bluuebunny/crossref_metadata_2025_split huggingface dataset:
Contribution to Safety Enhancement for BDBE in Structure and Material Fields
Authors: Kasahara Naoto , Sato Takuya , Blahoianu Andrei 
Published: October 2018

This paper proposes research issues on contribution to safety enhancement for BDBE in structure and material fields.
There are large difference between DBE and BDBE. Objective of DBE is prevention of accident and conservative approach is adopted such as prevention of all assumed failure modes. As for BDBE, objectives are prevention of safety function loss and mitigation of accident consequences, risk approach is expected. Since DEC is a part of BDBE, approach against DEC is different from DBE. DEC requires best estimate.
Probabilistic Risk Assessment consists of best estimate plus uncertainty.
For stress test, identification of the weakest portions and cliff edges become important.
To realize above approach, prediction of realistic failure modes is essential. Furthermore, relative strength evaluation becomes important to predict order of failure location and their mode, even though absolute strength is not clear.
After Fukushima daiichi nuclear accident, there is a tendency to apply DBE design criteria to BDBE, however, conservative criteria for design are inappropriate for best estimate. Therefore, authors proposes failure mode maps to identify realistic failure mode and its application to mitigation to accident sequences such as fracture control.
Optimization Design of PDC Bit Gauge Using CFD-DPM and Central Composite Design
Authors: Hou Huihui , Liu Zhong , Zhu Hongwu , Zhang Boyu , Wu Tingting 
Published: October 2018

Polycrystalline Diamond Compact (PDC) bit has been widely used in natural gas/oil drilling industry for its advantages such as no rotating parts, highly rate of penetration (ROP) and durability. In recent decades, the comprehensive property of PDC bit has been improved rapidly due to the scientists’ efforts. Hydraulic design of PDC bit is very important because the retention of the cuttings may dramatically decrease ROP. Most current researches focus on how blade profile and nozzle parameters affect hydraulic performance of PDC bit, but, the understanding of gauge geometrical parameters (mainly refer to gauge width, gauge length and gauge spiral angle) effects on hydraulic performance of PDC bit is still incomplete.
In this paper, a simulation method and an optimization method were applied to find the best gauge parameter for 8.5 inch PDC bit. In the proposed approach, the approximate range of above three parameters were confirmed first based on previous studies. Then, the central composite design (CCD) was used to design numerical simulation scheme and the discrete particle modeling (DPM) was used to track the cuttings and the velocity of particles and fluid were evaluated to study the hydraulic performance of drill bit under different gauge geometrical parameters. Finally, the optimal geometrical parameter of gauge could be obtained according to the simulation results. Overall, the proposed approach can be used to design PDC bit gauge structure to improve the cuttings’ velocity away from bottomhole, which will result in huge improvement of ROP.
Stress Intensity Factor Coefficients for Circumferential OD Surface Flaws in Cylinders for Appendix A of ASME Section XI
Authors: Lee Darrell R. , Cipolla Russell C. , Liu Michael C. 
Published: October 2018

Linear elastic fracture mechanics based flaw evaluation procedures in Section XI of the ASME Boiler and Pressure Vessel Code require calculation of the stress intensity factor (KI). The 2015 Edition of ASME Section XI [1] implemented a number of significant improvements in Article A-3000, including closed-form equations for calculating stress intensity factor influence coefficients (Gi) for circumferential flaws on the inside surface of cylinders. In the 2017 Edition [2], closed-form equations for axial flaws on the inside and outside surfaces of cylinders have been implemented.
In this paper, closed-form equations are developed for circumferential cracks on the OD surface of cylinders, based on tabular data from API 579 (2007 Edition) [3]. The equations presented, represent a complete set of Ri/t, a/t, and a/ℓ ratios. The closed-form equations provide G0 and G1 coefficients while G2 through G4 are obtained using a weight function representation for the KI solutions for a surface crack. These equations permit the calculation of the Gi coefficients without the need to perform tabular interpolation. The equations are complete up to a fourth order polynomial representation of the applied stress. The fourth-order representation for stress will allow for more accurate fitting of highly non-linear stress distributions, such as those depicting high thermal gradients and weld residual stress fields.
Fatigue Crack Growth in Low Alloy Steels Under Tension-Compression Loading in Air
Authors: Azuma Kisaburo , Yamazaki Yasuhiro 
Published: October 2018

