarizes the results of three previous studies to evaluate and compare the effectiveness of sampling plans for steam generator tube inspections. An analytical evaluation and Monte Carlo simulation techniques were the methods used to evaluate sampling plan performance. To test the performance of candidate sampling plans under a variety of conditions, ranges of inspection system reliability were considered along with different distributions of tube degradation. Results from the eddy current reliability studies performed with the retired-from-service Surry 2A steam generator were utilized to guide the selection of appropriate probability of detection and flaw sizing models for use in the analysis. Different distributions of tube degradation were selected to span the range of conditions that might exist in operating steam generators. The principal means of evaluating sampling performance was to determine the effectiveness of the sampling plan for detecting and plugging defective tubes. A summary of key results from the eddy current reliability studies is presented. The analytical and Monte Carlo simulation analyses are discussed along with a synopsis of key results and conclusions

Fundamentally Flawed: Extension Administrative Practice (Part 1).

Science.gov (United States)

Patterson, Thomas F., Jr.

1997-01-01

Extension's current administrative techniques are based on the assumptions of classical management from the early 20th century. They are fundamentally flawed and inappropriate for the contemporary workplace. (SK)

Thermal performance of capillary micro tubes integrated into the sandwich element made of concrete

DEFF Research Database (Denmark)

Mikeska, Tomas; Svendsen, Svend

2013-01-01

integrated into the thin plate of sandwich element made of HPC can supply the energy needed for heating and cooling. The investigations were conceived as a low temperature concept, where the difference between the temperature of circulating fluid and air in the room was kept in range of 1 to 4°C. © (2013......The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of High Performance Concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CMT...

Automation in tube finishing bay

International Nuclear Information System (INIS)

Bhatnagar, Prateek; Satyadev, B.; Raghuraman, S.; Syama Sundara Rao, B.

1997-01-01

Automation concept in tube finishing bay, introduced after the final pass annealing of PHWR tubes resulted in integration of number of sub-systems in synchronisation with each other to produce final cut fuel tubes of specified length, tube finish etc. The tube finishing bay which was physically segregated into four distinct areas: 1. tube spreader and stacking area, 2. I.D. sand blasting area, 3. end conditioning, wad blowing, end capping and O.D. wet grinding area, 4. tube inspection, tube cutting and stacking area has been studied

Statistical prediction of AVB wear growth and initiation in model F steam generator tubes using Monte Carlo method

International Nuclear Information System (INIS)

Lee, Jae Bong; Park, Jae Hak; Kim, Hong Deok; Chung, Han Sub; Kim, Tae Ryong

2005-01-01

The growth of AVB wear in Model F steam generator tubes is predicted using the Monte Carlo Method and statistical approaches. The statistical parameters that represent the characteristics of wear growth and wear initiation are derived from In-Service Inspection (ISI) Non-Destructive Evaluation (NDE) data. Based on the statistical approaches, wear growth model are proposed and applied to predict wear distribution at the End Of Cycle (EOC). Probabilistic distributions of the number of wear flaws and maximum wear depth at EOC are obtained from the analysis. Comparing the predicted EOC wear flaw data with the known EOC data the usefulness of the proposed method is examined and satisfactory results are obtained

Statistical prediction of AVB wear growth and initiation in model F steam generator tubes using Monte Carlo method

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Bong; Park, Jae Hak [Chungbuk National Univ., Cheongju (Korea, Republic of); Kim, Hong Deok; Chung, Han Sub; Kim, Tae Ryong [Korea Electtric Power Research Institute, Daejeon (Korea, Republic of)

2005-07-01

The growth of AVB wear in Model F steam generator tubes is predicted using the Monte Carlo Method and statistical approaches. The statistical parameters that represent the characteristics of wear growth and wear initiation are derived from In-Service Inspection (ISI) Non-Destructive Evaluation (NDE) data. Based on the statistical approaches, wear growth model are proposed and applied to predict wear distribution at the End Of Cycle (EOC). Probabilistic distributions of the number of wear flaws and maximum wear depth at EOC are obtained from the analysis. Comparing the predicted EOC wear flaw data with the known EOC data the usefulness of the proposed method is examined and satisfactory results are obtained.

Flaw behavior in mechanically loaded clad plates

International Nuclear Information System (INIS)

Iskander, S.K.; Robinson, G.C.; Oland, C.B.

1989-01-01

A small crack near the inner surface of clad nuclear reactor pressure vessels is an important consideration in the safety assessment of the structural integrity of the vessel. Four-point bend tests on large plate specimens, conforming to ASTM specification for pressure vessel plates, alloy steels, quenched and tempered, Mn-Mo and Mn-Mo-Ni (A533) grade B six clad and two unclad with stainless steels 308, 309 and 312 weld wires, were performed to determine the effect of cladding upon the propagation of small surface cracks subjected to stress states. Results indicated that the tough surface layer composed of cladding and/or heat-affected zone has enhanced the load-bearing capacity of plates under conditions where unclad plates have ruptured. The results are interpreted in terms of fracture mechanics. The behavior of flaws in clad reactor pressure vessels is examined in the light of the test results. 11 refs., 8 figs., 2 tabs

Characteristics Testing of the ECT Bobbin Probe for Steam Generator Tube Inspection of Nuclear Power Plant

International Nuclear Information System (INIS)

Nam, Min Woo; Lee, Hee Jong; Cho, Chan hee; Yoo, Hyun Joo

2010-01-01

The steam generator management program(SGMP) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. The first way is to qualify the equipment or the probe by using the flaw mechanism and method of the pulled tubes from the heat exchangers or the artificial flawed tubes. The second way is to verify the equivalency with the characteristics of the qualified equipment or probe. In this case, the qualified equipment or probe may be modified to substitute or replace instruments or probes without re-qualification provided that the range of essential variables defined in the examination technique specification sheet are met. This study is to describe the result of the comparative performance evaluation of bobbin coil eddy current probes manufactured by KEPCO Research Institute and probes manufactured by a foreign manufacturer. As a result of this study, although there were minor differences between the two kinds of probes, it was evaluated that the two kinds of probes were almost identical in the significant performance characteristics described in the KEPCO Research Institute guideline

Characteristics Testing of the ECT Bobbin Probe for Steam Generator Tube Inspection of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Nam, Min Woo; Lee, Hee Jong; Cho, Chan hee; Yoo, Hyun Joo [KEPCO, Daejeon (Korea, Republic of)

2010-08-15

The steam generator management program(SGMP) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. The first way is to qualify the equipment or the probe by using the flaw mechanism and method of the pulled tubes from the heat exchangers or the artificial flawed tubes. The second way is to verify the equivalency with the characteristics of the qualified equipment or probe. In this case, the qualified equipment or probe may be modified to substitute or replace instruments or probes without re-qualification provided that the range of essential variables defined in the examination technique specification sheet are met. This study is to describe the result of the comparative performance evaluation of bobbin coil eddy current probes manufactured by KEPCO Research Institute and probes manufactured by a foreign manufacturer. As a result of this study, although there were minor differences between the two kinds of probes, it was evaluated that the two kinds of probes were almost identical in the significant performance characteristics described in the KEPCO Research Institute guideline

Biaxial loading and shallow-flaw effects on crack-tip constraint and fracture-toughness

International Nuclear Information System (INIS)

Pennell, W.E.; Bass, B.R.; Bryson, J.W.; McAfee, W.J.; Theiss, T.J.; Rao, M.C.

1993-01-01

Uniaxial tests of single-edged notched bend (SENB) specimens with both deep- and shallow-flaws have shown elevated fracture-toughness for the shallow flaws. The elevation in fracture-toughness for shallow flaws has been shown to be the result of reduced constraint at the crack-tip. Biaxial loading has the potential to increase constraint at the crack-tip and thereby reduce some of the shallow-flaw, fracture-toughness elevation. Biaxial fracture-toughness tests have shown that the shallow-flaw, fracture-toughness elevation is reduced but not eliminated by biaxial loading. Dual-parameter, fracture-toughness correlations have been proposed to reflect the effect of crack-tip constraint on fracture-toughness. Test results from the uniaxial and biaxial tests were analyzed using the dual-parameter technology. Discrepancies between analysis results and cleavage initiation site data from fractographic examinations indicate that the analysis models are in need of further refinement. Addition of a precleavage, ductile-tearing element to the analysis model has the potential to resolve the noted discrepancies

Development of an Intelligent Ultrasonic Signature Classification Software for Discrimination of Flaws in Weldments

International Nuclear Information System (INIS)

Kim, H. J.; Song, S. J.; Jeong, H. D.

1997-01-01

Ultrasonic pattern recognition is the most effective approach to the problem of discriminating types of flaws in weldments based on ultrasonic flaw signals. In spite of significant progress in the research on this methodology, it has not been widely used in many practical ultrasonic inspections of weldments in industry. Hence, for the convenient application of this approach in many practical situations, we develop an intelligent ultrasonic signature classification software which can discriminate types of flaws in weldments based on their ultrasonic signals using various tools in artificial intelligence such as neural networks. This software shows the excellent performance in an experimental problem where flaws in weldments are classified into two categories of cracks and non-cracks. This performance demonstrates the high possibility of this software as a practical tool for ultrasonic flaw classification in weldments

Calculation and evaluation methodology of the flawed pipe and the compute program development

International Nuclear Information System (INIS)

Liu Chang; Qian Hao; Yao Weida; Liang Xingyun

2013-01-01

Background: The crack will grow gradually under alternating load for a pressurized pipe, whereas the load is less than the fatigue strength limit. Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics (EPFM) criteria. Methods: In the compute, the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++. Results: The fluctuating load of the Reactor Coolant System transients, the initial flaw shape, the initial flaw orientation are all accounted here. Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not. (authors)

Advanced nondestructive examination of the reactor vessel head penetration tube welds

International Nuclear Information System (INIS)

Cvitanovic, M.; Zado, V.

1996-01-01

Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

Optimization of a Two Stage Pulse Tube Refrigerator for the Integrated Current Lead System

Science.gov (United States)

Maekawa, R.; Matsubara, Y.; Okada, A.; Takami, S.; Konno, M.; Tomioka, A.; Imayoshi, T.; Hayashi, H.; Mito, T.

2008-03-01

Implementation of a conventional current lead with a pulse tube refrigerator has been validated to be working as an Integrated Current Lead (ICL) system for the Superconducting Magnetic Energy Storage (SMES). Realization of the system is primarily accounted for the flexibility of a pulse tube refrigerator, which does not posses any mechanical piston and/or displacer. As for an ultimate version of the ICL system, a High Temperature Superconducting (HTS) lead links a superconducting coil with a conventional copper lead. To ensure the minimization of heat loads to the superconducting coil, a pulse tube refrigerator has been upgraded to have a second cooling stage. This arrangement reduces not only the heat loads to the superconducting coil but also the operating cost for a SMES system. A prototype two-stage pulse tube refrigerator, series connected arrangement, was designed and fabricated to satisfy the requirements for the ICL system. Operation of the first stage refrigerator is a four-valve mode, while the second stage utilizes a double inlet configuration to ensure its confined geometry. The paper discusses the optimization of second stage cooling to validate the conceptual design

Stress-induced light scattering method for the detection of latent flaws on fine polished glass substrates.

Science.gov (United States)

Sakata, Y; Sakai, K; Nonaka, K

2014-08-01

Fine polishing techniques, such as the chemical mechanical polishing treatment, are one of the most important technique to glass substrate manufacturing. Mechanical interaction in the form of friction occurs between the abrasive and the substrate surface during polishing, which may cause formation of latent flaws on the glass substrate surface. Fine polishing-induced latent flaws may become obvious during a subsequent cleaning process if glass surfaces are corroded away by chemical interaction with the cleaning liquid. Latent flaws thus reduce product yield. In general, non-destructive inspection techniques, such as the light-scattering methods, used to detect foreign matters on the glass substrate surface. However, it is difficult to detect latent flaws by these methods because the flaws remain closed. Authors propose a novel inspection technique for fine polishing-induced latent flaws by combining the light scattering method with stress effects, referred to as the stress-induced light scattering method (SILSM). SILSM is able to distinguish between latent flaws and particles on the surface. In this method, samples are deformed by an actuator and stress effects are induced around the tips of latent flaws. Due to the photoelastic effect, the refractive index of the material around the tip of a latent flaw is changed. This changed refractive index is in turn detected by a cooled charge-coupled device camera as variations in light scattering intensity. In this report, surface latent flaws are detected non-destructively by applying SILSM to glass substrates, and the utility of SILSM evaluated as a novel inspection technique.

Detecting accuracy of flaws by manual and automatic ultrasonic inspections

International Nuclear Information System (INIS)

Iida, K.

1988-01-01

As the final stage work in the nine year project on proving tests of the ultrasonic inspection technique applied to the ISI of LWR plants, automatic ultrasonic inspection tests were carried out on EDM notches, surface fatigue cracks, weld defects and stress corrosion cracks, which were deliberately introduced in full size structural components simulating a 1,100 MWe BWR. Investigated items are the performance of a newly assembled automatic inspection apparatus, detection limit of flaws, detection resolution of adjacent collinear or parallel EDM notches, detection reproducibility and detection accuracy. The manual ultrasonic inspection of the same flaws as inspected by the automatic ultrasonic inspection was also carried out in order to have comparative data. This paper reports how it was confirmed that the automatic ultrasonic inspection is much superior to the manual inspection in the flaw detection rate and in the detection reproducibility

Flaw characterization through nonlinear ultrasonics and wavelet cross-correlation algorithms

Science.gov (United States)

Bunget, Gheorghe; Yee, Andrew; Stewart, Dylan; Rogers, James; Henley, Stanley; Bugg, Chris; Cline, John; Webster, Matthew; Farinholt, Kevin; Friedersdorf, Fritz

2018-04-01

Ultrasonic measurements have become increasingly important non-destructive techniques to characterize flaws found within various in-service industrial components. The prediction of remaining useful life based on fracture analysis depends on the accurate estimation of flaw size and orientation. However, amplitude-based ultrasonic measurements are not able to estimate the plastic zones that exist ahead of crack tips. Estimating the size of the plastic zone is an advantage since some flaws may propagate faster than others. This paper presents a wavelet cross-correlation (WCC) algorithm that was applied to nonlinear analysis of ultrasonically guided waves (GW). By using this algorithm, harmonics present in the waveforms were extracted and nonlinearity parameters were used to indicate both the tip of the cracks and size of the plastic zone. B-scans performed with the quadratic nonlinearities were sensitive to micro-damage specific to plastic zones.

Pressure tube replication techniques using the advanced NDE system

International Nuclear Information System (INIS)

Isherwood, A.; Jarron, D.; Travers, J.; Hanley, K.

