In-service Inspection Research Articles

Metallic replica technique for residual life assessment for components serviced in creep regime

Summary For 9–12% Cr creep-resistant steels, a new kind of replica, contemporaneously morphological and extractive, is nowadays in use. This method is non-destructive and ‘in situ’ applicable via a customised portable equipment and allows the study of the steel microstructure evolution during in-service inspections. The technique produces a neat imprint of the steel microstructure, like a morphological replica, but contemporaneously extracts particles down to 15–20 nm diameters. The use of a good electrical conductor extraneous to the steel composition as a replication basis allows the investigation of the martensitic (sub-) microstructure together with the particle distribution, nanometric dimensional and chemical analyses using a tungsten electrode scanning electron microscope at magnifications up to 500000×. For component residual life assessment, besides a measuring technique and a damage witnessing criterion, a reference is needed against which spent lifetime through which or actual damage can be estimated. .

Finite Element Analysis of Internally Circumferential Surface Cracked Pipe Stainless Steel under Bending and Torsion Loadings

Early Light Water Reactor Plants (LWR plants) in Japan are nearly 30 years old and utilities have been conducting technical evaluations of facility aging. Accurate prediction of fracture behavior for ductile materials is one of the most important issues to guarantee the integrity of the structure such as pipes in plants. If a crack is detected during the in-service inspections, the structural rational maintenance of the cracked pipe has been promoted based on the defect evaluation methods specified ASME Boiler and Pressure Vessel Code and JSME Rules on Fitness-for-Service for Nuclear Power Plants. Recently, experiments and simulations have been performed on pipes subjected to bending and torsion loadings. However, the failure criterion is still under investigation, and it is significant to propose a method to evaluate it appropriately. In this study, a finite element model of a large circumferentially cracked stainless steel pipe was generated and elastic-plastic analysis of stationary crack and crack extension were performed to validity of the results.

Probabilistic assessment of fatigue crack propagation with accounting for effects of crack retardation due to overloads

The paper presents a probabilistic approach to fatigue strength assessment based on the Wheeler model that provides the opportunity to account for crack retardation due to overload in a probabilistic formulation. The probabilistic assessment is performed by Monte Carlo simulations based on random sampling. It consists of running deterministic assessments a sufficiently large number of times with different combinations of randomly sampled input data in order to build a statistical picture of the output quantities. A computer code in Matlab was developed to obtain the prior assessment of the probability of fatigue failure of a pipeline structural component. The Bayesian inference is used to update the prior assessment failure probability assessments as soon data from in-service technical inspections become available

In Situ Creep Rupture Life Prediction Methods for 316FR Austenitic Stainless-Steel Weldment

The present paper proposes new creep life prediction methods that can be used in situ as an in-service inspection tool for the condition-based monitoring of high-temperature structure components of 316FR stainless steel that has been developed in Japan for fast breeder reactors. The methods were based on two types of the modified theta methods and the combination of the modified theta and modified omega methods and were applied to creep deformation data during creep tests made at different applied stress levels at 823K and 923 K in air. Creep lives were predicted by the proposed three methods at arbitrary time in the continuous measurement of deformation during the creep tests. The results show that the combination of the modified and modified omega methods gives the most accurate prediction of creep rupture time at the two levels of test temperature for different applied stress levels.

Acoustic Emission Analysis during Fatigue Damage Process on Type 2 Accumulator Materials for Hydrogen Station

A research project is underway to apply the Acoustic Emission (AE) method to establish a non-destructive, non-opening, in-service inspection of high-pressure hydrogen Type 2 accumulators for hydrogen stations. The objective is to prevent hydrogen leakage due to fatigue crack initiation, propagation, and penetration in the steel liner of the accumulator caused by cyclic pressure loading. In this study, the relationship between fatigue crack initiation/growth and AE wave generation behavior is evaluated using a plate-type specimen that simulates the low-alloy steel liner of a pressure accumulator. As a result, AE generation is considered to capture the transition of a crack from Stage I to Stage II, and the usefulness of applying the AE method for in-service inspection of high-pressure hydrogen Type 2 accumulators is demonstrated.

