bluuebunny/crossref_metadata_2025_split huggingface dataset:
Numerical Study on Fracture Behavior of Complex Cracked Pipes
Authors: Ryu Ho-Wan , Han Jae-Jun , Kim Yun-Jae 
Published: November 2015

This study provides the application of damage model to complex cracked pipes which can be found especially in weld overlay region. From the perspective of structural integrity, enough basic and large-scale tests are required to accurately evaluate the components containing a crack-like defect. In this case, damage model using finite element (FE) method can be effectively used for the assessment of full-scale cracked pipes with minimum basic experiments data. The proposed method in this research is based on the stress-modified fracture strain damage model with stress reduction technique.
In this paper, Battelle full-scale complex cracked pipe tests are simulated by the proposed damage model with reasonable procedure. FE simulation is conducted for basic experiments to determine failure criteria with calibrations. Then, crack initiation and maximum loads are predicted to characterize the fracture behavior of full-scale complex cracked pipes. Damage model is applied to both of carbon and stainless steel materials and verification with comparing to test data is conducted.
Subsequent License Renewal in the U.S.
Authors: Young Garry G. , Remer Jason 
Published: November 2015

As of January 2015, the NRC has renewed the operating licenses for 75 nuclear units, allowing for up to 60 years of safe operation. In addition, the NRC has license renewal applications under review for 19 units and 5 additional units have announced plans to submit applications over the next few years. This brings the total of renewed licenses and plans for renewal to almost 100% of the operating nuclear units in the U.S. At the end of 2014 there were 38 nuclear plants that had operated for more than 40 years and are eligible to seek a second license renewal (or 38% of the operating units). [1]
Although some nuclear plant owners have shutdown operation recently and others have announced plans to shutdown before reaching 60 years, the majority are keeping the option open for long term operation beyond 60 years. NRC and the U.S. nuclear industry have made significant progress in preparing the way for second license renewal applications. For example, in October 2014, 7 nuclear units were identified as the leading candidates for subsequent license renewal in a New York Times news article. This paper presents the status of the issues being addressed for possible applications for second license renewals for up to 80 years of operation.
Formulae for Plastic ETA Factors Accounting for Strain Hardening in SEB and SET Specimens
Authors: Gilles Philippe , Zhang Gongchen , Wang Xiaobei 
Published: November 2015

Elasto-plastic fracture mechanics is commonly used for defect assessments in Nuclear Industry. Most of the approaches are based on the crack driving force J which is compared to a fracture resistance parameter. This fracture resistance parameter cannot be measured directly, but requires either J engineering formulae or Finite Element computations. The η(Eta) factor approach [1] is the most popular method allowing deriving fracture resistance from test records. One of the issue is the dependency of this factor to strain-hardening, namely for the shallow cracks.
This paper analyzes the effect of strain hardening on Eta factor solutions for two types of loading. Detailed Finite Element results are provided for Single Edge Cracked specimens under bending or tension. The results are compared to other formulae from the literature and new formulae are proposed.
A Probability-Based Leak-Before-Break Application for Small Modular Reactors
Authors: Xu Heqin , Liszkai Tamas 
Published: November 2015

The leak-before-break (LBB) applicability is stated in General Design Criterion 4 (GDC-4) of Title 10 of the Code of Federal Regulation Part 50 (10 CFR 50). GDC-4 requires that analyses reviewed and approved by the U.S. Nuclear Regulatory Commission (NRC) demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping, in order that dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis. Standard review plan 3.6.3 (SRP-3.6.3) further requires a simultaneous safety margin of two and ten on the flaw size and leak rate detectability, respectively, for deterministic analyses, believing that the very conservative and restrictive safety margins would lead to extremely low probability of fluid system piping rupture. The technology advancements of recent years make it possible to numerically quantify the probability of rupture with confidence.
In this study, we are going to utilize a simplified methodology based on the Univariate Dimension-Reduction (UDR) method to investigate the probability of piping rupture for postulated circumferential through-wall cracks in piping systems of small modular reactors (SMRs) inside the metal containment vessel, which typically range from 5” to 8” nominal diameter for feedwater and main steam lines. The unique design of the vacuumed metal CNV in an advanced SMR offers improved leak detection capability, which makes the LBB qualification possible even for piping with much smaller diameters than piping systems in the traditional large reactors, which typically range from 18” to 40” for LBB-qualified piping systems such as surge lines, cold legs, hot legs and main steam lines.
The conditional probability that a piping with postulated crack would fail by large break, when multiplied by the probability of having a circumferential through-wall crack during the life time of plant service, produces an overall probability of piping rupture. The major quantifiable uncertainties, such as the uncertainties associated with the material properties, and flow-path crack face morphology parameters will be modeled as correlated random variables in this paper. Efficient Dimension-Reduction methods will be applied to predict the conditional probability. As sample applications of the proposed method, the relationship between the conditional probabilities and the loading factor will be established for SMR piping systems.
Demonstration of Fatigue for LTO License of NPP Borssele
Authors: Hannink M. H. C. , Blom F. J. , Quist P. W. B. , de Jong A. E. , Besuijen W. 
Published: November 2015