Low alloy steels are extensively used in pressure boundary components of nuclear power plants. The structural integrity of the components made of low alloy steels can be evaluated by the procedure of flaw evaluation provided by Section XI of the ASME Boiler and Pressure Vessel Code. According to the Code, the range of stress intensity factor ΔK can be used to determine the fatigue crack growth rates of the material. However, it has been reported that crack closure behavior also strongly influence the fatigue crack growth rate under strong compressive load cycles.
This paper discusses the relation between ΔK and the fatigue crack growth rate for cracks in low alloy steels exposed to air. Compressive-tensile cyclic loadings were applied to center-notched plates to obtain the fatigue crack growth curves. The test data demonstrated that effective SIF range ΔKeff more accurately described the crack growth property due to plasticity induced crack closure. Comparing the test results with the reference crack growth curves in the ASME Code Section XI, it may seem that the crack growth prediction based on the Code underestimates the crack growth rates for compressive-tensile cyclic loadings under high stress level.
Flamanville EPR RPV’s Heads Carbon Segregation: Vessel Heads Manufacturing
Authors: Lefever Benoit , Marie Stéphane , Berger Thierry , Bobin Vastra Isabelle 
Published: October 2018

The FA3 RPV heads segregation has been an important issue for the Flamanville 3 EPR project. A high carbon content of carbon has been measured at the outer central part of the two heads. Manufactured in 2006 according to the RCC-M requirements, these pieces were also submitted to the ESPN (i.e. nuclear pressurized equipment) French order requirements, and in particular the ESPN technical qualification.
This paper presents the manufacturing of these heads to explain the origin of the segregation and place this process in its context, with in particular a clarification of the technical qualification requirements in 2011, which are different to the RCC-M qualification process. This qualification required additional chemical and mechanical characterizations in the segregated area of the representative vessel head. The additional Charpy-V test did not meet the 60 J requirement of the ESPN order and led to the large testing program to characterize the toughness properties of the segregated material.
Study on Promoting Use of Repair, Replacement and Mitigation Techniques in Maintenance Activities of Nuclear Power Plants
Authors: Dozaki Koji , Aoki Takayuki , Koyama Koji , Kanno Masanori , Saeki Ryoichi 
Published: October 2018

Maintenance of nuclear power plant facilities involves activities comprising a large system composed of both plant hardware and human subsystems to assure safe and reliable operation. Maintenance activities are composed of inspection, evaluation and corrective measures. Corrective measures are countermeasures for aging degradation, e.g., resetting the inspection period based on the results of inspection and evaluation; mitigation of degradation phenomenon; repair or replacement; preventive maintenance; etc. The corrective measures merit special attention as they are important and valuable actions in order to promote continued efficient and safe plant operations. It is necessary to develop a set of regulatory and industrial technical requirements for a well-structured, documented set of standards, so that corrective measures can be used and applied uniformly and effectively. Currently the code and standard system is less developed in Japan than in the United States.
In this study, the authors considered the relationship between degradation and maintenance and the difference of performance requirements between the plant construction stage and the in-service stage. This effort is intended to clarify the issues of regulation for maintenance activities, with an objective to help develop structured regulatory/industrial requirements with a code and standards consistent with appropriate corrective measures.
The Nuclear Regulation Authority (NRA), the regulatory body in Japan, has reviewed the present Japanese inspection system in response to suggestions from the Integrated Regulatory Review Service (IRRS) mission established by International Atomic Energy Agency (IAEA). The NRA has also been developing a new regulatory inspection system similar to the Reactor Oversight Process (ROP) used in the United States. The expectation for the new Japanese inspection system is to focus regulations on plant issues with higher risk importance, considering both plant hardware and human subsystems. The new Japanese regulatory system addressing maintenance is also expected to enhance electric utilities ability to assure safety is self-motivated and sustained.
SCC Crack Growth Analysis for a Dissimilar Metal Weld Using Advanced FEA
Authors: Ogawa Takuya , Itatani Masao , Hayashi Takahiro , Saito Toshiyuki 
Published: October 2018