2006-01-01

Periodic and in-service inspections of fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections has driven new tooling for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) Replication System has been designed to compliment the ANDE Inspection System by providing a two plate replica system. These plates deliver a compound that makes a positive 3D mould of known ID flaws to gather information for flaw assessment. The two plate system, and the ability to retrieve and recharge the moulds in the reactor vault allows for gathering defect information with minimal critical path time. The ANDE Replication System was built on the foundation of CIGAR experience by a solid design team familiar with 3D CAD and manufacturing techniques. The tooling and controls went through a series of integration stages in the laboratory and then later with the Universal Delivery Machine (UDM) before being used on reactor starting in 2003. Once the inspection phase of an outage has been completed, the analysis team provides a list of flaw candidates that require 'root radius' information to complete the flaw assessment. This is a measure of how sharp the corners are in the defect. This data is used as part of the stress calculation that ultimately determines how many shutdown cycles that the reactor can have before that flaw must be re-inspected. The inspection tool is then swapped out of the delivery machine in the reactor vault using the versatile connectorized umbilical. The replication tool is loaded on the machine, charged with replica compound on each of the two plates, and then sent to the target channel(s). On channel, the operators use the same console as the ANDE Inspection System, but have a separate control system with a graphical display of the tool that shows its position in the channel with respect to the E-face. The axial and

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

International Nuclear Information System (INIS)

Kim, Y. S.; Jeong, Y. M.; Ahn, S. B.

2005-03-01

Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Y S; Jeong, Y M; Ahn, S B [and others

2005-03-15

Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

Behavior of deep flaws in a thick-wall cylinder under thermal shock loading

International Nuclear Information System (INIS)

Cheverton, R.D.

1979-01-01

Behavior of inner-surface flaws in thick-walled vessels was studied in a 991-mm OD x 152 mm wall x 1220 mm length cylinder with toughness properties similar to those for HSST Plate. The initial temperature of 93 0 C and a thermal shock medium of liquid nitrogen (-197 0 C) were employed. The initial flaw selected was a sharp, 16 mm deep, long (1220 mm) axial crack. Crack arrest methodology was shown to be valid for deep flaws under severe thermal shock

Development of a crack growth analysis is program for reactor head penetration

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Yull; Choi, Kwang Hee; Park, Jeong Il [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Kang, Young Hwan; Park, Sung Ho; Kim, Il; Kim, Young Jong; Yoo, Young Joon; Yoo, Wan; Maeng, Wan Young; Choi, Suk Nam; Kim, Kee Suk; Yoon, Sung Won; Kim, Jee Ho; Park, Myung Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-12-31

Crack growth analysis program for Reactor Head Penetration is being developed for applying to plants such as, Kori 1, Kori 2, Kori 3,4 YoungKwang 1,2 and Uljin 1,2 (1) Stress Evaluation - The stress analysis is required to evaluate the structure integrity for the RVH penetration tubes. The RVH penetration tubes are geometrically non-symmetry except center one. Thus, 3D finite element analysis should be employed for the stress analysis. The magnitude and distribution of residual stress resulted from welding can be determined analytically by simulation welding procedure. (2) Flaw Evaluation - There are two objectives of the penetration tube flaw evaluation to predict the time required for a crack to propagate to the acceptance criteria. The first objective is to perform the parametric evaluation for a postulated crack. The second objective is to develop the flaw evaluation program for the crack detected during the inspection. (3) Characterization of Material Properties of Alloy 600 - These study is to provide data which similarly represent the properties of PWR power plants in Korea. The data is used for analyzing of the stress distribution around penetration tubes. And the PWSCC data will be used for the crack growth rate of the penetration tubes. (author). 92 refs., 121 figs.

AE/flaw characterization for nuclear pressure vessels

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

1984-01-01

This chapter discusses the use of acoustic emission (AE) detected during continuous monitoring to identify and evaluate growing flaws in pressure vessels. Off-reactor testing and on-reactor testing are considered. Relationships for identifying acoustic emission (AE) from crack growth and using the AE data to estimate flaw severity have been developed experimentally by laboratory testing. The purpose of the off-reactor vessel test is to evaluate AE monitoring/interpretation methodology on a heavy section steel vessel under simulated reactor operating conditions. The purpose of on-reactor testing is to evaluate the capability of a monitor system to function in the reactor environment, calibrate the ability to detect AE signals, and to demonstrate that a meaningful criteria can be established to prevent false alarms. An expanded data base is needed from application testing and methodology standardization

Helically coiled tube heat exchanger

International Nuclear Information System (INIS)

Harris, A.M.

1981-01-01

In a heat exchanger such as a steam generator for a nuclear reactor, two or more bundles of helically coiled tubes are arranged in series with the tubes in each bundle integrally continuing through the tube bundles arranged in series therewith. Pitch values for the tubing in any pair of tube bundles, taken transverse to the path of the reactor coolant flow about the tubes, are selected as a ratio of two unequal integers to permit efficient operation of each tube bundle while maintaining the various tube bundles of the heat exchanger within a compact envelope. Preferably, the helix angle and tube pitch parallel to the path of coolant flow are constant for all tubes in a single bundle so that the tubes are of approximately the same length within each bundle

Engineering approach for examining crack growth and stability in flawed structures

International Nuclear Information System (INIS)

Shih, C.F.

1980-01-01

Progress made in two research programs sponsored by the Electric Power Research Institute (EPRI), to identify viable parameters for characterizing crack initiation and continued extension, and to develop an engineering/design methodology, based on these parameters, for the assessment of crack growth and instability in engineering structures which are stressed beyond the regime of applicability of linear elastic fracture mechanics is reported. The goal in the development of such methodology is to establish an improved basis for analyzing the effect of flaws (postulated or detected) on the safety margins of pressure boundary components of light water-cooled type nuclear steam supply systems. The methodology can also be employed for structural integrity analyses of other engineering structures

A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

International Nuclear Information System (INIS)

Yoshihisa Shindo; Kazuo Haga

2005-01-01

Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

Development of a shallow-flaw fracture assessment methodology for nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Bass, B.R.; Bryson, J.W.; Dickson, T.L.; McAfee, W.J.; Pennell, W.E.

1996-01-01

Shallow-flaw fracture technology is being developed within the Heavy-Section Steel Technology (HSST) Program for application to the safety assessment of radiation-embrittled nuclear reactor pressure vessels (RPVs) containing postulated shallow flaws. Cleavage fracture in shallow-flaw cruciform beam specimens tested under biaxial loading at temperatures in the lower transition temperature range was shown to be strain-controlled. A strain-based dual-parameter fracture toughness correlation was developed and shown to be capable of predicting the effect of crack-tip constraint on fracture toughness for strain-controlled fracture. A probabilistic fracture mechanics (PFM) model that includes both the properties of the inner-surface stainless-steel cladding and a biaxial shallow-flaw fracture toughness correlation gave a reduction in probability of cleavage initiation of more than two orders of magnitude from an ASME-based reference case

'Moral distress'--time to abandon a flawed nursing construct?

Science.gov (United States)

Johnstone, Megan-Jane; Hutchinson, Alison

2015-02-01

Moral distress has been characterised in the nursing literature as a major problem affecting nurses in all healthcare systems. It has been portrayed as threatening the integrity of nurses and ultimately the quality of patient care. However, nursing discourse on moral distress is not without controversy. The notion itself is conceptually flawed and suffers from both theoretical and practical difficulties. Nursing research investigating moral distress is also problematic on account of being methodologically weak and disparate. Moreover, the ultimate purpose and significance of the research is unclear. In light of these considerations, it is contended that the notion of moral distress ought to be abandoned and that concerted attention be given to advancing inquiries that are more conducive to improving the quality and safety of moral decision-making, moral conduct and moral outcomes in nursing and healthcare domains. © The Author(s) 2013.

Evaluation of sampling schemes for in-service inspection of steam generator tubing

International Nuclear Information System (INIS)

Hanlen, R.C.

1990-03-01

This report is a follow-on of work initially sponsored by the US Nuclear Regulatory Commission (Bowen et al. 1989). The work presented here is funded by EPRI and is jointly sponsored by the Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC). The goal of this research was to evaluate fourteen sampling schemes or plans. The main criterion used for evaluating plan performance was the effectiveness for sampling, detecting and plugging defective tubes. The performance criterion was evaluated across several choices of distributions of degraded/defective tubes, probability of detection (POD) curves and eddy-current sizing models. Conclusions from this study are dependent upon the tube defect distributions, sample size, and expansion rules considered. As degraded/defective tubes form ''clusters'' (i.e., maps 6A, 8A and 13A), the smaller sample sizes provide a capability of detecting and sizing defective tubes that approaches 100% inspection. When there is little or no clustering (i.e., maps 1A, 20 and 21), sample efficiency is approximately equal to the initial sample size taken. Thee is an indication (though not statistically significant) that the systematic sampling plans are better than the random sampling plans for equivalent initial sample size. There was no indication of an effect due to modifying the threshold value for the second stage expansion. The lack of an indication is likely due to the specific tube flaw sizes considered for the six tube maps. 1 ref., 11 figs., 19 tabs

Detection and Characterization of Flaws in Sprayed on Foam Insulation with Pulsed Terahertz Frequency Electromagnetic Waves

Science.gov (United States)

Winfree, William P.; Madaras, Eric I.

2005-01-01

The detection and repair of flaws such as voids and delaminations in the sprayed on foam insulation of the external tank reduces the probability of foam debris during shuttle ascent. The low density of sprayed on foam insulation along with it other physical properties makes detection of flaws difficult with conventional techniques. An emerging technology that has application for quantitative evaluation of flaws in the foam is pulsed electromagnetic waves at terahertz frequencies. The short wavelengths of these terahertz pulses make them ideal for imaging flaws in the foam. This paper examines the application of terahertz pulses for flaw detection in foam characteristic of the foam insulation of the external tank. Of particular interest is the detection of voids and delaminations, encapsulated in the foam or at the interface between the foam and a metal backing. The technique is shown to be capable of imaging small voids and delaminations through as much as 20 cm of foam. Methods for reducing the temporal responses of the terahertz pulses to improve flaw detection and yield quantitative characterizations of the size and location of the flaws are discussed.

Operational control and maintenance integrity of typical and atypical coil tube steam generating systems

Energy Technology Data Exchange (ETDEWEB)

Beardwood, E.S.

1999-07-01

Coil tube steam generators are low water volume to boiler horsepower (bhp) rating, rapid steaming units which occupy substantially less space per boiler horsepower than equivalent conventional tire tube and water tube boilers. These units can be retrofitted into existing steam systems with relative ease and are more efficient than the generators they replace. During the early 1970's they became a popular choice for steam generation in commercial, institutional and light to medium industrial applications. Although these boiler designs do not require skilled or certified operators, an appreciation for a number of the operational conditions that result in lower unscheduled maintenance, increased reliability and availability cycles would be beneficial to facility owners, managers, and operators. Conditions which afford lower operating and maintenance costs will be discussed from a practical point of view. An overview of boiler design and operation is also included. Pitfalls are provided for operational and idle conditions. Water treatment application, as well as steam system operations not conducive to maintaining long term system integrity; with resolutions, will be addressed.

Rejection index for pressure tubes

International Nuclear Information System (INIS)

Mitchell, A.B.; Meneley, D.

1989-10-01

The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

Assesment of integrity of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Brozova, A.; Zdarek, J.

1992-01-01

Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

Study of thermal performance of capillary micro tubes integrated into the building sandwich element made of high performance concrete

DEFF Research Database (Denmark)

Mikeska, Tomas; Svendsen, Svend

2013-01-01

The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of high performance concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CM...... and cooling purposes of future low energy buildings. The investigations were conceived as a low temperature concept, where the difference between the temperature of circulating fluid and air in the room was kept in range of 1–4 °C.......The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of high performance concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CMT...... HPC layer covering the CMT. This paper shows that CMT integrated into the thin plate of sandwich element made of HPC can supply the energy needed for heating (cooling) and at the same time create the comfortable and healthy environment for the occupants. This solution is very suitable for heating...

Determination of Flaw Size and Depth From Temporal Evolution of Thermal Response

Science.gov (United States)

Winfree, William P.; Zalameda, Joseph N.; Cramer, Elliott; Howell, Patricia A.

2015-01-01

Simple methods for reducing the pulsed thermographic responses of flaws have tended to be based on either the spatial or temporal response. This independent assessment limits the accuracy of characterization. A variational approach is presented for reducing the thermographic data to produce an estimated size for a flaw that incorporates both the temporal and spatial response to improve the characterization. The size and depth are determined from both the temporal and spatial thermal response of the exterior surface above a flaw and constraints on the length of the contour surrounding the delamination. Examples of the application of the technique to simulation and experimental data acquired are presented to investigate the limitations of the technique.

Integration of a photocatalytic multi-tube reactor for indoor air purification in HVAC systems: a feasibility study.

Science.gov (United States)

van Walsem, Jeroen; Roegiers, Jelle; Modde, Bart; Lenaerts, Silvia; Denys, Siegfried

2018-04-24

This work is focused on an in-depth experimental characterization of multi-tube reactors for indoor air purification integrated in ventilation systems. Glass tubes were selected as an excellent photocatalyst substrate to meet the challenging requirements of the operating conditions in a ventilation system in which high flow rates are typical. Glass tubes show a low-pressure drop which reduces the energy demand of the ventilator, and additionally, they provide a large exposed surface area to allow interaction between indoor air contaminants and the photocatalyst. Furthermore, the performance of a range of P25-loaded sol-gel coatings was investigated, based on their adhesion properties and photocatalytic activities. Moreover, the UV light transmission and photocatalytic reactor performance under various operating conditions were studied. These results provide vital insights for the further development and scaling up of multi-tube reactors in ventilation systems which can provide a better comfort, improved air quality in indoor environments, and reduced human exposure to harmful pollutants.

A Laboratory Experimental Study: An FBG-PVC Tube Integrated Device for Monitoring the Slip Surface of Landslides

Science.gov (United States)

Zhang, Shaojie; Chen, Jiang; Teng, Pengxiao; Wei, Fangqiang; Chen, Qiao

2017-01-01

A new detection device was designed by integrating fiber Bragg grating (FBG) and polyvinyl chloride (PVC) tube in order to monitor the slip surface of a landslide. Using this new FBG-based device, a corresponding slope model with a pre-set slip surface was designed, and seven tests with different soil properties were carried out in laboratory conditions. The FBG sensing fibers were fixed on the PVC tube to measure strain distributions of PVC tube at different elevation. Test results indicated that the PVC tube could keep deformation compatible with soil mass. The new device was able to monitor slip surface location before sliding occurrence, and the location of monitored slip surface was about 1–2 cm above the pre-set slip surface, which basically agreed with presupposition results. The monitoring results are expected to be used to pre-estimate landslide volume and provide a beneficial option for evaluating the potential impact of landslides on shipping safety in the Three Gorges area. PMID:29084157

Comparison of steam generator methods in PISC

International Nuclear Information System (INIS)

Lahdenperae, K.; Kankare, M.

1996-01-01

The main objective of the study (PISC III, action 5) was the experimental evaluation of the performance of methods used in in-service inspection of steam generator tubes used in nuclear power plants. The study was organized by the Joint Research Center of the European Community (JRC). The round robin test with blind boxes started in 1991. During the study training boxes and blind boxes were circulated in 29 laboratories in Europe, Japan and the USA. The boxes contained steam generator tubes with artificial and natural (chemically induced) flaws. The material was inconell. The blind boxes contained 66 tubes and 95 flaws. All flaws were introduced into different discontinuities, under support plates, above the tube sheet and into U-bends. The flaws included volumetric flaws (wastage, pitting, wear), axial and circumferential notches and chemically induced SCC cracks and IGA. After the round robin test the reference laboratory performed the destructive examination of reported flaws. The flaw detection probability (FDP) for all flaws and for teams inspecting all tubes was 60-85%. The detection of flaws deeper than 40% of the wall thickness was good. Flaws with a depth of less than 20% were not detected. When all flaws were considered, depth sizing was found to have a wide dispersion. Similarly, measured lengths did not as a rule correlate with true lengths. The classification of flaws in cracks and of volumetric flaws was not very successful, the correct classification probability being only about 70%. Evaluation of the flaws showed some shortcomings. The correct rejection probability was at best 83% for teams inspecting all boxes. (3 refs.)

