bluuebunny/crossref_metadata_2025_split huggingface dataset:
The Law of Crack Growth Life Under Load Controlled Creep-Fatigue Conditions for W-Added 12% Cr Steel
Authors: Nagumo Yoshiko , Yokobori A. Toshimitsu , Fukuda Takahiro , Takahashi Yoshiki , Sugiura Ryuji 
Published: December 2016

W-added 12% Cr ferritic heat resistant steel has been used as a steam turbine rotor material. The turbine rotor material is damaged by high temperature creep and fatigue due to starts and stops and changing load of power generation, which results in crack initiation and growth.
In the studies done before, the law of crack growth life under creep-fatigue conditions was characterized and clarified that the characteristics of crack growth life of various load frequencies under different temperatures change from fatigue to creep behavior through an inflection region. The law of crack growth life under creep-fatigue interactive conditions has been reported and evaluated by monotonous linear law. On the other hand, it has been indicated that the characteristics of crack growth life under creep-fatigue condition can be represented mathematically by a three dimensional characteristic curved surface based on non-equilibrium science.
In this study, crack growth tests using standardized C(T) specimens of W-added 12% Cr ferritic heat resistant steel were conducted under various conditions of stress holding time, applied stress and temperature. To evaluate the effect of cycle dependent and time dependent mechanisms on crack growth life, a method of separate estimation of cycle dependent mechanism from the time dependent mechanism based on the concept of three dimensional characteristic curved surface based on non-equilibrium science were used. As a result, the effect of load frequency on crack growth life was characterized and the predictive law of crack growth life for W-added 12% Cr ferritic heat resistant steel under creep-fatigue interactive conditions based on the concept of Q* with the transition function of crack growth life was estimated.
Comparison of Header Flat End Plate Designs for Operation in the Creep Range
Authors: Cooch Michael , Dewees Dave , Komora George 
Published: December 2016

Section I of the ASME Boiler and Pressure Vessel Code currently has a Code Case that provides rules for the construction of unstayed flat heads with integral flanges (butt welded hubs) from forged material. This Code Case is currently being considered for adoption into the Section I rules; therefore, the existing requirements of the Code Case are under review. In particular, feedback on the Code Case indicated that the required minimum hub length presents manufacturing challenges in certain situations, so the necessity of this hub is being reviewed.
As this type of flat head is commonly used in high temperature operation, evaluation of the expected creep behavior is critical, especially in the proposed weld area at the head junction to the attached shell (typically a header). Example geometries were developed using the current Code Case requirements and their suitability for high temperature operation was initially evaluated using elastic analysis. Detailed inelastic analysis was then performed to further quantify the expected time-dependent high temperature behavior.
Additionally, a series of failures of end plates designed to the TRD code was reported to have occurred in Europe in the 1990s. In order to assess the potential relevance of these failures to heads designed per the proposed Section I rules, additional examples were developed using the design rules of the TRD code rules. The predicted creep behavior and damage for these examples were then compared to the Code Case designed examples.
Reactor Pressure Vessel Integrity Assessment: French Simplified Analysis Method Applied to Underclad Postulated Defect in Nozzle
Authors: Arrieta-Ruiz Adolfo , Meister Eric , Vidard Stéphane 
Published: December 2016