Management of plant service life is a key issue for improving the safety of light water reactors. Some incidents of primary water stress corrosion cracking (PWSCC) of pressurized water reactor (PWR) components, such as a primary loop piping/nozzle weld, and intergranular stress corrosion cracking (IGSCC) of boiling water reactor (BWR) components, such as a shroud support weld, have been reported in the past. When a crack is detected, crack growth analysis is required as part of the structural integrity assessment of the component with the crack.
In Japan, the “Rules on Fitness-for-Service for Nuclear Power Plants” of the Japan Society of Mechanical Engineers (JSME FFS Code) describes the conventional methodology for analyzing crack growth. The methodology assumes a semi-elliptical crack shape and is based on crack growth calculation at only the deepest and surface points of the crack. However, the actual crack growth behavior is likely to be very different from that analyzed by the conventional methodology due to the complex distribution of residual stress and dependency of crack growth properties on the materials composing the weld joint, particularly in the case of cracks in a dissimilar metal weld. Recently, crack growth analysis techniques using finite element analysis (FEA) have been used to analyze crack growth behavior in more detail.
In this study, a program code was developed for SCC crack growth analysis that consists of fracture mechanics analysis by “ABAQUS”, crack growth calculation and automatic remesh of the FE model by in-house code. Case studies of SCC crack growth analysis for a dissimilar metal weld were performed and the analysis results were compared with those obtained by the conventional methodology. As a result, it was confirmed that the conventional methodology provides a conservative estimation of crack growth behavior. It was also found that the difference in crack growth properties of individual materials composing the weld joint had a significant effect on the crack growth behavior, particularly on a dissimilar metal weld. Furthermore, the effect of the material anisotropy of the SCC crack growth rate for the weld metal on the crack growth behavior was investigated.
Leak Rate Computation: Flow Resistance vs. Thermal-Hydraulic Aspect
Authors: Heckmann Klaus , Sievers Jürgen , Weyermann Fabian 
Published: October 2018

The computation of mass flow rates through crack-like defects in piping systems of light water reactors requires typically the description of two-phase flow conditions. The computed discharge rate depends on the crack opening area, the thermal-hydraulic modeling of the flow, and the flow resistance of the crack. Several models have been proposed to characterize the critical flow through crack-like defects. An evaluation of advantages and shortcomings of the different models with regard to the interaction of the three different parts (crack opening area, thermal-hydraulic modeling, flow resistance) has been performed.
In this paper, the flow resistance modeling from several approaches is discussed, and compared with a database from eight different testing programs. Five different flow models are applied to analyze a database of more than 800 leak rate measurements for subcooled water from twelve different experimental programs. It is shown that the correct modeling of the flow resistance is crucial for a best estimate reproduction of the measured data. It turns out that generally, equilibrium models are about as good as non-equilibrium models. The data were processed with the GRS software WinLeck which includes different analytical approaches for the calculation of crack sizes and leak rates in piping components. The most reliable results within the model selection are produced by the CDR model (Critical Discharge Rate) of the ATHLET code (Analysis of Thermal-hydraulics of Leaks and Transients) developed by GRS.
As a conclusion, the accurate modeling of form losses and frictional pressure losses for critical discharge flow rates through crack-like leaks are essential for a reliable prediction of flow rates. Uncertainties in leak rate computations results arise due to the lack of information about the flow geometry and its associated drag. The assumed flow resistance of a through-wall crack influences the computed leak rate as significant as the phase-change- and flow-models. The manifest difference between equilibrium models (Pana, Estorf) and non-equilibrium models (Henry, ATHLET-CDR) seems to be less significant than the pressure loss issue. One can conjecture that, for crack-like through-wall defects, friction effects play a more important role than non-equilibrium effects.
StrainLife: Efficient Fatigue Life Data Generation for an Enhanced Ageing Assessment of Metallic Components
Authors: Heckmann Klaus , Sievers Jürgen , Schopf Tim , Schuler Xaver , de Acosta Ruth , Starke Peter , Boller Christian , Jamrozy Michael , Knyazeva Marina , Walther Frank 
Published: October 2018