Probabilistic Fracture Mechanics of Reactor Pressure Vessels with Populations of Flaws

Energy Technology Data Exchange (ETDEWEB)

Spencer, Benjamin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Backman, Marie [Univ. of Tennessee, Knoxville, TN (United States); Williams, Paul [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Dickson, Terry [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bass, B. Richard [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Klasky, Hilda [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-01

This report documents recent progress in developing a tool that uses the Grizzly and RAVEN codes to perform probabilistic fracture mechanics analyses of reactor pressure vessels in light water reactor nuclear power plants. The Grizzly code is being developed with the goal of creating a general tool that can be applied to study a variety of degradation mechanisms in nuclear power plant components. Because of the central role of the reactor pressure vessel (RPV) in a nuclear power plant, particular emphasis is being placed on developing capabilities to model fracture in embrittled RPVs to aid in the process surrounding decision making relating to life extension of existing plants. A typical RPV contains a large population of pre-existing flaws introduced during the manufacturing process. The use of probabilistic techniques is necessary to assess the likelihood of crack initiation at one or more of these flaws during a transient event. This report documents development and initial testing of a capability to perform probabilistic fracture mechanics of large populations of flaws in RPVs using reduced order models to compute fracture parameters. The work documented here builds on prior efforts to perform probabilistic analyses of a single flaw with uncertain parameters, as well as earlier work to develop deterministic capabilities to model the thermo-mechanical response of the RPV under transient events, and compute fracture mechanics parameters at locations of pre-defined flaws. The capabilities developed as part of this work provide a foundation for future work, which will develop a platform that provides the flexibility needed to consider scenarios that cannot be addressed with the tools used in current practice.

Simulating the x-ray image contrast to setup techniques with desired flaw detectability

Science.gov (United States)

Koshti, Ajay M.

2015-04-01

The paper provides simulation data of previous work by the author in developing a model for estimating detectability of crack-like flaws in radiography. The methodology is developed to help in implementation of NASA Special x-ray radiography qualification, but is generically applicable to radiography. The paper describes a method for characterizing the detector resolution. Applicability of ASTM E 2737 resolution requirements to the model are also discussed. The paper describes a model for simulating the detector resolution. A computer calculator application, discussed here, also performs predicted contrast and signal-to-noise ratio calculations. Results of various simulation runs in calculating x-ray flaw size parameter and image contrast for varying input parameters such as crack depth, crack width, part thickness, x-ray angle, part-to-detector distance, part-to-source distance, source sizes, and detector sensitivity and resolution are given as 3D surfaces. These results demonstrate effect of the input parameters on the flaw size parameter and the simulated image contrast of the crack. These simulations demonstrate utility of the flaw size parameter model in setting up x-ray techniques that provide desired flaw detectability in radiography. The method is applicable to film radiography, computed radiography, and digital radiography.

Validation of favor code linear elastic fracture solutions for finite-length flaw geometries

International Nuclear Information System (INIS)

Dickson, T.L.; Keeney, J.A.; Bryson, J.W.

1995-01-01

One of the current tasks within the US Nuclear Regulatory Commission (NRC)-funded Heavy Section Steel Technology Program (HSST) at Oak Ridge National Laboratory (ORNL) is the continuing development of the FAVOR (Fracture, analysis of Vessels: Oak Ridge) computer code. FAVOR performs structural integrity analyses of embrittled nuclear reactor pressure vessels (RPVs) with stainless steel cladding, to evaluate compliance with the applicable regulatory criteria. Since the initial release of FAVOR, the HSST program has continued to enhance the capabilities of the FAVOR code. ABAQUS, a nuclear quality assurance certified (NQA-1) general multidimensional finite element code with fracture mechanics capabilities, was used to generate a database of stress-intensity-factor influence coefficients (SIFICs) for a range of axially and circumferentially oriented semielliptical inner-surface flaw geometries applicable to RPVs with an internal radius (Ri) to wall thickness (w) ratio of 10. This database of SIRCs has been incorporated into a development version of FAVOR, providing it with the capability to perform deterministic and probabilistic fracture analyses of RPVs subjected to transients, such as pressurized thermal shock (PTS), for various flaw geometries. This paper discusses the SIFIC database, comparisons with other investigators, and some of the benchmark verification problem specifications and solutions

Detecting flaws in welds

International Nuclear Information System (INIS)

Woodacre, A.; Lawton, H.

1979-01-01

An apparatus and a method for detecting flaws in welds in a workpiece, the portion of the workpiece containing the weld is maintained at a constant temperature and the weld is scanned by an infra red detector. The weld is then scanned again with the workpiece in contact with a cooling probe to produce a steeper temperature gradient across the weld. Comparison of the signals produced by each scan reveals the existence of defects in the welds. The signals may be displayed on an oscilloscope and the display may be observed by a TV camera and recorded on videotape. (UK)

Origin and type of flaws in heat engine ceramic materials and components

International Nuclear Information System (INIS)

Govila, R.K.

1995-01-01

A number of ceramic materials such as Silicon Nitrides and Carbides, Sialons, Whisker-Reinforced Ceramic Composites and Partially-Stabilized Zirconias (PSZs) have been developed for use as structural components in heat engine applications. The reliability and durability of a structural engine component is critically dependent on the size, density of distribution and location of flaws. This information is critical for the processing and design engineers in order to design structural components using suitable materials and thus minimize stress intensity. In general, the failure initiating flaws are associated or produced due to material impurity, processing methods and parameters, and fabrication techniques (machining and grinding). Examples of each type of flaws associated with material impurity, processing methods and fabrication techniques are illustrated

A new digital correlation flaw detection system

International Nuclear Information System (INIS)

Lee, B.B.; Furgason, E.S.

1981-01-01

A new portable digital random signal flaw detection system is described which uses a digital delay line to replace the acoustic delay line of the original random signal system. Using this new system, a comparison was made between the two types of transmit signals which have been used in previous systems--m-sequences and random signals. This comparison has not been possible with these previous correlation flaw detection systems. Results indicated that for high-speed short code operation, the m-sequences produced slightly lower range sidelobes than typical samples of a clipped random signal. For normal long code operation, results indicated that system performance is essentially equivalent in resolution and signal-to-noise ratio using either m-sequences or clipped and sampled random signals. Further results also showed that for normal long code operation, the system produces outputs equivalent in resolution to pulse-echo systems, but with the added benefit of signal-to-noise ratio enhancement

Development of the Automated Ultrasonic Testing System for Inspection of the flaw in the Socket Weldment

International Nuclear Information System (INIS)

Lee, Jeong Ki; Park, Moon Ho; Park, Ki Sung; Lee, Jae Ho; Lim, Sung Jin

2004-01-01

Socket weldment used to change the flow direction of fluid nay have flaws such as lack of fusion and cracks. Liquid penetrant testing or Radiography testing have been applied as NDT methods for flaw detection of the socket weldment. But it is difficult to detect the flaw inside of the socket weldment with these methods. In order to inspect the flaws inside the socket weldment, a ultrasonic testing method is established and a ultrasonic transducer and automated ultrasonic testing system are developed for the inspection. The automated ultrasonic testing system is based on the portable personal computer and operated by the program based Windows 98 or 2000. The system has a pulser/receiver, 100MHz high speed A/D board, and basic functions of ultrasonic flaw detector using the program. For the automated testing, motion controller board of ISA interface type is developed to control the 4-axis scanner and a real time iC-scan image of the automated testing is displayed on the monitor. A flaws with the size of less than 1mm in depth are evaluated smaller than its actual site in the testing, but the flaws larger than 1mm appear larger than its actual size on the contrary. This tendency is shown to be increasing as the flaw size increases. h reliable and objective testing results are obtained with the developed system, so that it is expected that it can contribute to safety management and detection of repair position of pipe lines of nuclear power plants and chemical plants

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

International Nuclear Information System (INIS)

Cunha, Sérgio B.; Netto, Theodoro A.

2012-01-01

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3″ diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: ► An analytical model for the burst of a pipe with a volumetric flaw is developed. ► Deformation, strain and stress are modeled in the elastic and plastic domains. ► The model is comprehensively validated by experiments and numerical simulations. ► The burst pressure model’s accuracy is equivalent to finite element simulations.

Crack propagation from a filled flaw in rocks considering the infill influences

Science.gov (United States)

Chang, Xu; Deng, Yan; Li, Zhenhua; Wang, Shuren; Tang, C. A.

2018-05-01

This study presents a numerical and experimental study of the cracking behaviour of rock specimen containing a single filled flaw under compression. The primary aim is to investigate the influences of infill on crack patterns, load-displacement response and specimen strength. The numerical code RFPA2D (Rock Failure Process Analysis) featured by the capability of modeling heterogeneous materials is employed to develop the numerical model, which is further calibrated by physical tests. The results indicate that there exists a critical infill strength which controls crack patterns for a given flaw inclination angle. For case of infill strength lower than the critical value, the secondary or anti-cracks are disappeared by increasing the infill strength. If the infill strength is greater than the critical value, the filled flaw has little influence on the cracking path and the specimen fails by an inclined crack, as if there is no flaw. The load-displacement responses show specimen stiffness increases by increasing infill strength until the infill strength reaches its critical value. The specimen strength increases by increasing the infill strength and almost keeps constant as the infill strength exceeds its critical value.

Manufacturing of tailored tubes with a process integrated heat treatment

Science.gov (United States)

Hordych, Illia; Boiarkin, Viacheslav; Rodman, Dmytro; Nürnberger, Florian

2017-10-01

The usage of work-pieces with tailored properties allows for reducing costs and materials. One example are tailored tubes that can be used as end parts e.g. in the automotive industry or in domestic applications as well as semi-finished products for subsequent controlled deformation processes. An innovative technology to manufacture tubes is roll forming with a subsequent inductive heating and adapted quenching to obtain tailored properties in the longitudinal direction. This processing offers a great potential for the production of tubes with a wide range of properties, although this novel approach still requires a suited process design. Based on experimental data, a process simulation is being developed. The simulation shall be suitable for a virtual design of the tubes and allows for gaining a deeper understanding of the required processing. The model proposed shall predict microstructural and mechanical tube properties by considering process parameters, different geometries, batch-related influences etc. A validation is carried out using experimental data of tubes manufactured from various steel grades.

Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

International Nuclear Information System (INIS)

1980-03-01

The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

Inspection of small multi-layered plastic tubing during extrusion, using low-energy X-ray beams

International Nuclear Information System (INIS)

Armentrout, C.; Basinger, T.; Beyer, J.; Colesa, B.; Olsztyn, P.; Smith, K.; Strandberg, C.; Sullivan, D.; Thomson, J.

1999-01-01

The automotive industry uses nylon tubing with a thin ETFE (ethylene-tetrafluroethylene) inner layer to carry fuel from the tank to the engine. This fluorocarbon inner barrier layer is important to reduce the migration of hydrocarbons into the environment. Pilot Industries has developed a series of real-time inspection stations for dimensional measurements and flaw detection during the extrusion of this tubing. These stations are named LERA TM (low-energy radioscopic analysis), use a low energy X-ray source, a special high-resolution image converter and intensifier (ICI) stage, image capture hardware, a personal computer, and software that was specially designed to meet this task. Each LERA TM station operates up to 20 h a day, 6 days a week and nearly every week of the year. The tubing walls are 1-2 mm thick and the outer layer is nylon and the inner 0.2 mm thick layer is ethylene-tetrafluroethylene

Proposal of limit moment equation applicable to planar/non-planar flaw in wall thinned pipes under bending

International Nuclear Information System (INIS)

Tsuji, Masataka; Meshii, Toshiyuki

2011-01-01

Highlights: → A limit moment equation applicable to planar/non-planar flaw of 0 ≤ θ ≤ π found in wall thinned straight pipes was proposed. → An idea to rationally classify planar/non-planar flaw in wall thinned pipes was proposed. → The equation based on the experimental observation focused on the fracture mode. - Abstract: In this paper, a limit bending moment equation applicable to all types of planar and non-planar flaws in wall-thinned straight pipes under bending was proposed. A system to rationally classify the planar/non-planar flaws in wall-thinned pipes was suggested based on experimental observations focused on the fracture mode. The results demonstrate the importance of distinguishing between axial and circumferential long flaws in wall-thinned pipes.

A study on the development of a real-time intelligent system for ultrasonic flaw classification

International Nuclear Information System (INIS)

Song, Sung Jin; Kim, Hak Joon; Lee, Hyun; Lee, Seung Seok

1998-01-01

In spite of significant progress in research on ultrasonic pattern recognition it is not widely used in many practical field inspection in weldments. For the convenience of field application of this methodology, following four key issues have to be suitably addressed; 1) a software where the ultrasonic pattern recognition algorithm is efficiently implemented, 2) a real-time ultrasonic testing system which can capture the digitized ultrasonic flaw signal so the pattern recognition software can be applied in a real-time fashion, 3) database of ultrasonic flaw signals in weldments, which is served as a foundation of the ultrasonic pattern recognition algorithm, and finally, 4) ultrasonic features which should be invariant to operational variables of the ultrasonic test system. Presented here is the recent progress in the development of a real-time ultrasonic flaw classification by the novel combination of followings; an intelligent software for ultrasonic flaw classification in weldments, a computer-base real-time ultrasonic nondestructive evaluation system, database of ultrasonic flaw signals, and invariant ultrasonic features called 'normalized features.'

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

Energy Technology Data Exchange (ETDEWEB)

Cunha, Sergio B., E-mail: sbcunha@petrobras.com.br [PETROBRAS/TRANSPETRO, Av. Pres. Vargas 328 - 7th floor, Rio de Janeiro, RJ 20091-060 (Brazil); Netto, Theodoro A., E-mail: tanetto@lts.coppe.ufrj.br [COPPE, Federal University ot Rio de Janeiro, Ocean Engineering Department, PO BOX 68508, Rio de Janeiro - RJ (Brazil)

2012-01-15

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3 Double-Prime diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: Black-Right-Pointing-Pointer An analytical model for the burst of a pipe with a volumetric flaw is developed. Black-Right-Pointing-Pointer Deformation, strain and stress are modeled in the elastic and plastic domains. Black-Right-Pointing-Pointer The model is comprehensively validated by experiments and numerical simulations. Black-Right-Pointing-Pointer The burst pressure model's accuracy is equivalent to finite element simulations.

An integrated automatic system for the eddy-current testing of the steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1995-12-31

This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

An integrated automatic system for the eddy-current testing of the steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1996-12-31

This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

Flaw assessment procedure for high temperature reactor components

International Nuclear Information System (INIS)

Ainsworth, R.A.; Takahashi, Y.