Structural integrity of the Reactor Pressure Vessel (RPV) is one of the main concerns regarding safety and lifetime of Nuclear Power Plants (NPP) since this component is considered as not reasonably replaceable. Fast fracture risk is the main potential damage considered in the integrity assessment of RPV.
In France, deterministic integrity assessment for RPV vis-à-vis the brittle fracture risk is based on the crack initiation stage. As regards the core area in particular, the stability of an under-clad postulated flaw is currently evaluated under a Pressurized Thermal Shock (PTS) through a dedicated fracture mechanics simplified method called “beta method”. However, flaw stability analyses are also carried-out in several other areas of the RPV. Thence-forward performing uniform simplified inservice analyses of flaw stability is a major concern for EDF.
In this context, 3D finite element elastic-plastic calculations with flaw modelling in the nozzle have been carried out recently and the corresponding results have been compared to those provided by the beta method, codified in the French RSE-M code for under-clad defects in the core area, in the most severe events.
The purpose of this work is to validate the employment of the core area fracture mechanics simplified method as a conservative approach for the under-clad postulated flaw stability assessment in the complex geometry of the nozzle.
This paper presents both simplified and 3D modelling flaw stability evaluation methods and the corresponding results obtained by running a PTS event. It shows that the employment of the “beta method” provides conservative results in comparison to those produced by elastic-plastic calculations for the cases here studied.
Consideration of Irradiation Effect in Components Design: Illustration of a Way to Introduce an Environmental Effect in the Design Process
Authors: Pétesch C. , Lebarbé T. , Pascal C. , Dubiez-Le Goff S. 
Published: December 2016

During the last years, a particular attention has been paid to the consideration and evaluation of environmental effects in the design of mechanical components, particularly for the consequences of Pressure Water Reactors environment on fatigue damages. In the frame of RCC-MRx design code, developed for Sodium Fast Reactors (SFR), Research Reactors (RR) and Fusion Reactors (FR-ITER), the specific features are creep damage (covered by the design rules responding to the needs of fast breeder reactors and high temperatures) and irradiated material (covered by the design rules responding to the needs of research reactors developed for the JHR project).
The purpose of this document is to detail the way to consider an environmental effect such irradiation in the material design rules. It covers:
- the background for these rules, linked with the changes in materials behavior,
- their content, in term of damage impact,
- and the challenges to be met for their development to other applications.
Evaluation of the Lowest Temperature for Ductile Crack Initiation in an Irradiated RPV
Authors: Wilkowski G. , Brust F. W. , Uddin M. , Kalyanam S. , Lacroix V. 
Published: December 2016

In this paper, work done for defining the lowest temperature for ductile fracture initiation in piping and fittings was extended to a sample reactor pressure vessel (RPV). The methodology used is called “Master Curve of Fracture Transition Temperatures” and was developed from correlations of thousands of laboratory tests and hundreds of pipe tests originally presented in 2005. Since then it has been extended to; low and high strength line-pipe steel base metals (oil/gas industry), blunt flaws as well as sharp cracks, girth welds for the oil/gas industry, fracture of pipe fittings/valves, and is the technical basis for the lowest temperature for ductile crack initiation in ferritic piping in Appendix C of ASME Section XI. Since the methodology is quite different than traditional approaches for nuclear component applications, the general methodology will be reviewed, as well as analysis results showing how surveillance capsule Charpy data could be used to predict the lowest temperature where ductile crack initiation would occur in RPVs. Once this temperature is established, then the upper-shelf toughness values can be used to determine if the failure is EPFM or limit-load, and the associated failure stresses. This temperature could be used for defining the pressure-temperature limits to assure that the RPV material has a high flaw tolerance. This methodology was proposed for the Doel 3 and Tihange 2 RPVs in Belgium.
One key concern for operation of RPVs is determining the lowest operating temperature where ductile crack initiation is still expected which can be based on studies of irradiation effects on the measured toughness from Charpy tests of material in surveillance capsules. With ductile crack initiation the flaw tolerance is quite high even with long-term irradiation damage to the material. In fact, if the toughness is closer to limit-load than LEFM in a FAD analysis, then as long as there is ductile initiation, the irradiation effects increase the strength which could increase the flaw tolerance. There are some ongoing efforts within the ASME Section XI activities to define the minimum temperature where ductile initiation occurs in fracture toughness testing (typically based on a mixture of Charpy and 1T CT specimen data); however, there are still thickness and constraint effects on the fracture toughness for more precise application to a thick-walled vessel with a postulated axial surface crack.
Sensitivity Study of Boundary Conditions for a 9-Meter Drop of a Shielded Cask for Operational Handling of a Container of Liquid
Authors: Yaksh Michael C. , Babu Suresh 
Published: December 2016