Design and manufacture of graphite components for 21st century small modular reactors

Successful deployment of graphite in 21st century nuclear reactors depends on the rigor and quality of communication between the component designer and manufacturer. For construction in the U.S., a thorough understanding of meeting the regulatory requirements, ASME International graphite component design and construction code requirements, and ASTM International material specifications enables efficiency and cost reduction during construction, reactor operation, and decommissioning. These codes and standards (C&S) are interlinked in several technical, component design, and reactor operation functional requirements. The focus in this technical note (TN) is to establish a framework for engaging the graphite component designer (GCD) and the graphite component manufacturer (GCM). This TN discusses five significant areas for active interactions between the designer and manufacturer, which begin at the preconceptual design stage and continue through material and fabrication method selection, component manufacturing, installation, reactor operation, decommissioning, and disposal. Performance success and overall cost reduction could likely be enhanced by adapting proven ceramic component probabilistic brittle design principles and incorporating features to minimize worker dose exposure during in-service inspection, component replacement, and decommissioning.

Corrosion damage and in-service inspection of retractable sootblower lances in recovery boilers

Several reports of accidents involving serious mechanical failures of sootblower lances in chemical recovery boilers are known in the pulp and paper industry. These accidents mainly consisted of detachment and ejection of the lance tip, or even of the entire lance, to the inside of the furnace, towards the opposite wall. At least one of these cases known to the author resulted in a smelt-water explosion in the boiler. In other events, appreciable damage or near-miss conditions have already been experienced. The risk of catastrophic consequences of the eventual detachment of the lance tip or the complete lance of a recovery boiler soot-blower has caught the attention of manufacturers, who have adjusted their quality procedures, but this risk also needs to be carefully considered by the technical staff at pulp mills and in industry committees. This paper briefly describes the failure mechanisms that prevailed in past accidents, while recommending inspection and quality control policies to be applied in order to prevent further occurrences of these dangerous and costly component failures. Digital radiography, in conjunction with other well known inspection techniques, appears to be an effective means to ensure the integrity of sootblower lances in chemical recovery boilers used in the pulp and paper industry.

Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP Regarding Stress Corrosion Cracking in Nickel Based Alloy Weld Joint of Piping System in Boiling Water Reactor

Abstract In the past few decades, cracks because of stress corrosion cracking (SCC) have been detected in the dissimilar weld joints using nickel-based alloy in piping system of boiling water reactors (BWRs). Thus, the structural integrity assessment for such weld joints has become important. Nowadays, probabilistic fracture mechanics (PFM) analysis is recognized as a rational method for structural integrity assessment because it can consider inherent uncertainties of various influencing factors as probability distributions and quantitatively evaluate the failure probability of a cracked component. The Japan Atomic Energy Agency has developed a PFM analysis code PASCAL-SP for a weld joint of piping system in nuclear power plant. This study improves the analysis functions of PASCAL-SP for weld joint using nickel-based alloy in BWR susceptible to SCC. As an analysis example of the improved version of PASCAL-SP, the failure probability of a weld joint is quantitatively evaluated. Furthermore, sensitivity analyses are conducted concerning the effect of leak detection and in-service inspection. From the analysis results, it is concluded that the improved version of PASCAL-SP is useful for structural integrity assessment.

Evolution of Standardized Specifications on Materials, Manufacturing and In-Service Inspection of Nuclear Reactor Vessels

The cataloguing and revision of reactor pressure vessels (RPV) manufacturing and in-service inspection codes and their standardized material specifications—as a technical heritage—are essential for understanding the historical evolution of criteria and for enabling the comparison of the various national regulations, integrating the most relevant results from the scientific research. The analysis of the development of documents including standardized requirements and the comparison of regulations is crucial to be able to implement learned lessons and comprehend the progress of increasingly stringent safety criteria, contributing to sustainable nuclear power generation in the future. A novel methodology is presented in this work where a thorough review of the regulations and technical codes for the manufacture and in-service inspection of RPVs, considering the implementation of scientific advances, is performed. In addition, an analysis focused on the differences between irradiation embrittlement prediction models and acceptance criteria for detected defects (both during manufacturing and in-service inspection) described by the different technical codes as required by different national regulations such as American, German, French or Russian is performed. The most stringent materials requirements for RPV manufacturing are provided by the American and German codes. The French code is the most stringent with respect to the reference defect size using as a criterion in the in-service inspection.