1990-01-01

An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the USA, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program. 20 refs

Flaw density examinations of a clad boiling water reactor pressure vessel segment

International Nuclear Information System (INIS)

Cook, K.V.; McClung, R.W.

1986-01-01

Flaw density is the greatest uncertainty involved in probabilistic analyses of reactor pressure vessel failure. As part of the Heavy-Section Steel Technology (HSST) Program, studies have been conducted to determine flaw density in a section of reactor pressure vessel cut from the Hope Creek Unit 2 vessel [nominally 0.7 by 3 m (2 by 10 ft)]. This section (removed from the scrapped vessel that was never in service) was evaluated nondestructively to determine the as-fabricated status. We had four primary objectives: (1) evaluate longitudinal and girth welds for flaws with manual ultrasonics, (2) evaluate the zone under the nominal 6.3-mm (0.25-in.) clad for cracking (again with manual ultrasonics), (3) evaluate the cladding for cracks with a high-sensitivity fluorescent penetrant method, and (4) determine the source of indications detected

Methods and means of the radioisotope flaw detection of the nuclear power reactors components

International Nuclear Information System (INIS)

Dekopov, A.S.; Majorov, A.N.; Firsov, V.G.

1979-01-01

Methods and means are considered for the radioisotopic flaw detection of the nuclear reactors pressure vessels and structural components of the reactor circuit. Methods of control are described as in the technological process of fabrication of the power reactors assemblies as during the systematic-preventive repair of the nuclear power station equipment during exploitation. Methodological base is given of the technology of radiation control of welded joints of the pressure vessel branch piper of the WWER-440 and WWER-1000 reactors in the process of assembling and exploitation and joining pipes with the pipe-plate of the steamgenerator in the process of fabrication. Methods of the radioisotope flaw detection in the process of exploitation take into consideration the influence of the radioisotope background, and ensure obtaining of the demanded by the rules of control, sensitivity. Methods of control of welded joints of the steamgenerator of nuclear power plants are based on the simultaneous examination of all joints with application of the shaped radiographic plate-holders. Special gamma-flaw-detection equipment is developed for control of the welded joints of the main branch-pipes. Design peculiarities are given of the installation for flaw detection. These installations are equipped with the system for emergency return of the radiation source into the storage position from the position for exposure. They have automatic exposure-meters for determination of the exposure time. Successfull exploitation of such installations in the Finland during assembling equipment for the nuclear reactor of the nuclear power plant ''Loviisa-1'' and in the USSR on the Novovoronezh nuclear power plant has shown possibility for detection of flaws having dimensions about 1% of the equipment used. For control of welded joints of pipes with pipe-plates at the steam generators, portable flaw-detectors are used. Sensitivity of these flaw-detectors towards detection of the wire standards has

A dynamic fatigue study of soda-lime silicate and borosilicate glasses using small scale indentation flaws

International Nuclear Information System (INIS)

Dabbs, T.P.; Lawn, B.R.; Kelly, P.L.

1982-01-01

The dynamic fatigue characteristics of two glasses, soda-lime silicate and borosilicate, in water have been studied using a controlled indentation flaw technique. It is argued that the indentation approach offers several advantages over more conventional fatigue testing procedures: (i) the reproducibility of data is relatively high, eliminating statistics as a basis of analysis: (ii) the flaw ultimately responsible for failure is well defined and may be conveniently characterised before and after (and during, if necessary) the strength test; (iii) via adjustment of the indentation load, the size of the flaw can be suitably predetermined. Particular attention is devoted to the third point because of the facility it provides for systematic investigation of the range of flaw sizes over which macroscopic crack behaviour remains applicable. The first part of the paper summarises the essential fracture mechanics theory of the extension of an indentation flaw to failure. In the next part of the paper the results of dynamic fatigue tests on glass rods in distilled water are described. Data are obtained for Vickers indentation loads in the range 0.05 to 100 N, corresponding to contact dimensions of 2 to 100 μm. Finally, the implications of the results in relation to the response of 'natural' flaws are discussed. (author)

On flaw tolerance of nacre: a theoretical study

Science.gov (United States)

Shao, Yue; Zhao, Hong-Ping; Feng, Xi-Qiao

2014-01-01

As a natural composite, nacre has an elegant staggered ‘brick-and-mortar’ microstructure consisting of mineral platelets glued by organic macromolecules, which endows the material with superior mechanical properties to achieve its biological functions. In this paper, a microstructure-based crack-bridging model is employed to investigate how the strength of nacre is affected by pre-existing structural defects. Our analysis demonstrates that owing to its special microstructure and the toughening effect of platelets, nacre has a superior flaw-tolerance feature. The maximal crack size that does not evidently reduce the tensile strength of nacre is up to tens of micrometres, about three orders higher than that of pure aragonite. Through dimensional analysis, a non-dimensional parameter is proposed to quantify the flaw-tolerance ability of nacreous materials in a wide range of structural parameters. This study provides us some inspirations for optimal design of advanced biomimetic composites. PMID:24402917

Encapsulation of Fluidic Tubing and Microelectrodes in Microfluidic Devices: Integrating Off-Chip Process and Coupling Conventional Capillary Electrophoresis with Electrochemical Detection.

Science.gov (United States)

Becirovic, Vedada; Doonan, Steven R; Martin, R Scott

2013-08-21

In this paper, an approach to fabricate epoxy or polystyrene microdevices with encapsulated tubing and electrodes is described. Key features of this approach include a fixed alignment between the fluidic tubing and electrodes, the ability to polish the device when desired, and the low dead volume nature of the fluidic interconnects. It is shown that a variety of tubing can be encapsulated with this approach, including fused silica capillary, polyetheretherketone (PEEK), and perfluoroalkoxy (PFA), with the resulting tubing/microchip interface not leading to significant band broadening or plug dilution. The applicability of the devices with embedded tubing is demonstrated by integrating several off-chip analytical methods to the microchip. This includes droplet transfer, droplet desegmentation, and microchip-based flow injection analysis. Off-chip generated droplets can be transferred to the microchip with minimal coalescence, while flow injection studies showed improved peak shape and sensitivity when compared to the use of fluidic interconnects with an appreciable dead volume. Importantly, it is shown that this low dead volume approach can be extended to also enable the integration of conventional capillary electrophoresis (CE) with electrochemical detection. This is accomplished by embedding fused silica capillary along with palladium (for grounding the electrophoresis voltage) and platinum (for detection) electrodes. With this approach, up to 128,000 theoretical plates for dopamine was possible. In all cases, the tubing and electrodes are housed in a rigid base; this results in extremely robust devices that will be of interest to researchers wanting to develop microchips for use by non-experts.

Internal Rot Detection with the Use of Low-Frequency Flaw Detector

Science.gov (United States)

Proskórnicki, Marek; Ligus, Grzegorz

2014-12-01

The issue of rot detection in standing timber or stocked wood is very important in forest management. Rot flaw detection used for that purpose is represented by invasive and non-invasive devices. Non-invasive devices are very accurate, but due to the cost and complicated operation they have not been applied on a large scale in forest management. Taking into account the practical needs of foresters a prototype of low-frequency flaw was developed. The principle of its operation is based on the difference in acoustic wave propagation in sound wood and wood with rot.

30 CFR 250.517 - Tubing and wellhead equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 2 2010-07-01 2010-07-01 false Tubing and wellhead equipment. 250.517 Section... Tubing and wellhead equipment. (a) No tubing string shall be placed in service or continue to be used unless such tubing string has the necessary strength and pressure integrity and is otherwise suitable for...

Structural and leakage integrity assessment of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

1996-01-01

The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

Working session 2: Tubing inspection

International Nuclear Information System (INIS)

Guerra, J.; Tapping, R.L.

1997-01-01

This session was attended by delegates from 10 countries, and four papers were presented. A wide range of issues was tabled for discussion. Realizing that there was limited time available for more detailed discussion, three topics were chosen for the more detailed discussion: circumferential cracking, performance demonstration (to focus on POD and sizing), and limits of methods. Two other subsessions were organized: one dealt with some challenges related to the robustness of current inspection methods, especially with respect to leaving cracked tubes in service, and the other with developing a chart of current NDE technology with recommendations for future development. These three areas are summarized in turn, along with conclusions and/or recommendations. During the discussions there were four presentations. There were two (Canada, Japan) on eddy current probe developments, both of which addressed multiarray probes that would detect a range of flaws, one (Spain) on circumferential crack detection, and one (JRC, Petten) on the recent PISC III results

Tube collector with integrated tracking parabolic concentrator

Energy Technology Data Exchange (ETDEWEB)

Grass, C.; Benz, N.; Hacker, Z.; Timinger, A. [ZAE Bayern, Bavarian Centre for Applied Energy Research, Muenchen (Germany)

2000-07-01

Low concentrating CPC collectors usually do not track the sun and are mounted in east-west direction with a latitude dependent slope angle. They are most suitable for maximum working temperatures up to 200 250 deg. C. We present a novel evacuated tube-collector with a trough-like concentrating mirror. Single-axis tracking of the mirror is realized with a magnetic mechanism. The mirror is mounted inside the evacuated tube and hence protected from environmental influences. One axis tracking in combination with a small acceptance angle allows for higher concentration as compared to non-tracking concentrating collectors. Ray-tracing analysis shows a half acceptance angle of about 5 deg. at a geometrical concentration ratio of 3.2. The losses of evacuated tube collectors are dominated by the radiation losses of the absorber. Hence, reducing the absorber size can lead to higher efficiencies at high operating temperature levels. With the presented collector we aim for operating temperatures up to 400 deg. C. At temperatures of 300 deg. C we expect efficiencies of 65 %. This allows for application in industrial process heat generation, high efficient solar cooling and power generation. A first prototype was tested at the ZAE Bayern. The optical efficiency was measured to be 75 %. (au)

Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study

International Nuclear Information System (INIS)

Bahn, Chi Bum; Majumdar, Saurin; Kasza, Ken E.; Shack, William J.

2013-01-01

To address concerns regarding excessive leakage from throughwall cracks in steam generator tube-to-tubesheet joints under severe accident conditions, leak rate testing was conducted using tube-to-collar joint specimens. The tube interior and the interface between tube and collar (crevice) were pressurized independently using nitrogen gas. The leak rate through the crevice was almost zero when the specimens were pressurized at ∼500 °C; this low leak rate is attributed to thermal mismatch effects preventing much leakage. The near zero leak rate was maintained until the onset of large leakage at higher temperatures. The leak rate behavior after the onset of the large leakage was not much affected by the crevice length or heat-to-heat variation of Alloy 600 tubes. This suggests that once the crevice gap opens, the creep rate of the low alloy steel collar becomes dominant. Specimens with different tube diameters behaved essentially the same way. To simulate a flawed steam generator tube in the tubesheet, the crevice region was pressurized through a hole in the tube. This simulation resulted in essentially the same behavior as those specimens whose tubes and crevices were pressurized independently. Oxidation of low alloy steel collars in air tests can increase the flow resistance, and thus tests using nitrogen gas would provide more conservative leak rate data. Highlights: ► Leak rates were measured by using tube-to-collar joint specimens under creep condition. ► Leak rate through the joint interface was almost zero at ∼500 °C due to thermal mismatch. ► The near zero leak rate was maintained until the onset of large leakage at ∼680 °C. ► The leak behavior after the onset of the large leakage was not affected by hydraulic expansion length or tube heats.

Measurement of flaw size in a weld sample by ultrasonic frequency analysis

International Nuclear Information System (INIS)

Whaley, H.L. Jr.; Adler, L.; Cook, K.V.; McClung, R.W.

1975-05-01

An ultrasonic frequency analysis technique has been developed and applied to the measurement of flaws in an 8-in.-thick heavy-section steel specimen belonging to the Pressure Vessel Research Committee program. Using the technique the flaws occurring in the weld area were characterized in quantitative terms of both dimension and orientation. Several modifications of the technique were made during the study to include the application of several transducers and to consider ultrasonic mode conversion. (U.S.)

WWER Steam Generators Tubing Performance and Aging Management

International Nuclear Information System (INIS)

Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

2006-01-01

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

Detector-device-independent quantum secret sharing with source flaws.

Science.gov (United States)

Yang, Xiuqing; Wei, Kejin; Ma, Haiqiang; Liu, Hongwei; Yin, Zhenqiang; Cao, Zhu; Wu, Lingan

2018-04-10

Measurement-device-independent entanglement witness (MDI-EW) plays an important role for detecting entanglement with untrusted measurement device. We present a double blinding-attack on a quantum secret sharing (QSS) protocol based on GHZ state. Using the MDI-EW method, we propose a QSS protocol against all detector side-channels. We allow source flaws in practical QSS system, so that Charlie can securely distribute a key between the two agents Alice and Bob over long distances. Our protocol provides condition on the extracted key rate for the secret against both external eavesdropper and arbitrary dishonest participants. A tight bound for collective attacks can provide good bounds on the practical QSS with source flaws. Then we show through numerical simulations that using single-photon source a secure QSS over 136 km can be achieved.

Flaw tolerance vs. performance: A tradeoff in metallic glass cellular structures

International Nuclear Information System (INIS)

Chen, Wen; Liu, Ze; Robinson, Hannah Mae; Schroers, Jan

2014-01-01

Stochastic cellular structures are prevalent in nature and engineering materials alike. They are difficult to manipulate and study systematically and almost always contain imperfections. To design and characterize various degrees of imperfections in perfect periodic, stochastic and natural cellular structures, we fabricate a broad range of metallic glass cellular structures from perfectly periodic to highly stochastic by using a novel artificial microstructure approach based on thermoplastic replication of metallic glasses. For these cellular structures, precisely controlled imperfections are implemented and their effects on the mechanical response are evaluated. It is found that the mechanical performance of the periodic structures is generally superior to that of the stochastic structures. However, the stochastic structures experience a much higher tolerance to flaws than the periodic structure, especially in the plastic regime. The different flaw tolerance is explained by the stress distribution within the various structures, which leads to an overall 'strain-hardening' behavior of the stochastic structure compared to a 'strain-softening' behavior in the periodic structure. Our findings reveal how structure, 'strain-hardening' and flaw tolerance are microscopically related in structural materials

Repair boundary for parent tube indications within the upper joint zone of hybrid expansion joint (HEJ) sleeved tubes

International Nuclear Information System (INIS)

Cullen, W.K.; Keating, R.F.

1997-01-01

In the Spring and Fall of 1994, and the Spring of 1995, crack-like indications were found in the upper hybrid expansion joint (HEJ) region of Steam Generator (S/G) tubes which had been sleeved using Westinghouse HEJ sleeves. As a result of these findings, analytic and test evaluations were performed to assess the effect of the degradation on the structural, and leakage, integrity of the sleeve/tube joint relative to the requirements of the United States Nuclear Regulatory Commission's (NRC) draft Regulatory Guide (RG) 1.121. The results of these evaluations demonstrated that tubes with implied or known crack-like circumferential parent tube indications (PTIs) located 1.1 inches or farther below the bottom of the hardroll upper transition, have sufficient, and significant, integrity relative to the requirements of RG 1.121. Thus, the purpose of this report is to provide background information related to the justification of the modified tube repair boundary

Integration of finite element analysis and design of experiments to analyse the geometrical factors in bi-layered tube hydroforming

International Nuclear Information System (INIS)

Alaswad, A.; Olabi, A.G.; Benyounis, K.Y.