Pressurized pipes in hot water environment are subject to ageing mechanisms such as fatigue and environmental-assisted fatigue. These ageing effects limit the in-service time of components due to the possibility of crack formation, initiation and growth. Furthermore, uncertainties in life time assessment evolve as a consequence of increased scattering and resulting deviations in material properties. The lack of appropriate information requires safety-oriented design and conservative margins in the acceptable operation time.
In this contribution, investigations on methods for an improved assessment of fatigue life assessment based on microstructural aspects (specifically in the view of piping systems) are presented. The focus is on the metastable austenitic steel AISI 347 (X6CrNiNb18-10) under boiling water reactor operational conditions. In addition to strain-controlled fatigue tests and the hysteresis information, microstructural characterizations of the damage evolution are performed. Fatigue tests are equipped with magnetic and resistometric sensors to measure a different kind of material response correlated with the damage states.
The information obtained is used for an improved derivation of fatigue life evaluations of piping systems. The StrainLife approach and its specific instrumentation is proposed and discussed. In this approach, the number of strain-controlled fatigue tests is basically reduced by an order of magnitude when compared to traditional procedures. A software tool (as a module of the structural analysis code PROST) has been developed to assist users in evaluating and incorporating the respective data and to derive the subsequent fatigue life curves. As an outlook, further development is addressed to transfer the methodology to the assessment of residual fatigue life of components being exposed to ageing in operation.
Advanced Structural Integrity Assessment Tools for Safe Long Term Operation (ATLAS+)
Authors: Lindqvist Sebastian , Wallin Kim , Moinereau Dominique , Smith Mike , Marie Stéphane , Dillström Peter , Keim Elisabeth , Szavai Szabolcs 
Published: October 2018

The main objective and mission of the ATLAS+ project is to develop advanced structural assessment tools to address the remaining technology gaps for the safe and long term operation of nuclear reactor pressure coolant boundary systems. This is achieved by development and validation of:
• innovative quantitative methodologies to transfer laboratory material properties to assess the structural integrity of large components,
• enhanced treatment of weld residual stresses when subjected to long term operation,
• advanced simulation tools based on fracture mechanics methods using physically based mechanistic models,
• improved engineering methods to assess components under long term operation taking into account specific operational demands,
• integrated probabilistic assessment methods to reveal uncertainties and justify safety margins.
Additionally, the objective is to disseminate the findings of the work through special training sessions and links to the NUGENIA association.
The project scope of work focuses on piping systems of the reactor coolant pressure boundary components (RCPB) excluding the reactor pressure vessel (RPV). The project is aimed on an experimental proof of concept and validates the developed methodology both at the laboratory scale and the full scale level.
The ATLAS+ project contains 4 main technical work packages and one training and dissemination package. These are summarised here.
Investigation Into Applications of Local Failure Criterion for X70 Pipeline With Corrosion Defect
Authors: Moon Ji-Hee , Huh Nam-Su , Kim Ki-Seok 
Published: October 2018

In this paper, the local failure criterion using stress modified critical strain method based on annex B of API 579 is applied to evaluate the ductile failure of API X70 pipelines with a volumetric corrosion defect. Ductile failure is quantified in terms of strain, representing the tensile strain capacity (TSC) which is commonly used in strain-based assessment for fitness-for-service of pipelines installed in frozen area where large-scale ground movement can arise due to earthquakes, freezing and thawing.
Based on the local failure criterion suggested for API X70 steel material, the TSCs of the corroded pipelines are evaluated by using the detailed finite element (FE) analyses. The effects of internal pressure and defect size (such as longitudinal length, circumferential width and depth in the direction of thickness) on TSC of pipelines subjected to axial displacement are systematically investigated. In addition, TSCs based on local failure criterion are compared with those based on net-section limit load. The TSCs from the present FE analyses for various defect geometries and internal pressure can be used to predict ductile failure of corroded pipelines and to build the framework for a strain-based assessment for in-service pipelines.
Comparison Between Strain-Based and Energy-Based Creep Failure Simulation
Authors: Kim Seung Jae , Oh Young Ryun , Kim Yun Jae 
Published: October 2018