Study on Fast Fracture Evaluation Method for Reactor Pressure Vessel with Excessive Carbon Content

The initial non-ductile transition temperature is an important input parameter in fast fracture evaluation according to the RCC-M code. The initial non-ductile transition temperature (RTNDT) will be affected by the carbon content, but there is no quantitative relationship between them. When the carbon content of the reactor pressure vessel exceeds the standard, it is necessary to complete the fast fracture evaluation without quantitative relationship to ensure the integrality of RPV. In this paper, the evaluation method of the fast fracture of RPV with excessive carbon content is studied. Taking the RPV core barrel with excessive carbon content as an example, the defect repair is considered, and the fast fracture analysis and evaluation are carried out by calculating the maximum initial non-ductile transition temperature. This method can support the operation and the in-service inspection of the pressure vessels with excessive carbon content.

Analysis and evaluation of the irradiated dimensional and volume change behavior of nuclear graphites

Literature on the changes in the specimen dimensions and volume after high energy neutron irradiation of more than a dozen graphites were examined. Some of these data have been previously used to design graphite components in gas cooled reactors. The dimensional change behavior of a graphite is rather unique and depends on both the irradiation temperature and accumulated irradiation dose. The characteristics of the dimensional and volume change behavior of several graphites were analyzed and evaluated from engineering design consideration of graphite components. Specifically, six parameters were derived from the dimensional and volume change behavior. These parameters can serve as “figures-of-merit” (FOM) to enable engineering judgement on the irradiation-induced dimensional and volume change behavior of nuclear graphites, and are referred to as “derived figure-of-merit (DFOM).” The DFOMs also provide engineering bases to examine the opportunity to select several competing graphites for fulfilling the design and operation of components in reactor service. The utility of these DFOMs can be realized potentially in designing inservice inspection and component condition monitoring (CCM) protocols that are effective and efficient both technically and economically. The DFOMs, in combination, also can guide the reactor operator to arrive at optimum component replacement schedule for graphite components, which ensure maintaining graphite components functional requirements and related reactor safety issues.

Multilevel modeling of diametral creep in pressure tubes of Korean CANDU units

In this work, we applied a multilevel modeling technique to estimate the diametral creep in the pressure tubes of Korean Canada Deuterium Uranium (CANDU) units. Data accumulated from in-service inspections were used to develop the model. To confirm the strength of the multilevel models, a 2-level multilevel model considering the relationship between channels for a CANDU unit was compared with existing linear models. The multilevel model exhibited a very robust prediction accuracy compared to the linear models with different data pooling methods. A 3-level multilevel model, which considered individual bundles, channels, and units, was also implemented. The influence of the channel installation direction was incorporated into the three-stage multilevel model. For channels that were previously measured, the developed 3-level multilevel model exhibited a very good predictive power, and the prediction interval was very narrow. However, for channels that had never been measured before, the prediction interval widened considerably. This model can be sufficiently improved by the accumulation of more data and can be applied to other CANDU units.

Concept design of the upper port remote handling equipment for ITER neutral beam cell maintenance

Among the eighteen upper ports of the ITER Tokamak, the upper ports #4, #5, #6 and #7 are located in the Neutral Beam (NB) cell. A roughing filter and an upper port plug are located inside each upper port. Vacuum Vessel (VV) in-service inspection and neutron flux monitor ports are located on the balcony area near the third Heating Neutral Beam (HNB). This paper describes the concept design of the Upper Port Remote Handling Equipment (UPRHE) that provides means to maintain the components at the upper ports and the balcony area within the NB cell. The UPRHE can move along the toroidal direction to access the upper ports and can deploy various remote handling (RH) tools into the upper ports to access the components that require maintenance. The UPRHE consists of permanent installations in the NB cell and removable equipment that is transferred from the Hot Cell Facility to the NB cell only when carrying out the maintenance activities. The permanent equipment is the cabling device mounted on wall. The removable equipments are the upper port trolley, manipulator transporter, end-effectors, manipulators, and other RH tools. The concept design and the RH compatibility assessment demonstrate that the proposed UPRHE can perform the identified tasks and is ready to move on to the next design phase.