2011-01-01

Tube hydroforming (THF) is a type of unconventional metal forming process in which high fluid pressure and axial feed are used to deform a tube blank in the desired shape. Bi-layered tube hydroforming is suitable to produce bi-layered joints to be used in special applications such as aerospace, oil production and nuclear power plants. In this work, a finite element study along with response surface methodology (RSM) for design of experiment (DOE) has been used to construct models for three responses namely: bulge height, thickness reduction, and wrinkle height as a function of geometrical factors for X shape bi-layered tube hydroforming. A finite element model was built and experimentally validated. The models developed using finite element analysis (FEA) and RSM was found to be educated. The factors effect and their interactions on the three responses were determined and discussed. Such integration was proved to be a successful technique that can be used to predict the geometry of the hydroformed part.

Safety significance of steam generator tube degradation mechanisms

Energy Technology Data Exchange (ETDEWEB)

Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

1991-07-01

Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

A robust indicator based on singular value decomposition for flaw feature detection from noisy ultrasonic signals

Science.gov (United States)

Cui, Ximing; Wang, Zhe; Kang, Yihua; Pu, Haiming; Deng, Zhiyang

2018-05-01

Singular value decomposition (SVD) has been proven to be an effective de-noising tool for flaw echo signal feature detection in ultrasonic non-destructive evaluation (NDE). However, the uncertainty in the arbitrary manner of the selection of an effective singular value weakens the robustness of this technique. Improper selection of effective singular values will lead to bad performance of SVD de-noising. What is more, the computational complexity of SVD is too large for it to be applied in real-time applications. In this paper, to eliminate the uncertainty in SVD de-noising, a novel flaw indicator, named the maximum singular value indicator (MSI), based on short-time SVD (STSVD), is proposed for flaw feature detection from a measured signal in ultrasonic NDE. In this technique, the measured signal is first truncated into overlapping short-time data segments to put feature information of a transient flaw echo signal in local field, and then the MSI can be obtained from the SVD of each short-time data segment. Research shows that this indicator can clearly indicate the location of ultrasonic flaw signals, and the computational complexity of this STSVD-based indicator is significantly reduced with the algorithm proposed in this paper. Both simulation and experiments show that this technique is very efficient for real-time application in flaw detection from noisy data.

Reactor pressure vessel structural integrity research

International Nuclear Information System (INIS)

Pennell, W.E.; Corwin, W.R.

1994-01-01

Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

Fabrication and Characterization of All-Polystyrene Microfluidic Devices with Integrated Electrodes and Tubing.

Science.gov (United States)

Pentecost, Amber M; Martin, R Scott

2015-01-01

A new method of fabricating all-polystyrene devices with integrated electrodes and fluidic tubing is described. As opposed to expensive polystyrene (PS) fabrication techniques that use hot embossing and bonding with a heated lab press, this approach involves solvent-based etching of channels and lamination-based bonding of a PS cover, all of which do not need to occur in a clean room. PS has been studied as an alternative microchip substrate to PDMS, as it is more hydrophilic, biologically compatible in terms of cell adhesion, and less prone to absorption of hydrophobic molecules. The etching/lamination-based method described here results in a variety of all-PS devices, with or without electrodes and tubing. To characterize the devices, micrographs of etched channels (straight and intersected channels) were taken using confocal and scanning electron microscopy. Microchip-based electrophoresis with repetitive injections of fluorescein was conducted using a three-sided PS (etched pinched, twin-tee channel) and one-sided PDMS device. Microchip-based flow injection analysis, with dopamine and NO as analytes, was used to characterize the performance of all-PS devices with embedded tubing and electrodes. Limits of detection for dopamine and NO were 130 nM and 1.8 μM, respectively. Cell immobilization studies were also conducted to assess all-PS devices for cellular analysis. This paper demonstrates that these easy to fabricate devices can be attractive alternative to other PS fabrication methods for a wide variety of analytical and cell culture applications.

Imaging flaws in thin metal plates using a magneto-optic device

Science.gov (United States)

Wincheski, B.; Prabhu, D. R.; Namkung, M.; Birt, E. A.

1992-01-01

An account is given of the capabilities of the magnetooptic/eddy-current imager (MEI) apparatus in the case of aging aircraft structure-type flaws in 2024-T3 Al alloy plates. Attention is given to images of cyclically grown fatigue cracks from rivetted joints in fabricated lap-joint structures, electrical discharge machining notches, and corrosion spots. Although conventional eddy-current methods could have been used, the speed and ease of MEI's use in these tests is unmatched by such means. Results are displayed in real time as a test piece is scanned, furnishing easily interpreted flaw images.

A study on the extraction of feature variables for the pattern recognition for welding flaws

International Nuclear Information System (INIS)

Kim, J. Y.; Kim, C. H.; Kim, B. H.

1996-01-01

In this study, the researches classifying the artificial and natural flaws in welding parts are performed using the pattern recognition technology. For this purpose the signal pattern recognition package including the user defined function was developed and the total procedure including the digital signal processing, feature extraction, feature selection and classifier selection is treated by bulk. Specially it is composed with and discussed using the statistical classifier such as the linear discriminant function classifier, the empirical Bayesian classifier. Also, the pattern recognition technology is applied to classification problem of natural flaw(i.e multiple classification problem-crack, lack of penetration, lack of fusion, porosity, and slag inclusion, the planar and volumetric flaw classification problem). According to this results, if appropriately teamed the neural network classifier is better than stastical classifier in the classification problem of natural flaw. And it is possible to acquire the recognition rate of 80% above through it is different a little according to domain extracting the feature and the classifier.

Influence of circumferential flaw length on internal burst pressure of a wall-thinned pipe

Energy Technology Data Exchange (ETDEWEB)

Tsuji, Masataka, E-mail: tsuji-m@u-fukui.ac.jp [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui, Fukui (Japan); Meshii, Toshiyuki [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui, Fukui (Japan)

2013-02-15

Highlights: ► The effect of θ on p{sub f} was examined by experimental analysis and FEA. ► Here θ is the circumferential angle of a flaw, p{sub f} is the internal burst pressure. ► p{sub f} decreased as θ increased in some cases. ► The effect of θ on p{sub f} should be taken into consideration in evaluating p{sub f}. -- Abstract: This paper examines the effect of the circumferential angle of a flaw θ on the internal burst pressure p{sub f} of pipes with artificial wall-thinned flaws. The effect of θ has conventionally been regarded as unimportant in the evaluation of the p{sub f} of wall-thinned straight pipes. Therefore, a burst pressure equation for an axial crack inside a cylinder (Fig. 1, left), such as Kiefner's equation (Kiefner et al., 1973), has been widely applied (ANSI/ASME B31.G., 1991; Hasegawa et al., 2011). However, the following implicit assumptions notably exist when applying the equation to planar flaws in situations with non-planar flaws. 1)The fracture mode of the non-planar flaw under consideration is identical to that of the crack. 2)The effect of θ on p{sub f}, which is not considered for an axial crack, is small or negligible. However, the experimental results from the systematic burst tests for carbon steel pipes with artificial wall-thinned flaws examined in this paper showed that these implicit assumptions may be incorrect. In this paper the experimental results are evaluated in further detail. The purpose of the evaluation was to clarify the effect of θ on p{sub f}. Specifically, the significance of the flaw configuration (axial length δ{sub z} and wall-thinning ratio t{sub 1}/t) was studied for its effects on θ and p{sub f}. In addition, a simulation of this effect was conducted using a large strain elastic-plastic Finite Element Analysis (FEA) model. As observed from the experimental results, θ tended to affect p{sub f} in cases with large δ{sub z}, and t{sub 1}/t was also correlated with a decrease in p{sub f

Applying computer modeling to eddy current signal analysis for steam generator and heat exchanger tube inspections

International Nuclear Information System (INIS)

Sullivan, S.P.; Cecco, V.S.; Carter, J.R.; Spanner, M.; McElvanney, M.; Krause, T.W.; Tkaczyk, R.

2000-01-01

Licensing requirements for eddy current inspections for nuclear steam generators and heat exchangers are becoming increasingly stringent. The traditional industry-standard method of comparing inspection signals with flaw signals from simple in-line calibration standards is proving to be inadequate. A more complete understanding of eddy current and magnetic field interactions with flaws and other anomalies is required for the industry to generate consistently reliable inspections. Computer modeling is a valuable tool in improving the reliability of eddy current signal analysis. Results from computer modeling are helping inspectors to properly discriminate between real flaw signals and false calls, and improving reliability in flaw sizing. This presentation will discuss complementary eddy current computer modeling techniques such as the Finite Element Method (FEM), Volume Integral Method (VIM), Layer Approximation and other analytic methods. Each of these methods have advantages and limitations. An extension of the Layer Approximation to model eddy current probe responses to ferromagnetic materials will also be presented. Finally examples will be discussed demonstrating how some significant eddy current signal analysis problems have been resolved using appropriate electromagnetic computer modeling tools

Characterization of type, position and dimension of flaws by transit time locus curves of ultrasonic inspections - ALOK. Pt. 2

International Nuclear Information System (INIS)

Grohs, B.; Barbian, O.A.; Kappes, W.; Paul, H.

1981-01-01

With automatic ultrasonic testing, flaws can be detected and described and thus characterized according to their type, position and dimensions. During scanning of a test object, the flaws are registered by many different pathways and many different acoustic irradiation directions. The transit time locus curve represents the distance between the relfecting points of a flaw and the source in dependence of the probe position; hence, information on flaw position and dimensions can be derived from this curve. If the sound velocity is known, the transit path can then be calculated from the transit time. This requires, above all, a constant sound velocity along the whole transit path. Various methods are presented for reconstructing the flaw border in the plane of incidence. (orig./RW) [de

Size determinations, by ultrasonic techniques, of cracks in hydride blisters formed in Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Trujillo Badillo, Giovanna; Desimone, Carlos; Domizzi, Gladys

1999-01-01

Non destructive techniques (NDT) are very useful in the detection of flaws produced in structural components in service. During the service of CANDU nuclear power reactors, it is possible that pressure tubes (PT) may contact calandria tubes (CT). After the PT/CT contact, zirconium hydride blisters may form at the point of contact depending on the concentration of hydrogen/deuterium. Zirconium hydride is brittle and is therefore prone to cracking under stress. Ultrasonic NDT is routinely use during PT in service inspection. In order to be able of detecting cracked blisters, it is of great importance the development of standards to calibrate the employed equipment. On this purpose, hydride blisters were grown, in laboratory, on sections of pressure tube. The cracks in the blisters were detected and measured by ultrasonic techniques. The obtained results were compared with measurements carried out in optic microscope, on successive sections of the samples. The crack tip diffraction technique was found to be the more effective for the mentioned ends. (author)

Detection of plane, poorly oriented wide flaws using focused transducers

International Nuclear Information System (INIS)

Vadder, D. de; Azou, P.; Bastien, P.; Saglio, R.

1976-01-01

The detection of plane, poorly oriented, wide flaws by ultrasonic non destructive testing is distinctly improved when using focused transducers. An increased echo can be obtained crossing the defect limit [fr

Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

International Nuclear Information System (INIS)

Crespi, J.C.

1992-01-01

This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

CLEAVAGE FRACTURE ANALYSIS OF CLADDED BEAMS WITH AN EMBEDDED FLAW UNDER FOUR-POINT BENDING

International Nuclear Information System (INIS)

Yin, Shengjun; Williams, Paul T; Bass, Bennett Richard

2008-01-01

Semi-large scale embedded flaw beams were tested at Nuclear Research Institute (NRI) Rez in the Czech Republic for the 6th Network for Evaluating Structural Components (NESC-VI) project. The experiments included, among others, a series of semi-large scale tests on cladded beam specimens containing simulated sub-clad flaws. Oak Ridge National Laboratory (ORNL) conducted numerical studies to analyze the constraint issues associated with embedded flaws using various fracture mechanics methods, including T-Stress, hydrostatic stress based QH stress, and the Weibull stress model. The recently developed local approach using the modified Weibull stress model combined with the Master Curve methodology was also utilized to predict the failure probability (Pf) of semi-large scale beams. For this study, the Weibull statistical model associated with the Master Curve methodology was employed to stochastically simulate the fracture toughness data using the available Master Curve reference temperature T0 for the tested base material from the 'aged' WWER-440 Reactor Pressure Vessel (RPV). The study was also conducted to investigate the sensitivity of predicated probability of failure of semi-large scale beams with embedded flaw with different Weibull shape parameters, m

Creating a YouTube-Like Collaborative Environment in Mathematics: Integrating Animated Geogebra Constructions and Student-Generated Screencast Videos

Science.gov (United States)

Lazarus, Jill; Roulet, Geoffrey

2013-01-01

This article discusses the integration of student-generated GeoGebra applets and Jing screencast videos to create a YouTube-like medium for sharing in mathematics. The value of combining dynamic mathematics software and screencast videos for facilitating communication and representations in a digital era is demonstrated herein. We share our…

Trial manufacture of simple integrated tube-type pyranometer by phycoerythrin and measurements of transmittance of solar radiation in crop canopies

International Nuclear Information System (INIS)

Yamamoto, H.; Honjo, H.; Kamota, F.; Suzuki, Y.; Hayakawa, S.

1998-01-01

We tried to construct a simple integrated tube-type pyranometer using phycoerythrin from seaweed pigment. The maximum sensitive wavehand of phycoerythrin was 550 nm - 560 nm, and this waveband was in the photosynthetically active radiation range. The acrylic tubes (outside diameter, 22 mm, length, 100 cm) were spread with white paints except for a strip 15 mm in width, and phycoerythrin was put into the acrylic tube. In the results from the outdoor measurements, the tube-type pyranometer showed a positive correlation between the transmittance of phycoerythrin (%) and the measured accumulated solar radiation (MJ n(-2)), but the slope of the linear equation was different in summer and winter. In an artificial climate room, the relationship between the transmissions of phycoerythrin and the accumulated solar radiation could be approximated by a quadratic equation at every temperature. In the measurements made outdoors, the accumulated solar radiation could be estimated using the transmittance of phycoerythrin and the mean air temperature during measurements

Vitamin D and depression: a systematic review and meta-analysis comparing studies with and without biological flaws.

Science.gov (United States)

Spedding, Simon

2014-04-11

Efficacy of Vitamin D supplements in depression is controversial, awaiting further literature analysis. Biological flaws in primary studies is a possible reason meta-analyses of Vitamin D have failed to demonstrate efficacy. This systematic review and meta-analysis of Vitamin D and depression compared studies with and without biological flaws. The systematic review followed the Preferred Reporting Items for Systematic Reviews and Meta-Analyses (PRISMA) guidelines. The literature search was undertaken through four databases for randomized controlled trials (RCTs). Studies were critically appraised for methodological quality and biological flaws, in relation to the hypothesis and study design. Meta-analyses were performed for studies according to the presence of biological flaws. The 15 RCTs identified provide a more comprehensive evidence-base than previous systematic reviews; methodological quality of studies was generally good and methodology was diverse. A meta-analysis of all studies without flaws demonstrated a statistically significant improvement in depression with Vitamin D supplements (+0.78 CI +0.24, +1.27). Studies with biological flaws were mainly inconclusive, with the meta-analysis demonstrating a statistically significant worsening in depression by taking Vitamin D supplements (-1.1 CI -0.7, -1.5). Vitamin D supplementation (≥800 I.U. daily) was somewhat favorable in the management of depression in studies that demonstrate a change in vitamin levels, and the effect size was comparable to that of anti-depressant medication.