The power plant is required to operate under high temperature and pressure for high efficiency. In order to predict reliable life time of power plant under high temperature, creep-low cycle fatigue life prediction method should be proposed.
In this paper, strain based and energy based failure model are proposed to simulate notch bar creep tensile test. Modification factors considering multiaxial fracture and strain rate effect were proposed in order to simulate notch bar creep tensile test using FE analysis.
Using proposed models, FE result of strain based and energy based damage model are compared with notch bar creep tensile test. As a result, both strain and energy based damage model simulates crack growth well during creep, However, when tertiary creep behavior is considered, energy based failure model simulate rupture time longer than strain based model. It can be inferred that plastic damage accumulation of energy based model is slower than that of strain based model.
Approach for the Dimensioning of Bolted Joint Connections With Threaded Flanges for High Pressure Applications
Authors: Bortz Matthias , Fuenmayor David , Wink Rolf 
Published: October 2018

High pressure flange connections for LDPE plants are commonly designed using lens ring gaskets and bolted flange connections. The rules given in the ASME Code Sections B31.3, VIII Div. 2 or VIII Div. 3 for bolted-flange joints offer no guidance for the calculation of threaded flanges.
In this paper, an approach is presented that considers internal pressure and external loads on a flange connection and fulfills the requirements of ASME Section VIII Div.3 and ASME B31.3. Of specific interest is the dimensioning of bolts due to the bending moment with a method that avoids the over-conservatism of the rules given in ASME VIII Div.2 and ASME B31.3. Furthermore, a simplified calculation for threaded flanges considering the stress classification and stress limits given in ASME VIII Div. 3 and using Finite Element Analyses is described.
Presentation of DEFI-PROSAFE Project: Probabilistic Methodology to Assess Margin in Deterministic RPV Integrity Assessment and Proposed Benchmark
Authors: Tiete Ralf , Trewin Richard , Blasset Sebastien , Pistora Vladislav , Posta Miroslav , Niffenegger Markus , Qian Guian 
Published: October 2018

In this paper the project DEFI-PROSAFE is presented.
In the frame of Nugenia+ project, a work package was dedicated to the “DEFInition of reference case studies for harmonized PRObabilistic evaluation of SAFEty margins in integrity assessment for long-term operation of reactor pressure vessel” (acronym DEFI-PROSAFE). A methodology is proposed to assess safety margins in RPV integrity accounting for uncertainties propagation, because no commonly accepted European approach exists for probabilistic assessment of RPV fast fracture risk.
The DEFI-PROSAFE methodology, which is based on the comparison between deterministic and probabilistic assessment, will be detailed. The experience gained from the US Screening Criteria (NUREG-1806) and past projects (ICAS, PROSIR), and guideline (IAEA TecDoc 1627) as well as aspects specific to European deterministic integrity approach have been considered. Usually for probabilistic fracture assessment of the RPV, the parameters (describing flaws, material and neutron fluence) are sampled and the RPV cylindrical region is assessed using deterministic thermal hydraulic loading for evaluation of initiation or failure conditional probability. Within DEFI-PROSAFE methodology the uncertainties in thermal-hydraulic input parameters are taken into account and their propagation in the structural assessment is considered. Comparison of RELAP5 thermal hydraulic results and mixing calculation results (KWU-MIX) with experiment relevant to PTS assessment was performed. Furthermore the DEFI-PROSAFE methodology considers RPV discontinuity regions (like RPV nozzle) and specific PIRT analysis has been performed for selection of the TH-parameters.
A new benchmark for probabilistic assessment of RPV was defined within the DEFI-PROSAFE project. The benchmark definition and case studies based on previous R&D project (ICAS, PROSIR) will be presented. The performance of the new benchmark may be submitted as new European project.