Assessment of Icephobic Coatings—How Can We Monitor Performance Durability?

Significant progress in the field of icephobic coatings has raised a demand for evaluation criteria to assess and monitor the related icephobic effects and their durability. The initial coating performance in preventing ice formation and reducing ice adhesion needs to be proven over a given period by withstanding technically relevant stressors. In this study, silanized polyurethane (PUR) coatings are assessed in conjunction with a standardized accelerated ultraviolet (UV)-ageing procedure in order to identify potential monitoring tools that are also applicable during in-service inspections. Wettability and roughness parameters are recorded after pre-defined ageing intervals, compared with the ice adhesion strength, and tested using a modified centrifuge. Correlation assessments indicate that the chosen parameters cannot generally be used for the monitoring of icephobic effects for the selected material class. It is more likely that specific coating parameter sets need to be defined for in-service monitoring, as an important step towards the integration of icephobic coatings into technical applications.

A fracture mechanics framework for optimising design and inspection of offshore wind turbine support structures against fatigue failure

Abstract. Offshore wind turbine (OWT) support structures need to be designed against fatigue failure under cyclic aerodynamic and wave loading. The fatigue failure can be accelerated in a corrosive sea environment. Traditionally, a stress–life approach called the S–N (stress–number of cycles) curve method has been used for the design of structures against fatigue failure. There are a number of limitations in the S–N approach related to welded structures which can be addressed by the fracture mechanics approach. In this paper the limitations of the S–N approach related to OWT support structure are addressed and a fatigue design framework based on fracture mechanics is developed. The application of the framework to a monopile OWT support structure is demonstrated and optimisation of in-service inspection of the structure is studied. It was found that both the design of the weld joint and non-destructive testing (NDT) techniques can be optimised to reduce in-service inspection frequency. Furthermore, probabilistic fracture mechanics as a form of risk-based design is outlined and its application to the monopile support structure is studied. The probabilistic model showed a better capability to account for NDT reliability over a range of possible crack sizes as well as to provide a risk associated with the chosen inspection time which can be used in inspection cost–benefit analysis. There are a number of areas for future research, including a better estimate of fatigue stress with a time-history analysis, the application of the framework to other types of support structures such as jackets and tripods, and integration of risk-based optimisation with a cost–benefit analysis.

In-service Piping Inspection Work-aid Tool for Oil & Gas Industries

Piping systems are important in the oil and gas plant’s operation, but continuous damage is harming the piping system due to extensive usage resulting in the increase of accident cases. API 570 Piping Inspection Code: In-service Inspection, Rating, Repair, and Alteration of Piping Systems has proposed an in-service piping inspection practice to ensure the in-service piping is functionable by identifying the remaining life of the piping system. However, the in-service piping inspection process is numerous and complicated where certain steps may be skipped, and data collected may lost along the process. The in-service piping inspection framework followed by a work-aid tool is developed in this study based on API 570 to guide the user the piping inspection process along with providing a depository database for document storage. Validation test is conducted by collecting feedback from professional piping engineer using System Usability Scale followed by conducting a case study using secondary data and sample attachment to test the functionality of the work-aid tool. The work-aid tool can guide the piping inspection process and provide a systematic documentation method for corresponding inspection documents. Case study in the industry is recommended to test the usability of the tool in the industry.