Self-shielding flex-circuit drift tube, drift tube assembly and method of making

Science.gov (United States)

Jones, David Alexander

2016-04-26

The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.

Fracture assessment of shallow-flaw cruciform beams tested under uniaxial and biaxial loading conditions

International Nuclear Information System (INIS)

Bass, B.R.; McAfee, W.J.; Williams, P.T.; Pennell, W.E.

1999-01-01

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow, surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a far-field, out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure-temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for an RPV material. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies, namely, the J-Q formulation, the Dodds-Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness; the conventional maximum principal stress criterion indicated no effect. A three-parameter Weibull model based on the hydrostatic stress criterion is shown to correlate with the experimentally observed biaxial effect on cleavage fracture toughness by providing a scaling mechanism between uniaxial and biaxial loading states. (orig.)

Stress analysis of HTR-10 steam generator heat exchanging tubes

International Nuclear Information System (INIS)

Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

2001-01-01

Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

Experimental evaluation of emergency operating procedures on multiple steam generator tube rupture in INER integral system test facility

International Nuclear Information System (INIS)

Liu, T.J.; Lin, Y.M.; Lee, C.H.; Chang, C.Y.; Hong, W.T.

1997-01-01

The multiple steam generator tube rupture (SGTR) scenario in Westinghouse type pressurized water reactor (PWR) has been investigated at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. This reduced-height and reduced-pressure test facility was designed to simulate the main features of Maanshan nuclear power plant. The SGTR test scenario assumes the double-ended break of one-, two- and six- tubes without other failures. The major operator actions follow the related symptom-oriented Emergency Operating Procedure (EOP) on the reference plant. This study focuses on the investigation of thermal-hydraulics phenomena and the adequacy of associated EOP to limit primary-to-secondary leakage. Through this study, it is found that the adequacy of current EOP in minimizing the radioactivity release demands early substantial operator involvement, especially in the multi-tubes break events. Also, the detailed mechanism of the main thermal-hydraulic phenomena during the SGTR transient are explored. (author)

Detecting and Preventing Type flaws at Static Time

DEFF Research Database (Denmark)

Bodei, Chiara; Brodo, Linda; Degano, Pierpaolo

2010-01-01

A type flaw attack on a security protocol is an attack where an honest principal is cheated on interpreting a field in a message as the one with a type other than the intended one. In this paper, we shall present an extension of the LYSA calculus to cope with types, by using tags to represent...

Reconstructing flaw image using dataset of full matrix capture technique

Energy Technology Data Exchange (ETDEWEB)

Lee, Tae Hun; Kim, Yong Sik; Lee, Jeong Seok [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2017-02-15

A conventional phased array ultrasonic system offers the ability to steer an ultrasonic beam by applying independent time delays of individual elements in the array and produce an ultrasonic image. In contrast, full matrix capture (FMC) is a data acquisition process that collects a complete matrix of A-scans from every possible independent transmit-receive combination in a phased array transducer and makes it possible to reconstruct various images that cannot be produced by conventional phased array with the post processing as well as images equivalent to a conventional phased array image. In this paper, a basic algorithm based on the LLL mode total focusing method (TFM) that can image crack type flaws is described. And this technique was applied to reconstruct flaw images from the FMC dataset obtained from the experiments and ultrasonic simulation.

An corrective method to correct of the inherent flaw of the asynchronization direct counting circuit

International Nuclear Information System (INIS)

Wang Renfei; Liu Congzhan; Jin Yongjie; Zhang Zhi; Li Yanguo

2003-01-01

As a inherent flaw of the Asynchronization Direct Counting Circuit, the crosstalk, which is resulted from the randomicity of the time-signal always exists between two adjacent channels. In order to reduce the counting error derived from the crosstalk, the author propose an effective method to correct the flaw after analysing the mechanism of the crosstalk

Capability evaluation of Eddy current and ultrasonic in-service inspections of steam generator tubes. A status report of PISC III Action 5

International Nuclear Information System (INIS)

Bieth, M.; Birac, C.; Comby, R.

1998-01-01

Document summarizes the PISC III (Programme for the Inspection of Steel Components) report No. 41, full description of the PISC III Action 5 on Steam Generator Tubes Inspection, containing all details and final conclusions which are still to be approved by the PISC III Management Board. The report was prepared by the reference laboratory of PISC under guidance and with continuous contribution of the members of the Data Analysis Group (DAG) of this PISC III. There were several procedures which demonstrated good detection capability of major flaws in typical locations of the steam generator. Conclusions of the exercise indicate that capability demonstration is necessary to qualify in service inspection procedures for steam generator tubes

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2006-01-01

Considering a rational maintenance rule of Light Water Reactor piping, reliable flaw evaluation criteria are essential to determine how a detected flaw is detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes to be considered for carbon steel piping, and can be analyzed by the elastic-plastic fracture mechanics. Some analytical efforts have been provided as flaw evaluation criteria using load correction factors such like the Z-factors in the JSME codes on fitness-for-service for nuclear power plants or the ASME boiler and pressure vessel code section XI. The present correction factors were conventionally determined taken conservatism and simplicity into account, however, the effect of internal pressure which would be an important factor under an actual plant condition was not adequately considered. Recently, a J-estimation scheme, 'LBB. ENGC' for ductile fracture analysis of circumferentially through-wall-cracked pipes subjected combined loading was newly developed to have a better prediction with more realistic manner. This method is explicitly incorporated the contribution of both bending and tension due to internal pressure by means of the scheme compatible with an arbitrary combined loading history. In this paper, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. A correction factor based on the new J-estimation scheme was compared with the present correction factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of internal pressure. (author)

Flaw evaluation of thermally aged cast stainless steel in light-water reactor applications

International Nuclear Information System (INIS)

Lee, S.; Kuo, P.T.; Wichman, K.; Chopra, O.

1997-01-01

Cast stainless steel may be used in the fabrication of the primary loop piping, fittings, valve bodies, and pump casings in light-water reactors. However, this material is subject to embrittlement due to thermal aging at the reactor temperature, that is 290 o C (550 o F). The Argonne National Laboratory (ANL) recently completed a research program and the results indicate that the lower-bound fracture toughness of thermally aged cast stainless steel is similar to that of submerged arc welds (SAWs). Thus, the US Nuclear Regulatory Commission (NRC) staff has accepted the use of SAW flaw evaluation procedures in IWB-3640 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code to evaluate flaws in thermally aged cast stainless steel for a license renewal evaluation. Alternatively, utilities may estimate component-specific fracture toughness of thermally aged cast stainless steel using procedures developed at ANL for a case-by-case flaw evaluation. (Author)

Gamma flaw detectors for radiographic control of welded joint quality under mounting conditions

International Nuclear Information System (INIS)

Khoroshev, V.N.; Galash, T.F.; Andreev, V.L.; Grigor'ev, V.M.; Medvedev, N.E.

1978-01-01

Main characteristics are presented of gamma flaw detector models used for radiographic control of the quality of welded steel and pipeline joints during assembly. Specially developed experimental models, operating with 75 Se, 90 Sr, 170 Tm, 137 Cs and 192 Ir sources are considered. The new instruments have been made on a single structural base, which creates a foundation for standardizing individual units of radiation heads, manual control panels, containers, exterior packings, devices and accessories, maintenance techniques, and repair techniques. They are distinguished by small sizes and weight, possibility of using a set of radiation sources ensuring control of 3-40 mm thick joints, and reliable protection. Special devices permit to reduce 2-3-folds the time needed for installing and orienting the flaw detectors. The expected economic effect from implementation of the new gamma flaw detectors into industry will amount to 1.5-10.0 thousand roubles per annum for one detector at approximate cost of each detector equal to 3.5-6.0 thousand roubles

Vitamin D and Depression: A Systematic Review and Meta-Analysis Comparing Studies with and without Biological Flaws

Directory of Open Access Journals (Sweden)

Simon Spedding

2014-04-01

Full Text Available Efficacy of Vitamin D supplements in depression is controversial, awaiting further literature analysis. Biological flaws in primary studies is a possible reason meta-analyses of Vitamin D have failed to demonstrate efficacy. This systematic review and meta-analysis of Vitamin D and depression compared studies with and without biological flaws. The systematic review followed the Preferred Reporting Items for Systematic Reviews and Meta-Analyses (PRISMA guidelines. The literature search was undertaken through four databases for randomized controlled trials (RCTs. Studies were critically appraised for methodological quality and biological flaws, in relation to the hypothesis and study design. Meta-analyses were performed for studies according to the presence of biological flaws. The 15 RCTs identified provide a more comprehensive evidence-base than previous systematic reviews; methodological quality of studies was generally good and methodology was diverse. A meta-analysis of all studies without flaws demonstrated a statistically significant improvement in depression with Vitamin D supplements (+0.78 CI +0.24, +1.27. Studies with biological flaws were mainly inconclusive, with the meta-analysis demonstrating a statistically significant worsening in depression by taking Vitamin D supplements (−1.1 CI −0.7, −1.5. Vitamin D supplementation (≥800 I.U. daily was somewhat favorable in the management of depression in studies that demonstrate a change in vitamin levels, and the effect size was comparable to that of anti-depressant medication.

Detecting and revising flaws in OWL object property expressions

CSIR Research Space (South Africa)

Keet, CM

2012-10-01

Full Text Available to the ontologist's intention. However, the more one can do, the higher the chance modelling flaws are introduced; hence, an unexpected or undesired classification or inconsistency may actually be due to a mistake in the object property box, not the class axioms. We...

Acoustic emission/flaw relationships for inservice monitoring of LWRs

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Friesel, M.A.; Skorpik, J.R.; Dawson, J.F.

1991-10-01

The program concerning Acoustic Emission/Flaw Relationships for Inservice Monitoring of LWRs was initiated in FY76 with the objective of validating the application of acoustic emission (AE) to monitor nuclear reactor pressure-containing components during operation to detect cracking. The program has been supported by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Research and development has been performed by Pacific Northwest Laboratory, operated for the Department of Energy by Battelle Memorial Institute. The program has shown the feasibility of continuous, on-line AE monitoring to detect crack growth and produced validated methods for applying the technology. Included are relationships for estimating flaw severity from AE data and field applications at Watts Bar Unit 1 Reactor, Limerick Unit 1 Reactor, and the High Flux Isotope Reactor. This report discusses the program scope and organization, the three program phases and the results obtained, standard and code activities, and instrumentation and software developed under this program

Failure analysis of retired steam generator tubings

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

2005-04-15

Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

Development of technology on the material surveillance of CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author).

Development of technology on the material surveillance of CANDU pressure tubes

International Nuclear Information System (INIS)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author)

Probabilistic Estimation of Critical Flaw Sizes in the Primary Structure Welds of the Ares I-X Launch Vehicle

Science.gov (United States)

Pai, Shantaram S.; Hoge, Peter A.; Patel, B. M.; Nagpal, Vinod K.

2009-01-01

The primary structure of the Ares I-X Upper Stage Simulator (USS) launch vehicle is constructed of welded mild steel plates. There is some concern over the possibility of structural failure due to welding flaws. It was considered critical to quantify the impact of uncertainties in residual stress, material porosity, applied loads, and material and crack growth properties on the reliability of the welds during its pre-flight and flight. A criterion--an existing maximum size crack at the weld toe must be smaller than the maximum allowable flaw size--was established to estimate the reliability of the welds. A spectrum of maximum allowable flaw sizes was developed for different possible combinations of all of the above listed variables by performing probabilistic crack growth analyses using the ANSYS finite element analysis code in conjunction with the NASGRO crack growth code. Two alternative methods were used to account for residual stresses: (1) The mean residual stress was assumed to be 41 ksi and a limit was set on the net section flow stress during crack propagation. The critical flaw size was determined by parametrically increasing the initial flaw size and detecting if this limit was exceeded during four complete flight cycles, and (2) The mean residual stress was assumed to be 49.6 ksi (the parent material s yield strength) and the net section flow stress limit was ignored. The critical flaw size was determined by parametrically increasing the initial flaw size and detecting if catastrophic crack growth occurred during four complete flight cycles. Both surface-crack models and through-crack models were utilized to characterize cracks in the weld toe.

Centrifugal LabTube platform for fully automated DNA purification and LAMP amplification based on an integrated, low-cost heating system.

Science.gov (United States)

Hoehl, Melanie M; Weißert, Michael; Dannenberg, Arne; Nesch, Thomas; Paust, Nils; von Stetten, Felix; Zengerle, Roland; Slocum, Alexander H; Steigert, Juergen

2014-06-01

This paper introduces a disposable battery-driven heating system for loop-mediated isothermal DNA amplification (LAMP) inside a centrifugally-driven DNA purification platform (LabTube). We demonstrate LabTube-based fully automated DNA purification of as low as 100 cell-equivalents of verotoxin-producing Escherichia coli (VTEC) in water, milk and apple juice in a laboratory centrifuge, followed by integrated and automated LAMP amplification with a reduction of hands-on time from 45 to 1 min. The heating system consists of two parallel SMD thick film resistors and a NTC as heating and temperature sensing elements. They are driven by a 3 V battery and controlled by a microcontroller. The LAMP reagents are stored in the elution chamber and the amplification starts immediately after the eluate is purged into the chamber. The LabTube, including a microcontroller-based heating system, demonstrates contamination-free and automated sample-to-answer nucleic acid testing within a laboratory centrifuge. The heating system can be easily parallelized within one LabTube and it is deployable for a variety of heating and electrical applications.

A Study on the Profile Change Measurement of Steam Generator Tubes with Tube Expansion Methods

International Nuclear Information System (INIS)

Kim, Young Kyu; Song Myung Ho; Choi, Myung Sik

2011-01-01

Steam generator tubes for nuclear power plants contain the local shape transitions on their inner or outer surface such as dent, bulge, over-expansion, eccentricity, deflection, and so on by the application of physical force during the tube manufacturing and steam generator assembling and by the sludge (that is, corrosion products) produced during the plant operation. The structural integrity of tubes will be degraded by generating the corrosive crack at that location. The profilometry using the traditional bobbin probes which are currently applied for measuring the profile change of tubes gives us basic information such as axial locations and average magnitudes of deformations. However, the three-dimensional quantitative evaluation on circumferential locations, distributional angle, and size of deformations will have to be conducted to understand the effects of residual stresses increased by local deformations on corrosive cracking of tubes. Steam generator tubes of Korean standard nuclear power plants expanded within their tube-sheets by the explosive expansion method and suffered from corrosive cracks in the early stage of power operation. Thus, local deformations of steam generator tubes at the top of tube-sheet were measured with an advanced rotating probe and a laser profiling system for the two cases where the tubes expanded by the explosive expansion method and hydraulic expansion. Also, the trends of eccentricity, deflection, and over-expansion of tubes were evaluated. The advanced eddy current profilometry was confirmed to provide accurate information of local deformations compared with laser profilometry

Development of flaw evaluation and acceptance procedures for flaw indications in the cooling water system at the Savannah River site K reactor

International Nuclear Information System (INIS)

Tandon, S.; Bamford, W.H.; Cowfer, C.D.; Ostrowski, R.