A shielded cask is used to move welded containers whose contents are liquid. The requirement controlling the design of the shielded cask was the 9-meter drop. Since the orientation of the cask is arbitrary, it is required to assume an orientation which would result in the most damage to the cask. For such drops the target is usually considered to be an unyielding surface. The shielded cask of interest is not designed with components to mitigate the damage due to such drops. The shielded cask contains a thick shell of lead which is considered to move and deform during the 9-meter drop. Additionally, the container of liquid is not physically attached to the cask, and is free to move within the confined space during the event as well. Each component has its own unique stiffness and mass characteristics which could result in a different dynamic response. Since the dynamic response of each component is different, the most damage to a particular component may be sensitive to the boundary conditions. The unyielding surface would maximize the damage to the impact surface of the cask, but as a result, could mitigate the maximum loading applied to other components of the cask. Most actual targets are comprised of concrete. The evaluations are performed using an explicit finite element computer code. Consequently, it is necessary to monitor certain energies, such as the hourglass energy or a sliding energy indicating the behavior of the contact surface associated with the target. These parameters confirm the accurate behavior of the elements comprising the finite element model. Given that components can have a different response, the hourglass energy may also vary. Varying the boundary conditions will affect these types of parameters.
In this paper, the authors present the results of a study of the effect of the boundary conditions on the shielded cask components response to the 9-meter drop. The primary orientation of interest is the end drop. The end drop maximizes the axial loading to the container. It is this orientation which could result in the most compression of the lead shield leading to increased radiation exposure. The container is considered as a pressure vessel and its integrity would be evaluated using the plastic strain based criteria contained in Section III, Division 3. The shielded cask, however, is not a pressure vessel and was evaluated using Section VIII, Division 2, Part 5. Both evaluations used the plastic strains and triaxiality factors determined from the drop evaluations.
Regulatory Perspective for the Definition of Probabilistic Acceptance Criteria for CANDU Pressure Tubes
Authors: Laxman Sankar , Carroll Blair , Jin John 
Published: December 2016

For the assurance of fitness-for-service of CANDU Pressure Tubes (PTs), guidelines and acceptance criteria are provided in Canadian Standard Association (CSA) N285.8-15, Technical requirements for in-service evaluation of zirconium alloy pressure Tubes in CANDU reactors. With respect to the assessment of risk of operation associated with degradation mechanisms and aging of the PTs in the entire core of a given reactor Unit, Clause 7 of CSA N285.8 allows Licensee’s to use either a deterministic or probabilistic method to assess the likelihood of PT failures. When a probabilistic method is used, the Licensee is obligated to demonstrate that the combined frequency of PT failure(s) over the evaluation period, due to the various potential degradation mechanisms, is less than the maximum acceptable frequency provided in Table C.1 of CSA N285.8-15.
The maximum acceptable frequency provided in Table C.1 of CSA N285.8-15 was developed in the early-1990’s based on reactor operating experience and knowledge at that time, Station Siting Guides and Consultative Regulatory Guide C-006 (Revision 1). A task group was established by the CSA N285.8 Technical Steering Committee to re-evaluate the allowable failure frequencies to confirm that they remain relevant given the current state of knowledge and the additional evaluation tools available. This paper provides Canadian Nuclear Safety Commission staff views regarding the technical basis for revisions to the allowable frequencies based upon current industry practices in conducting probabilistic core assessments.
Nonlinear Finite Element Analysis of Pipe Bends Subjected to Inplane Bending
Authors: Akula Venkata M. K. , Martin David W. 
Published: December 2016