Long Term Operation (LTO) of nuclear power plants (NPPs) requires an ageing management review and a revalidation of Time Limited Ageing Analyses (TLAAs) of structures and components important for nuclear safety. An important ageing effect to manage is fatigue. Generally, the basis for this is formed by the fatigue analyses of the safety relevant components. In this paper, the methodology for the revalidation of fatigue TLAAs is demonstrated for LTO of NPP Borssele in the Netherlands.
The LTO demonstration starts with a scoping survey to determine the components and locations having relevant fatigue loadings. The scope was defined by assessment against international practice and guidelines and engineering judgment. Next, a methodical review was performed of all existing fatigue TLAAs. This also includes the latest international developments regarding environmental effects. In order to reduce conservatism, a comparison was made between the number of cycles in the analyses and the number of cycles projected to the end of the intended LTO period. The projected number of cycles is based on transient counting. The loading conditions used in the analyses were assessed by means of temperature measurements by the fatigue monitoring system (FAMOS). As a result of the review, further fatigue assessment or assessment of environmental effects was necessary for certain locations. New analyses were performed using state-of-the-art calculation and assessment methods. The methodology is demonstrated by means of an example of the surge line. The model includes the piping, as well as the nozzles on the pressurizer and the main coolant line. The thermal loadings for the fatigue analysis are based on temperature measurements. Fatigue management of the NPP is ensured by means of the fatigue concept where load monitoring, transient counting and fatigue assessment are coupled through an integrated approach during the entire period of LTO.
Development of a Testing System for Piezoelectric Wafer Active Sensors Exposed to Extreme Temperatures
Authors: Roth William , Poddar Banibrata , Bao Jingjing 
Published: November 2015

Piezoelectric Wafer Active Sensors (PWAS) are a viable option for monitoring the structural integrity of pressure vessels and piping systems. They are inexpensive, small and unobtrusive sensors which can be permanently attached to structures for long term monitoring without interfering with operations, such as operating in areas with limited head space. PWAS are used to inspect the structure through several methods which include; pitch catch or pulse echo wave propagation, and electromechanical impedance spectroscopy. Since the PWAS could be exposed to a range of environmental and/or operating conditions while attached to the structure, the change in the properties and electromechanical characteristics of the sensor must be known at a given condition. Accordingly, there is a need for a testing system which can measure the PWAS properties while exposing the sensor to a wide range of temperatures.
The focus in this paper is on elevated temperatures, but the same methodology could be used for low temperature or environmental testing. The requirements which were imposed on the design include: providing an electrical connection from each electrode to the exterior of an industrial oven; allow the sensor to expand and contract both in plane and out of plane; withstand an extended duration at elevated temperatures; the equipment must not influence the measured quantities. The challenges include how to place the sensor in an oven and make electrical contact while allowing free motion, how to implement wiring and electrical connections at elevated temperatures, how to allow the thermal expansion of components and account for thermal mismatching, and how to maintain electrical isolation of the two electrodes. This paper discusses how these requirements were met and challenges overcome, as well as experimental validation of the system.
A Rational Basis for Deterministic Margins in CUF-Based Fatigue Evaluation With Uncertainty
Authors: Garud Y. S. , Steininger David A. 
Published: November 2015

One of the aging issues of concern in the light-water reactor power plants is the fatigue degradation due to cyclic loads for which the updated regulatory guidelines have factored in the water environment effects that were not in the original design basis analyses. These guidelines are also applicable to newer plant designs not yet in operation. In both these cases, notably, the design/analysis is deterministic with several factors conservatively used as an engineering judgment to address various sources of uncertainty in the cumulative usage factor (CUF) based fatigue evaluation. The use of deterministic approach to an inherently statistical issue with large and unspecified uncertainty of analysis does not provide a consistent or quantifiable measure of actual safety margin, nor does it lend itself well to assess the impact of various conservative assumptions, or their potential reductions, on the gain/loss in the margin. While a probabilistic methodology is expected to be pertinent and useful it is also recognized that the deterministic approach has the simplicity and familiarity to its credit. These aspects of significance to the long-term assessment and management of fatigue degradation are addressed in this paper with a new approach focused on its engineering application.
Presented in this paper is a rational basis that relates the results of a probabilistic assessment of CUF to the common notion of deterministic margins. The objective is to account for major sources of uncertainty in the evaluation of CUF and its allowable or performance criterion, with the goal of allowing for a more realistic assessment of fatigue damage and a consistent quantification of fatigue margins. The basis itself is developed using the stress–strength (or load–resistance) interference methodology that is probabilistic. The results of application of this methodology are expressed in the form of suitable margins that can be treated in the usual deterministic manner. The relation between application dependent acceptable risk level and the quantified fatigue margin is examined to further utilize the results of proposed approach in a simplified engineering manner. Limitations of the key assumptions in the development of this basis/approach, and their expected practical significance, are discussed with suggestions for further enhancements from the application point of view.
Computational Aspect of Nonlinear Fracture Mechanics Application
Authors: Orynyak Igor , Oryniak Andrii 
Published: November 2015