1993-01-01

This paper describes the methodology used in determining the criteria for acceptance of inspection indications in the K-Reactor Cooling Water System at the Savannah River Plant. These criteria have been developed in a manner consistent with the development of similar criteria in the ASME Code Section XI for commercial light water reactors, but with a realistic treatment of the operating conditions in the cooling water system. The technical basis for the development of these criteria called ''Acceptance Standards'' is contained in this paper. A second portion of this paper contains the methodology used in the construction of flaw evaluation charts which have been developed for each specific line size in the cooling water system. The charts provide the results of detailed fracture mechanics calculations which have been completed to determine the largest flaw which can be accepted in the cooling water system without repair. These charts are designed for use in conjunction with in-service inspections of the cooling water system, and only require inspection results to determine acceptability

Effect of combined loading due to bending and internal pressure on pipe flaw evaluation criteria

International Nuclear Information System (INIS)

Miura, Naoki; Sakai, Shinsuke

2008-01-01

Considering a rule for the rationalization of maintenance of Light Water Reactor piping, reliable flaw evaluation criteria are essential for determining how a detected flaw will be detrimental to continuous plant operation. Ductile fracture is one of the dominant failure modes that must be considered for carbon steel piping and can be analyzed by elastic-plastic fracture mechanics. Some analytical efforts have provided various flaw evaluation criteria using load correction factors, such as the Z-factors in the JSME codes on fitness-for-service for nuclear power plants and the section XI of the ASME boiler and pressure vessel code. The present Z-factors were conventionally determined, taking conservativity and simplicity into account; however, the effect of internal pressure, which is an important factor under actual plant conditions, was not adequately considered. Recently, a J-estimation scheme, LBB.ENGC for the ductile fracture analysis of circumferentially through-wall-cracked pipes subjected to combined loading was developed for more accurate prediction under more realistic conditions. This method explicitly incorporates the contributions of both bending and tension due to internal pressure by means of a scheme that is compatible with an arbitrary combined-loading history. In this study, the effect of internal pressure on the flaw evaluation criteria was investigated using the new J-estimation scheme. The Z-factor obtained in this study was compared with the presently used Z-factors, and the predictability of the current flaw evaluation criteria was quantitatively evaluated in consideration of the internal pressure. (author)

Stepped frequency imaging for flaw monitoring: Final report

International Nuclear Information System (INIS)

Hildebrand, B.P.

1988-09-01

This report summarizes the results of research into the usefulness of stepped frequency imaging (SFI) to nuclear power plant inspection. SFI is a method for producing ultrasonic holographic images without the need to sweep a two-dimensional aperture with the transducer. Instead, the transducer may be translated along a line. At each position of the transducer the frequency is stepped over a finite preselected bandwidth. The frequency stepped data is then processed to synthesize the second dimension. In this way it is possible to generate images in regions that are relatively inaccessible to two-dimensional scanners. This report reviews the theory and experimental work verifying the technique, and then explores its possible applications in the nuclear power industry. It also outlines how this new capability can be incorporated into the SDL-1000 Imaging System previously developed for EPRI. The report concludes with five suggestions for uses for the SFI method. These are: monitoring suspect or repaired regions of feedwater nozzles; monitoring pipe cracks repaired by weld overlay; monitoring crack depth during test block production; imaging flaws where access is difficult; and imaging flaws through cladding without distortion

An integrated leak detection system for the ALMR steam generator

International Nuclear Information System (INIS)

Dayal, Y.; Gaubatz, D.C.; Wong, K.K.; Greene, D.A.

1995-01-01

The steam generator (SG) of the Advanced Liquid Metal Reactor (ALMR) system serves as a heat exchanger between the shell side secondary loop hot liquid sodium and the tube side water/steam mixture. A leak in the tube will result in the injection of the higher pressure water/steam into the sodium and cause an exothermic sodium-water reaction. An initial small leak (less than 1 gm/sec) can escalate into an intermediate size leak in a relatively short time by self enlargement of the original flaw and by initiating leaks in neighboring tubes. If not stopped, complete rupture of one or more tubes can cause injection rates of thousands of gm/sec and result in the over pressurization of the secondary loop rupture disk and dumping of the sodium to relieve pressure. The down time associated with severe sodium-water reaction damage has great adverse economic consequence. An integrated leak detection system (ILDS) has been developed which utilizes both chemical and acoustic sensors for improved leak detection. The system provides SG leak status to the reactor operator and is reliable enough to trigger automatic control action to protect the SG. The ILDS chemical subsystem uses conventional in-sodium and cover gas hydrogen detectors and incorporates knowledge based effects due to process parameters for improved reliability. The ILDS acoustic subsystem uses an array of acoustic sensors and incorporates acoustic beamforming technology for highly reliable and accurate leak identification and location. The new ILDS combines the small leak detection capability of the chemical system with the reliability and rapid detection/location capability of the acoustic system to provide a significantly improved level of protection for the SG over a wide range of operation conditions. (author)

High temperature ceramic-tubed reformer

Science.gov (United States)

Williams, Joseph J.; Rosenberg, Robert A.; McDonough, Lane J.

1990-03-01

The overall objective of the HiPHES project is to develop an advanced high-pressure heat exchanger for a convective steam/methane reformer. The HiPHES steam/methane reformer is a convective, shell and tube type, catalytic reactor. The use of ceramic tubes will allow reaction temperature higher than the current state-of-the-art outlet temperatures of about 1600 F using metal tubes. Higher reaction temperatures increase feedstock conversion to synthesis gas and reduce energy requirements compared to currently available radiant-box type reformers using metal tubes. Reforming of natural gas is the principal method used to produce synthesis gas (primarily hydrogen and carbon monoxide, H2 and CO) which is used to produce hydrogen (for refinery upgrading), methanol, as well as several other important materials. The HiPHES reformer development is an extension of Stone and Webster's efforts to develop a metal-tubed convective reformer integrated with a gas turbine cycle.

A Flue Gas Tube for Thermoelectric Generator

DEFF Research Database (Denmark)

2013-01-01

The invention relates to a flue gas tube (FGT) (1) for generation of thermoelectric power having thermoelectric elements (8) that are integrated in the tube. The FTG may be used in combined heat and power (CHP) system (13) to produce directly electricity from waste heat from, e.g. a biomass boiler...

PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

International Nuclear Information System (INIS)

Mattar Neto, Miguel; Cruz, Julio R.B.

1999-01-01

One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

Preliminary Stress Analysis of an IHX Tube Support Plate in Prototype SFR

International Nuclear Information System (INIS)

Kim, Sung Kyun; Koo, Gyeong Hoi

2013-01-01

In this paper, the structural integrity about the conceptual design of IHX tube support plate was reviewed and the design should be changed because of its high stress concentration at the outer rim area. For reducing its maximum stress, two alternatives were proposed and reviewed for the structural integrity point of view. In both proposing support designs, the maximum stress decreases up to the stress design limit. Tube support plates (TSPs) of the intermediate heat exchanger (IHX) in Prototype GenIV Sodium Cooled Fast Reactor (PGSFR) act to horizontally support IHX tubes against hydraulic loadings and they have numerous flow holes where a primary sodium flows downward and secondary sodium flows upward. Due to its many penetrations, its geometric shape is quite complex and structurally its integrity is quite weaker than other parts. In this study, we investigated the structural integrity of the conceptually designed IHX tube support plate. In addition, TSP's supporting concepts were proposed to increase its structural integrity, and confirmed its integrity by using a finite element analysis

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

Detection and characterization of flaws in segments of light water reactor pressure vessels

International Nuclear Information System (INIS)

Cook, K.V.; Cunningham, R.A. Jr.; McClung, R.W.

1988-01-01

Studies have been conducted to determine flaw density in segments cut from light water reactor )LWR) pressure vessels as part of the Oak Ridge National Laboratory's Heavy-Section Steel Technology (H SST) Program. Segments from the Hope Creek Unit 2 vessel and the Pilgrim Unit 2 Vessel were purchased from salvage dealers. Hope Creek was a boiling water reactor (BWR) design and Pilgrim was a pressurized water reactor (PWR) design. Neither were ever placed in service. Objectives were to evaluate these LWR segments for flaws with ultrasonic and liquid penetrant techniques. Both objectives were successfully completed. One significant indication was detected in a Hope Creek seam weld by ultrasonic techniques and characterized by further analyses terminating with destructive correlation. This indication [with a through-wall dimension of ∼6 mm (∼0.24 in.)] was detected in only 3 m (10 ft) of weldment and offers extremely limited data when compared to the extent of welding even in a single pressure vessel. However, the detection and confirmation of the flaw in the arbitrarily selected sections implies the Marshall report estimates (and others) are nonconservative for such small flaws. No significant indications were detected in the Pilgrim material by ultrasonic techniques. Unfortunately, the Pilgrim segments contained relatively little weldment; thus, we limited our ultrasonic examinations to the cladding and subcladding regions. Fluorescent liquid penetrant inspection of the cladding surfaces for both LWR segments detected no significant indications [i.e., for a total of approximately 6.8 m 2 (72 ft 2 ) of cladding surface]. (author)

Full length channel Pressure Tube sagging under completely voided full length pressure tube of an Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Negi, Sujay, E-mail: negi.sujay@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Kumar, Ravi, E-mail: ravikfme@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukopadhyay, D., E-mail: dmukho@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India)

2017-03-15

Highlights: • At 16 kW/m input, thermal stability was attained at 595 °C, without PT-CT contact. • At 20 kW/m step input, PT-CT contact occurred at 637 °C near bottom-center of the tube. • PT integrity was maintained throughout the experiment. - Abstract: An experimental investigation was conducted to simulate the sagging behavior of a full length Pressure Tube of a channel of 220 MWe Indian PHWR. The investigation aimed to recreate a condition resembling Loss of Coolant Accident (LOCA) with Emergency Core Cooling System (ECCS) failure in a nuclear power plant. A full length channel assembly immersed in moderator was subjected to electrical resistance heating of Pressure Tube (PT) to simulate the residual heat after shutting down of reactor. The temperature of PT started rising and the contact between PT and CT was established at the center of the tube where average bottom temperature was 637 °C. The integrity of PT was maintained throughout the experiment and the PT heat up was arrested on contact with the CT due to transfer of heat to the moderator.

Study on simplified estimation of J-integral under thermal loading

International Nuclear Information System (INIS)

Takahashi, Y.

1993-01-01

For assessing structural integrity or safety of nuclear power plants, strength of structures under the presence of flaws sometimes needs to be evaluated. Because relative large inelastic deformation is anticipated in the liquid metal reactor components even without flaws due to high operating temperature and large temperature gradients, inelastic effects should be properly taken into account in the flaw assessment procedures. It is widely recognized that J-integral and its variations - e.g. fatigue J-integral range and creep J-integral - play substantial roles in the flaw assessment under the presence of large inelastic deformation. Therefore their utilization has been promoted in the recent flaw assessment procedure both for low and high temperature plants. However, it is not very practical to conduct a detailed numerical computation for cracked structures to estimate the values of these parameters for the purpose of trailing crack growth history. Thus development of simplified estimation methods which do not require full numerical calculation for cracked structures is desirable. A method using normalized J-integral solutions tabulated in the handbook is a direct extension of linear fracture mechanics counterpart and it can be used for standard specimen and simple structural configurations subjected to specified loading type. The reference stress method has also been developed but in this case limit load solutions, which are often difficult to obtain for general stress distribution, are necessary instead of nonlinear J-integral solutions. However, both methods have been developed mainly for mechanical loading and thus applying these techniques to thermal stress problem is rather difficult except the cases where the thermal stress can be properly substituted by equivalent mechanical loading as in the case of simple thermal expansion loading. Therefore alternative approach should be pursued for estimating J-integral and their variations in thermal stress problems

Evaluation of constraint methodologies applied to a shallow-flaw cruciform bend specimen tested under biaxial loading conditions

International Nuclear Information System (INIS)

Bass, B.R.; McAfee, W.J.; Williams, P.T.; Pennell, W.E.

1998-01-01

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a prototypic, far-field. out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure-temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for RPV materials. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies. namely, the J-Q formulation, the Dodds-Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness, the conventional maximum principal stress criterion indicated no effect

According to Jim: The Flawed Normal Curve of Intelligence

Science.gov (United States)

Gallagher, James J.

2008-01-01

In this article, the author talks about the normal curve of intelligence which he thinks is flawed and contends that wrong conclusions have been drawn based on this spurious normal curve. An example is that of racial and ethnic differences wherein some authors maintain that some ethnic and racial groups are clearly superior to others based on…

Evaluation of Manual Ultrasonic Examinations Applied to Detect Flaws in Primary System Dissimilar Metal Welds at North Anna Power Station

International Nuclear Information System (INIS)

Anderson, Michael T.; Diaz, Aaron A.; Doctor, Steven R.

2012-01-01

During a recent inservice inspection (ISI) of a dissimilar metal weld (DMW) in an inlet (hot leg) steam generator nozzle at North Anna Power Station Unit 1, several axially oriented flaws went undetected by the licensee's manual ultrasonic testing (UT) technique. The flaws were subsequently detected as a result of outside diameter (OD) surface machining in preparation for a full structural weld overlay. The machining operation uncovered the existence of two through-wall flaws, based on the observance of primary water leaking from the DMW. Further ultrasonic tests were then performed, and a total of five axially oriented flaws, classified as primary water stress corrosion cracking (PWSCC), were detected in varied locations around the weld circumference.

Signal to noise ratio enhancement for Eddy Current testing of steam generator tubes in PWR's

International Nuclear Information System (INIS)

Georgel, B.

1985-01-01

Noise reduction is a compulsory task when we try to recognize and characterize flaws. The signals we deal with come from Eddy Current testings of steam generator steel tubes. We point out the need for a spectral invariant in digital spectral analysis of 2 components signals. We make clear the pros and cons of classical passband filtering and suggest the use of a new noise cancellation method first discussed by Moriwaki and Tlusty. We generalize this tricky technique and prove it is a very special case of the well-known Wiener filter. In that sense the M-T method is shown to be optimal. 6 refs

Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

Energy Technology Data Exchange (ETDEWEB)

Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

Gastrostomy Tube (G-Tube)

Science.gov (United States)

... any of these problems: a dislodged tube a blocked or clogged tube any signs of infection (including redness, swelling, or warmth at the tube site; discharge that's yellow, green, or foul-smelling; fever) excessive bleeding or drainage from the tube site severe abdominal pain lasting ...

Fracture assessment of HSST Plate 14 shallow-flaw cruciform bend specimens tested under biaxial loading conditions

Energy Technology Data Exchange (ETDEWEB)

Bass, B.R.; McAfee, W.J.; Williams, P.T.; Pennell, W.E.