Pipe bends are structural components that provide flexibility to accommodate length changes in pipelines while allowing fluid flow. Estimating the collapse load is critical to ensuring the structural integrity of the pipeline. This research discusses modeling and analysis of pipe bends utilizing the finite element method. Three-dimensional models utilizing elbow elements, shell elements, and brick elements are generated to predict the collapse moment of pipe bends subjected to in-plane loading. All simulation was performed using Abaqus.
To obtain a more physically-consistent response, material, geometric, and boundary nonlinearities are all included. A MPC user subroutine is utilized to capture the end behavior of the pipe bends correctly when utilizing shell and brick elements. Experimental data from two sources, available in literature, was used to evaluate the effect of the different element types on the predicted structural response.
Finally, utilizing the shell, brick, and the elbow elements, parameter sensitivity analyses are performed to identify the key parameters influencing the response of pipe bends. Multiple parameters are varied independently of each other to fully understand and capture their influence on the response. SIMULIA’s Isight software was used to automate the workflow and vary the model parameters about their respective baseline values.
The Effects of Discretization Errors on the High Frequency Content of In-Structure Response Spectra
Authors: Mertz Greg , Spears Robert , Houston Thomas 
Published: December 2016

The next generation ground motion prediction equations predict significant high frequency seismic input for rock sites in the Central Eastern United States (CEUS). This high frequency motion is transmitted to basemat supported components and may be transmitted to components supported on elevated slabs. The existing ASCE 4 analysis requirements were initially developed based on seismic motions having lower frequencies, typical of ground motions in the Western United States (WUS). The adequacy of the existing ASCE 4 analysis requirements are examined using high frequency CEUS spectral shapes and the potential error inherent in using the existing approach to computing in structure response spectra is quantified. Modifications to reduce potential error in the existing ASCE 4 criteria are proposed.
In structure response spectra are typically generated for a subsystem given the time history response of a building region. The building time history response is based on analyses that use either modal time history superposition, direct integration or complex frequency response analysis of the building and supporting soil. Input to the building analyses consist of either real or synthetic discretized ground motion records. The discretized ground motion records are often based on recorded ground motion seeds and are often limited to a 0.005 second time step. Thus the time step of the seed record often limits the frequency content of the problem.
Both the building analyses and in structure response spectra subsystem analysis may interpolate the discretized ground motion records to obtain stable results. This interpolation generates errors that are propagated through the analyses used to calculate in structure response spectra. These errors may result in extraneous high frequency content in the in structure response spectra. Errors are quantified by comparison of time history parameters, Fourier components and in structure response spectra.
Pipeline Design Method Against Large Displacement of Strike-Slip Fault
Authors: Oda Keita , Ishihara Takahiro , Miyajima Masakatsu 
Published: December 2016

This study proposes a method for designing a water pipeline system against fault displacement by incorporating earthquake resistant ductile iron pipes (ERDIPs). An ERDIP pipeline is capable of absorb the large ground displacements that occur during severe earthquakes by movement of its joint (expansion, contraction and deflection) and the use of the joint locking system. Existing ERDIP pipelines have been exposed to several severe earthquakes such as the 1995 Kobe Earthquake and the 2011 Great East Japan Earthquake, and there has been no documentation of their failure in the last 40 years.
In the case of a pipeline that crosses a fault, there is the possibility of the occurrence of a local relative displacement between the pipeline and the ground. It is known that an ERDIP pipeline withstands a fault of axial compression direction by past our study. Hence, this present study was targeted at developing a method for designing an ERDIP pipeline that is capable of withstanding a strike-slip fault of axial tensile direction for a pipeline. This was done by FEM analysis wherein the ERDIPs and spring elements were used to model the soil and ERDIP joints.
An ERDIP pipeline can accommodate a fault displacement of about 2 m by joint expansion/contraction and deflection, while maintaining the stress in the pipeline within the elastic limit. However, additional countermeasure is required when the fault displacement exceeds 2 m because such could stress the pipeline beyond the elastic limit.
The use of large displacement absorption unit is an effective countermeasure for displacements exceeding 2 m. The expansion/contraction capacity of a unit is 10 times that of an ERDIP joint and it is able to absorb a locally-concentrated axial displacement of the pipeline. It was confirmed in the present study that an ERDIP pipeline with large displacement absorption unit, referred to as a large displacement absorption system, could accommodate fault displacement in excess of 2 m within the elastic stress range of the pipeline.
Deterministic and Probabilistic PTS Study for a Reactor Pressure Vessel Considering Plume Cooling Effects
Authors: Qian Guian , Niffenegger Markus , Sharabi Medhat , Lafferty Nathan 
Published: December 2016