소듐냉각고속로 원자로용기 가동중검사를 위한 고온 전자기음향 트랜스듀서 성능 개선

200°C 이상의 고온과 협소한 검사 공간 및 액체 접촉 매질의 사용이 제한되는 가혹한 조건에서 수행되어야 하는 소듐냉각고속로 원자로용기 가동중검사에 대한 적용성 향상을 위해 기 개발된 고온 전자기음향 트랜스듀서의 성능을 개선하였다. 기존 PPM 방식 영구자석 배열 대비 높은 자속밀도를 얻을 수 있는 Halbach 방식 영구자석 배열을 도입하였고, 자기장 해석을 통해 기존 PPM 방식 대비 1.4배 이상 큰 자속밀도를 얻을 수 있는 최적의 자석 크기 비율을 선정하였다. 선정된 최적 자석 크기 비율을 기반으로 개선된 고온 전자기음향 트랜스듀서를 제작하고 성능검증시험을 수행하여 기존 대비 33% 이상 증가된 리프트 오프에서도 원자로용기 두께 대비 5% 깊이의 내면 결함 탐지가 가능함을 확인하였다.

Supervisory practices and challenges faced by senior high school principals in Greater Monrovia, Liberia: implications for quality education

Education is an inseparable development element that influences a nation's socio-economic growth and its citizenry for a quality life. This study aimed to explore the supervisory practices by senior high school principals in Greater Monrovia, Liberia and the challenges faced during supervision. The study employed qualitative design and included 30 participants. Findings indicated that senior high school principals used supervisory practices such as inspection of teachers' lesson notes, punctuality, organized in-service training and inspection of classroom managerial skills of teachers. Notwithstanding these supervisory practices by school principals, the study found that supervision was impeded by lack of resources for teaching and learning, limited teaching staff, inexperience teachers, and improper attitudes by teachers and students.

Radiological aspects during cutting and removal of L-08 coolant channel from the core of 540 MWe TAPS-4 nuclear reactor

The pressurized heavy-water reactors (PHWR) consist of a low-pressure horizontal reactor vessel, calandria, containing heavy water as moderator. The calandria is pierced by a large number of coolant tubes (also called pressure tubes [PTs], 392 in 540 MWe PHWR), which contains fuel bundles, and through which pressurized heavy-water coolant circulates. During biennial shut down in 2017, in-service inspection of coolant channels in Tarapur atomic power station-4 (TAPS-4) had been carried out and after the review of results of inspection, it was recommended by various groups of experts that channel L-08 should be removed for postirradiation examination at Bhabha Atomic Research Center (BARC). Before the channel removal job, one special shielding flask was designed to shift the removed channel to BARC. The integrity test of special shielding flask was carried out by safely placing cobalt-60 (source strength 851000 MBq) capsule source inside the shielding flask with the help of cranking unit mechanism followed by dose rate mapping on the outer surface of the flask. To establish the hydrogen pickup rates in L-08 PT, sliver samples were collected and separately sent to BARC. Four metallic sliver samples were obtained at four different distances from north E-face. The activity content present in each sliver sample was also estimated. The maximum activity estimated was 2313.24 MBq. Subsequently, L-08 coolant channel was cut from both sides using a chipless tool. Jobs involving heavy water (D2O) collection work were carried out with a ventilated plastic suit. Derived air concentration (DAC) of tritium at the work location was maintained below 1DAC during the entire activity. Particulate DAC was found below the detectable limit. Floor contamination checks and floor decontamination were conducted at regular intervals to avoid buildup of contamination. As a result of such high-quality radiological safety measures, only 25 workers, out of 270 radiation workers, have received a cumulative dose of more than 3 mSv in direct reading dosimeter with a maximum individual dose of 8.45 mSv and maximum individual uptake of 0.39 GBq/m3. Job was completed with a total collective dose of 324.35P-mSv which is 14.5% lower than estimated. This article highlights some of the critical tasks involved in the cutting and the removal of irradiated coolant channel from the core of 540 Mwe TAPS-4 reactor which is a first of kind activity in nature.

Difference analysis of in-service hydrostatic test for second and third generation PWR

Hydraulic test of nuclear power plant is an important part of in-service inspection of nuclear power plant, and it is one of the important technical means to test the pressure bearing capacity of equipment. In this paper, the technical requirements of the second generation PWR based on safety classification and the third generation reactor based on radioactive release classification are discussed from the aspects of equipment classification, primary circuit hydrostatic test, steam generator secondary side hydraulic test, hydrostatic test of vessel and pipeline, as well as the difficulty of implementation in hydrostatic test.