The development of powerful commercial computer programs made the concept of J-integral as computational parameter of fracture mechanics to be a very attractive one. It is equivalent to SIF in linear case, it converges in numerical calculation and the same results are obtained by different codes (programs). Besides, it is widely thought that elastic-plastic analysis gives bigger values than elastic SIF ones what is good from regulatory point of view. Such stand was reflected in the recommended by IAEA TECDOC 1627 (February 2010) devoted to pressurized thermal shock analysis of reactor pressure vessels, where the embedded crack in FEM mesh, elastic-plastic analysis with simultaneous determination of J-integral was stated as the best option of analysis.
But at that time all the most widely used software were not able to treat the residual stresses, the thermal stresses in case of two different materials. Such a contradiction between requirements and the possibilities made a lot of problems for honest contractors especially in countries where the regulator had no own experience in calculation and completely relied on the authority of international documents. This means that at that time the said recommendations were harmful. The main reason of such a situation was the absence of the carefully elaborated examples.
Now the capabilities and accuracy of such software are increasing. Nevertheless, some principal ambiguities and divergences of computations results in various J-integral contours around the crack tip still exist. They are exhibited when the large plastic zone emerges at the crack tip. Other problem is influence of the history of loading and the specification of the time of crack insertion in the mesh including the time of emergence of residual stress.
This paper is invitation for discussion of the accuracy and restriction of computational J-integral. With this aim the detailed analysis of some simplified 2D examples of calculation of elastic -plastic J-integral for surface crack with accounting for residual stress, thermal stress and inner pressure are performed and commented.
The attainment of consensus among the engineering society for treating the outcome results is the prerequisite for practical application of computational elastic plastic J-integral.
Non-Contact Gas Accumulation Detection in Water Loaded Pipes Using Guided Waves
Authors: Tian Zhenhua , Sun Xiaoyi , Ma Linlin , Yu Lingyu 
Published: November 2015

The U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter 2008-01 due to the safety concern associated with gas accumulation events in emergency core cooling pipes of nuclear power plants. Since the gas accumulation may critically damage pipes, pumps, and valves, and affect the safety operation of nuclear power plants, the gas accumulation needs to be detected as well as quantified.
This paper presents a quantitative gas accumulation detection method for water loaded pipes by using guided waves. To establish the detection method, we investigated the differences between guided waves in a free pipe and those in a water loaded pipe. The guided waves in both cases were measured by using a scanning laser Doppler vibrometer, and analyzed by using frequency wavenumber analysis. Analysis results show that guided wave characteristics such as wavenumbers and wave speeds are different between the free and water loaded pipes. Based on those findings, we developed a gas accumulation detection method that can also provide quantitative information of the gas accumulation. Through a proof-of-concept test, the quantitative gas accumulation detection method was verified.
The Effect of Torsion on Moment Capacity in Piping Bends
Authors: Azadi-Borujemi Bijan , Asadkarami Ali , Reinhardt Wolf 
Published: November 2015

Piping analysis according to NB-3600 treats bending and torsion loads as equivalent. Piping sections are generally subjected to loading that includes a combination of bending and torsion. Such loading results in a biaxial stress state in the pipe wall. Although NB-3600 would still assign a single bounding stress index to the resultant moment vector in bends and elbows, detailed analysis shows that bending moments of different orientations and torsion loading have a different effect on the calculated stresses.
This paper shows the effect of applied moment direction on the bounding bend limit load that would be evaluated in a detailed (e.g. NB-3200) piping primary stress analysis. This gives the analyst some guidance regarding the potential benefit of a detailed analysis. Further, it is shown that pure bending is not necessarily the worst loading mode in an elbow or bend. The addition of a torsion component could increase the bend stress and lead to a more severe loading condition even if the magnitude of the resultant applied moment vector remains the same.
Rules for Temporary Repair Techniques in JSME Fitness-for-Service Code and Their Challenges
Authors: Dozaki Koji , Kobayashi Hiroyuki , Adachi Hiroyuki , Muraoka Katsuhide , Kono Yusuke 
Published: November 2015