A reactor pressure vessel (RPV) is assumed to be subjected to pressurized thermal shocks (PTSs) as a result of the emergency cooling water injected during a loss-of-coolant accident (LOCA). The cooling flow is not homogeneous but typically in a plume shape (stripe cooling) flowing from the cold leg through the inlet nozzles. This paper aims to analyze the non-uniform cooling effect on the RPV integrity.
In this paper, both deterministic and probabilistic methods are used to analyze the integrity of a model RPV subjected to PTS. RELAP5, GRS-MIX, CFD and other semi-analytical methods are used to analyze the transient with and without considering plume cooling effect. Finite element method (FEM), extended finite element method (XFEM) and weight function method are used to calculate KI of the postulated cracks. The FAVOR code is used to calculate the conditional probabilities for crack initiation and failure of the RPV considering different crack distributions.
KI based on CFD input is the highest, followed by that based on reference transient, GRS-MIX and RELAP5. Peak KI of the cracks inside the plume increases about 33% compared with that outside. According to the maximum criteria, the maximum allowed RTNDT are 56.9 °C, 90.2 °C, 98 °C, 115.7 °C and 136.2 °C for the crack in the nozzle region based on CFD transient, the cracks in the ring region based on the CFD, reference data, GRS-MIX and RELAP5 calculated transient, respectively. These values are 36 °C, 68.5 °C, 73 °C, 81 °C and 104 °C according to the tangent criteria. The conditional probability inside the plume is more than nine orders of magnitude higher than outside the plume. Considering plume cooling effects increases the total failure frequency by 1–2 orders of magnitude. In order to be conservative, it is necessary to consider the plume effect in the integrity assessment.
Root Cause Analysis of the Unexpected Behaviour of a Flaked Material Under Irradiation and Transferability to the Doel 3/Tihange 2 Reactor Pressure Vessels
Authors: De Bruycker Evy , De Vroey Séverine , Hallet Xavier , Stubbe Jacqueline , Nardone Steve 
Published: December 2016

During the 2012 outage at Doel 3 (D3) and Tihange 2 (T2) Nuclear Power Plants (NPP), a large number of nearly-laminar indications were detected mainly in the lower and upper core shells. The D3/T2 shells are made from solid casts that were pierced and forged.
Restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes. These tests showed unexpected results regarding the shift in the Reference Temperature for Nil Ductility Transition (RTNDT) of the flaked material VB395 (Steam Generator shell rejected because of flakes) after irradiation.
This paper presents the root cause analysis of this unexpected behaviour and its transferability (or not) to the D3/T2 Reactor Pressure Vessels (RPVs).
A mechanistic and a manufacturing based approach were used, aiming at identifying the microstructural mechanisms responsible for the atypical embrittlement of VB395 and evaluating the plausibility of these mechanisms in the D3/T2 RPVs.
This work was based on expert’s opinions, literature data and test results. Both flaked and unflaked samples have been investigated in irradiated and non-irradiated condition.
All hydrogen-related mechanisms were excluded as root cause of the unexpected behaviour of VB395. Two possible mechanisms at the basis of the atypical embrittlement of VB395 were identified, but are still open to discussion. These mechanisms could be linked to the specific manufacturing history of the rejected VB395 shell.
Since the larger than predicted shift in transition temperature after irradiation of VB395 is not linked with the hydrogen flaking and since none of the specific manufacturing history features that are possible root causes are reported for the D3/T2 RPVs, the D3/T2 shells should not show the unexpected behaviour observed in VB395.