1998-06-01

A technology to determine shallow-flaw fracture toughness of reactor pressure vessel (RPV) steels is being developed for application to the safety assessment of RPVs containing postulated shallow surface flaws. Matrices of cruciform beam tests were developed to investigate and quantify the effects of temperature, biaxial loading, and specimen size on fracture initiation toughness of two-dimensional (constant depth), shallow, surface flaws. The cruciform beam specimens were developed at Oak Ridge National Laboratory (ORNL) to introduce a far-field, out-of-plane biaxial stress component in the test section that approximates the nonlinear stresses resulting from pressurized-thermal-shock or pressure-temperature loading of an RPV. Tests were conducted under biaxial load ratios ranging from uniaxial to equibiaxial. These tests demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower transition temperature region for an RPV material. The cruciform fracture toughness data were used to evaluate fracture methodologies for predicting the observed effects of biaxial loading on shallow-flaw fracture toughness. Initial emphasis was placed on assessment of stress-based methodologies, namely, the J-Q formulation, the Dodds-Anderson toughness scaling model, and the Weibull approach. Applications of these methodologies based on the hydrostatic stress fracture criterion indicated an effect of loading-biaxiality on fracture toughness; the conventional maximum principal stress criterion indicated no effect. A three-parameter Weibull model based on the hydrostatic stress criterion is shown to correlate the experimentally observed biaxial effect on cleavage fracture toughness by providing a scaling mechanism between uniaxial and biaxial loading states.

A general approach to flaw simulation in castings by superimposing projections of 3D models onto real X-ray images

International Nuclear Information System (INIS)

Hahn, D.; Mery, D.

2003-01-01

In order to evaluate the sensitivity of defect inspection systems, it is convenient to examine simulated data. This gives the possibility to tune the parameters of the inspection method and to test the performance of the system in critical cases. In this paper, a practical method for the simulation of defects in radioscopic images of aluminium castings is presented. The approach simulates only the flaws and not the whole radioscopic image of the object under test. A 3D mesh is used to model a flaw with complex geometry, which is projected and superimposed onto real radioscopic images of a homogeneous object according to the exponential attenuation law for X- rays. The new grey value of a pixel, where the 3D flaw is projected, depends only on four parameters: (a) the grey value of the original X-ray image without flaw; (b) the linear absorption coefficient of the examined material; (c) the maximal thickness observable in the radioscopic image; and (d) the length of the intersection of the 3D flaw with the modelled X-ray beam, that is projected into the pixel. A simulation of a complex flaw modelled as a 3D mesh can be performed in any position of the castings by using the algorithm described in this paper. This allows the evaluation of the performance of defect inspection systems in cases where the detection is known to be difficult. In this paper, we show experimental results on real X-ray images of aluminium wheels, in which 3D flaws like blowholes, cracks and inclusions are simulated

Preclinical animal anxiety research - flaws and prejudices.

Science.gov (United States)

Ennaceur, Abdelkader; Chazot, Paul L

2016-04-01

The current tests of anxiety in mice and rats used in preclinical research include the elevated plus-maze (EPM) or zero-maze (EZM), the light/dark box (LDB), and the open-field (OF). They are currently very popular, and despite their poor achievements, they continue to exert considerable constraints on the development of novel approaches. Hence, a novel anxiety test needs to be compared with these traditional tests, and assessed against various factors that were identified as a source of their inconsistent and contradictory results. These constraints are very costly, and they are in most cases useless as they originate from flawed methodologies. In the present report, we argue that the EPM or EZM, LDB, and OF do not provide unequivocal measures of anxiety; that there is no evidence of motivation conflict involved in these tests. They can be considered at best, tests of natural preference for unlit and/or enclosed spaces. We also argued that pharmacological validation of a behavioral test is an inappropriate approach; it stems from the confusion of animal models of human behavior with animal models of pathophysiology. A behavioral test is developed to detect not to produce symptoms, and a drug is used to validate an identified physiological target. In order to overcome the major methodological flaws in animal anxiety studies, we proposed an open space anxiety test, a 3D maze, which is described here with highlights of its various advantages over to the traditional tests.

Application of a finite element method to leak before break (LBB) of a heat exchanger

International Nuclear Information System (INIS)

Lee, Choon-Yeol; Kwon, Jae-Do; Lee, Yong-Sun

2003-01-01

The leak before break (LBB) concept is difficult to apply to a structure with a thin tube that is immersed in a water environment. A heat exchanger in a nuclear power plant is such a structure. The present paper addresses an application of the LBB concept to a heat exchanger in a nuclear power plant. The minimum leaked coolant amount containing the radioactive material which can activate the radiation detector device installed near the heat exchanger is assumed. The postulated initial flaw size that cannot grow to the critical flaw size within the time period to activate the radiation detector is justified. In this case, the radiation detector can activate the warning signal caused by coolant leakage from initially postulated flaws of the heat exchanger. The nuclear plant can safely shutdown when this occurs. Since the postulated initial flaw size can not grow to the critical flaw size, the structural integrity of the heat exchanger is not impeded. Particularly the informational scenario presented in this paper discusses an actual nuclear plant. (author)

Twenty years of fracture mechanics and flaw evaluation applications in the ASME Nuclear Code

International Nuclear Information System (INIS)

Riccardella, P.C.

1991-01-01

The paper presents a retrospective on the development and applications of fracture mechanics-based toughness requirements and flaw evaluation methodology in Sections III and XI of the ASME Code. Section III developments range from the rules and requirements for thick section Class 1 pressure vessels to thinner section components in other Classes. Section XI applications include flaw acceptance standards and evaluation methodology for various components ranging from pressure vessels to thins section piping of carbon and austenitic steels. The experience gained in operating plant applications of these rules and procedures are also discussed

Statistical flaws in design and analysis of fertility treatment studies on cryopreservation raise doubts on the conclusions

Science.gov (United States)

van Gelder, P.H.A.J.M.; Nijs, M.

2011-01-01

Decisions about pharmacotherapy are being taken by medical doctors and authorities based on comparative studies on the use of medications. In studies on fertility treatments in particular, the methodological quality is of utmost importance in the application of evidence-based medicine and systematic reviews. Nevertheless, flaws and omissions appear quite regularly in these types of studies. Current study aims to present an overview of some of the typical statistical flaws, illustrated by a number of example studies which have been published in peer reviewed journals. Based on an investigation of eleven studies at random selected on fertility treatments with cryopreservation, it appeared that the methodological quality of these studies often did not fulfil the required statistical criteria. The following statistical flaws were identified: flaws in study design, patient selection, and units of analysis or in the definition of the primary endpoints. Other errors could be found in p-value and power calculations or in critical p-value definitions. Proper interpretation of the results and/or use of these study results in a meta analysis should therefore be conducted with care. PMID:24753877

Statistical flaws in design and analysis of fertility treatment -studies on cryopreservation raise doubts on the conclusions.

Science.gov (United States)

van Gelder, P H A J M; Nijs, M

2011-01-01

Decisions about pharmacotherapy are being taken by medical doctors and authorities based on comparative studies on the use of medications. In studies on fertility treatments in particular, the methodological quality is of utmost -importance in the application of evidence-based medicine and systematic reviews. Nevertheless, flaws and omissions appear quite regularly in these types of studies. Current study aims to present an overview of some of the typical statistical flaws, illustrated by a number of example studies which have been published in peer reviewed journals. Based on an investigation of eleven studies at random selected on fertility treatments with cryopreservation, it appeared that the methodological quality of these studies often did not fulfil the -required statistical criteria. The following statistical flaws were identified: flaws in study design, patient selection, and units of analysis or in the definition of the primary endpoints. Other errors could be found in p-value and power calculations or in critical p-value definitions. Proper -interpretation of the results and/or use of these study results in a meta analysis should therefore be conducted with care.

Application of Fourier elastodynamics to direct and inverse problems for the scattering of elastic waves from flaws near surfaces

International Nuclear Information System (INIS)

Richardson, J.M.; Fertig, K.W. Jr.

1983-01-01

In order to inspect flaws which lie too close to the surface a Fourier elastodynamic formalism is proposed which enables one to decompose the elastodynamic system into separately charterizable parts by means of planes perpendicular to the z-axis. The process can be represented by a generalized transfer function relating the near-field scattered waves to the waves incident on a slab of material containing the flaw. The Fourier elastodynamics are applied to the characterization of the total scattering process involving a flaw at various distances from a plastic-water interface. An abbreviated discussion of Fourier elastodynamics is presented, and the results specialized to the case of spherical voids and inclusions bear an interface. Finally, the computational results for several ranges of temporal frequency and for a sequence of values of the distance from the flaw center to the interface are discussed

YouTube and Facebook

DEFF Research Database (Denmark)

Robertson, Scott P.; Vatrapu, Ravi; Medina, Richard

This paper examines the links to YouTube from the Facebook “walls” of Barack Obama, Hillary Clinton, and John McCain over two years prior to the 2008 U.S. Presidential election. User-generated linkage patterns show how participants in these politically-related social networking dialogues used...... online video to make their points. We show a strong integration of the Web 2.0 and new media technologies of social networking and online video. We argue that political discussion in social networking environments can no longer be viewed as primarily textual, and that neither Facebook nor YouTube can...

Advanced evacuated tube collectors

Science.gov (United States)

Schertz, W. W.; Hull, J. R.; Winston, R.; Ogallagher, J.

1985-04-01

The essence of the design concept for these new collectors is the integration of moderate levels of nonimaging concentration inside the evacuated tube itself. This permanently protects the reflection surfaces and allows the use of highly reflecting front surface mirrors with reflectances greater than 95%. Previous fabrication and long term testing of a proof-of-concept prototype has established the technical success of the concept. Present work is directed toward the development of a manufacturable unit that will be suitable for the widest possible range of applications. Design alternatives include scaling up the original prototype's tube diameter from 5 cm to 10 cm, using an internal shaped metal concentrating reflector, using a variety of profile shapes to minimize so-called gap losses and accommodate both single ended and double-ended flow geometries, and allowing the use of heat pipes for the absorber tube.

Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

International Nuclear Information System (INIS)

Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

1982-01-01

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

Rotating flux-focusing eddy current probe for flaw detection

Science.gov (United States)

Wincheski, Russell A. (Inventor); Fulton, James P. (Inventor); Nath, Shridhar C. (Inventor); Simpson, John W. (Inventor); Namkung, Min (Inventor)

1997-01-01

A flux-focusing electromagnetic sensor which uses a ferromagnetic flux-focusing lens simplifies inspections and increases detectability of fatigue cracks about circular fasteners and other circular inhomogeneities in high conductivity material. The unique feature of the device is the ferrous shield isolating a high-turn pick-up coil from an excitation coil, The use of the magnetic shield is shown to produce a null voltage output across the receiving coil in the presence of an unflawed sample. A redistribution of the current flow in the sample caused by the presence of flaws, however, eliminates the shielding condition and a large output voltage is produced, yielding a clear unambiguous flaw signal. By rotating the probe in a path around a circular fastener such as a rivet while maintaining a constant distance between the probe and the center of a rivet, the signal due to current flow about the rivet can be held constant. Any further changes in the current distribution, such as due to a fatigue crack at the rivet joint, can be detected as an increase in the output voltage above that due to the flow about the rivet head.

Amplitude-independent flaw length determination using differential eddy current

Science.gov (United States)

Shell, E.

2013-01-01

Military engine component manufacturers typically specify the eddy current (EC) inspection requirements as a crack length or depth with the assumption that the cracks in both the test specimens and inspected component are of a similar fixed aspect ratio. However, differential EC response amplitude is dependent on the area of the crack face, not the length or depth. Additionally, due to complex stresses, in-service cracks do not always grow in the assumed manner. It would be advantageous to use more of the information contained in the EC data to better determine the full profile of cracks independent of the fixed aspect ratio amplitude response curve. A specimen with narrow width notches is used to mimic cracks of varying aspect ratios in a controllable manner. The specimen notches have aspect ratios that vary from 1:1 to 10:1. Analysis routines have been developed using the shape of the EC response signals that can determine the length of a surface flaw of common orientations without use of the amplitude of the signal or any supporting traditional probability of detection basis. Combined with the relationship between signal amplitude and area, the depth of the flaw can also be calculated.

Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

International Nuclear Information System (INIS)

Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

1965-11-01

The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

Development of a probe for inner profile measurement and flaw detection

Science.gov (United States)

Yoshizawa, Toru; Wakayama, Toshitaka; Kamakura, Yoshihisa

2011-08-01

It is one of the important necessities to precisely measure the inner diameter and/or the inner profile of pipes, tubes and other objects similar in shape. Especially in mechanical engineering field, there are many requests from automobile industry because the inner surface of engine blocks and other die casts are strongly required to be inspected and measured by non-contact methods (not by the naked eyes inspection using a borescope). If the inner diameter is large enough like water pipes or drain pipes, complicated and large equipment may be applicable. However, small pipes with a diameter ranging from 10mm to 100mm are difficult to be inspected by such a large instrument as is used for sewers inspection. And we have proposed an instrument which has no moving elements such as a rotating mirror or a prism for scanning a beam. Our measurement method is based on optical sectioning using triangulation. This optically sectioned profile of an inner wall of pipe-like objects is analyzed to produce numerical data of inner diameter or profile. Here, we report recent development of the principle and applications of the optical instrument with a simple and compact configuration. In addition to profile measurement, we found flaws and defects on the inner wall were also detected by using the similar principle. Up to now, we have developed probes with the diameter of 8mm to 25mm for small size objects and another probe (80 mm in diameter) for such a larger container with the dimensional size of 600mm.

Robust precision alignment algorithm for micro tube laser forming

NARCIS (Netherlands)

Folkersma, Ger; Brouwer, Dannis Michel; Römer, Gerardus Richardus, Bernardus, Engelina; Herder, Justus Laurens

2016-01-01

Tube laser forming on a small diameter tube can be used as a high precision actuator to permanently align small (optical)components. Applications, such as the alignment of optical fibers to photonic integrated circuits, often require sub-micron alignment accuracy. Although the process causes

Gun bore flaw image matching based on improved SIFT descriptor

Science.gov (United States)

Zeng, Luan; Xiong, Wei; Zhai, You

2013-01-01

In order to increase the operation speed and matching ability of SIFT algorithm, the SIFT descriptor and matching strategy are improved. First, a method of constructing feature descriptor based on sector area is proposed. By computing the gradients histogram of location bins which are parted into 6 sector areas, a descriptor with 48 dimensions is constituted. It can reduce the dimension of feature vector and decrease the complexity of structuring descriptor. Second, it introduce a strategy that partitions the circular region into 6 identical sector areas starting from the dominate orientation. Consequently, the computational complexity is reduced due to cancellation of rotation operation for the area. The experimental results indicate that comparing with the OpenCV SIFT arithmetic, the average matching speed of the new method increase by about 55.86%. The matching veracity can be increased even under some variation of view point, illumination, rotation, scale and out of focus. The new method got satisfied results in gun bore flaw image matching. Keywords: Metrology, Flaw image matching, Gun bore, Feature descriptor

Corrosion and Rupture of Steam Generator Tubings in PWRs

Energy Technology Data Exchange (ETDEWEB)

Hwang, Seong Sik; Kim, Hong Pyo

2007-08-15

This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned.

Corrosion and Rupture of Steam Generator Tubings in PWRs

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Hong Pyo

2007-08-01

This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

Flaw preparations for HSST program vessel fracture mechanics testing: mechanical-cyclic pumping and electron-beam weld-hydrogen-charge cracking schemes