Chapter of Repair, Replacement Activities (RRA) in Fitness-For-Service (FFS) Code of the Japanese Society of Mechanical Engineers (JSME) includes rules of article RB-3000 for temporary repair techniques in the use of covering leakage during operation of a Nuclear Power Plant (NPP). Temporary repair techniques are RB-3010 Plating, RB-3020 Adhesion and RB-3030 Infill. In this paper, rules of these temporary repair techniques are summarized as well as the meaning of ‘temporary’, intention and benefit of them.
On the other hand, these rules of temporary repair techniques were provided by Thermal and Nuclear Power Engineering Society (TENPES) at first in 1986, about thirty years ago. “Plant Operation and Maintenance Standards (POMS) Plan” developed by Japan Power Engineering and Inspection Corporation (JAPEIC) in 1996 incorporated those rules of temporary repair techniques, then JSME FFS Code has taken over them from POMS. Because the design rules of these temporary repair techniques have had much margin since origin, it may result in excessive design, or time-consuming procurement of parts. Especially, since these temporary repair techniques are often applied to the leakage around valve gland and flange, simpler and more practical modified rules could provide more benefit for effective repair activities. In this paper, an orientation of possible revision is described on temporary repair techniques rules in JSME FFS Code.
A Method to Derive the JC Value of a 1T SE(B) Using Charpy Impact Energy in the Lower Ductile to Brittle Transition
Authors: Smith Robin J. , Sherry Andrew H. , Horn Anthony J. , Bannister Adam C. 
Published: November 2015

This paper describes a method by which the elastic-plastic crack driving force, J, of a 1T SE(B) may be calculated using Charpy V-notch absorbed impact energy, Uel+pl,LLD. The method is applicable in the lower ductile-to-brittle transition regime of fracture behaviour and permits the calculation of equivalent critical J-integral values, Jc, using Uel+pl,LLD data. The method is demonstrated using a ferritic steel study material. Comparisons are made between the predictions of a Uel+pl,LLD scaling approach, which was derived using a Weibull stress model, and experimental test data for a ferritic steel. The approach was evaluated using experimental test data composed of instrumented Charpy impact test results and SE(B) fracture toughness test results.
The probabilistic predictions were found to be in good agreement with experimental values. Extension of the methodology is recommended to include other material flow properties. Further work is required to ascertain the accuracy of the approach at higher temperatures in the ductile-to-brittle transition temperature range. A practical method for applying the methodology in practice to allow description of the behaviour of a wide range of ferritic steel materials has been outlined.
Constraint Based Assessments of Large-Scale Cracked Straight Pipes and Elbows
Authors: Gintalas Marius , Ainsworth Robert A. 
Published: November 2015

The paper presents T-stress solutions developed to characterize constraint levels in large-scale cracked pipes and elbows. Stress intensity factor, KI, solutions for pipes and elbows are normalised by material fracture toughness to define the Kr parameter in fitness-for-service procedures, such as R6. Adding knowledge on levels of T-stress allows more advanced analysis through a normalised constraint parameter βT.
The paper presents analyses for 6 pipes and 8 elbows. Values of the normalised constraint parameter βT are calculated for each pipe and elbow at the experimentally measured crack initiation point. Comparison of constraint levels in the pipes and elbows with those in various types of fracture toughness specimen are used to predict the initiation loads using the R6 method and to provide guidelines for transferability.
Probabilistic Pressurized Thermal Shocks Analysis for a Reactor Pressure Vessel by Considering Plume Effect
Authors: Qian Guian , González-Albuixech V. F. , Niffenegger Markus , Sharabi Medhat 
Published: November 2015

The inner surface of a reactor pressure vessel (RPV) is assumed to be subjected to pressurized thermal shocks (PTSs) caused by the downstream of emergency cooling water. The downstream is not homogeneous but typically in a plume shape coming from the inlet nozzles.
In this paper, both deterministic and probabilistic methods are used to assess the integrity of a model RPV subjected to PTS. The FAVOR code is used to calculate the probabilities for crack initiation and failure of the RPV considering crack distributions based on cracks observed in the Shoreham and PVRUF RPVs.
The study shows that peak KI of the cracks inside the plume increases about 33% compared with that outside. The conditional probability inside the plume is more than eight orders of magnitude higher than outside the plume. In order to be conservative, it is necessary to consider the plume effect in the integrity assessment.