bluuebunny/crossref_metadata_2025_split huggingface dataset:
Evaluation of Flange Calculations Using Strain-Based Acceptance Criteria
Authors: Mutz Alexander , Schaaf Manfred 
Published: November 2019

Abstract
The Nuclear Power Plant KKG in Gösgen, Switzerland was designed according to the ASME Boiler and Pressure Vessel Code. The ASME BPVC, Section III, Appendix 11 regulates the flange calculation for class 2 and 3 components, it is also used for class 1 flanges. A standard for the determination of the required gasket characteristics is not well established which leads to a lack of clarity. As a hint different y and m values for different kinds of gasket are invented in ASME BPVC Section III [1].
The KTA 3201.2[2] and KTA 3211.2[3] regulate the calculation of bolted flanged joints in German nuclear power plants. The gasket characteristics required for these calculation methods are based on DIN 28090-1[4], they can be determined experimentally. In Europe, the calculation code EN 1591-1 [5] and the gasket characteristics according to EN 13555[6] are used for flange calculations. Because these calculation algorithms provide not only a stress analysis but also a tightness proof, it would be preferable to use them also in the NPP’s in Switzerland. Additionally, for regulatory approval also the requirements of the ASME BPVC must be fullfilled.
For determining the bolting up torque moment of flanges several tables for different nominal diameters of flanges using different gaskets and different combinations of bolt and flange material were established. As leading criteria for an allowable state, the gasket surface pressure, the allowable elastic stress of the bolts and the strain in the flange should be a good and conservative basis for determining allowable torque moments. The herein established tables show only a small part according to a previous paper [7] where different calculation methods for determining bolting up moments were compared to each other.
In this paper the bolting-up torque moments determined with the European standard EN 1591-1 for the flange, are assessed on the strain-based acceptance criteria in ASME BPVC, Section III, Appendices EE and FF. The assessment of the torque moment of the bolts remains elastically which should lead to a more conservative insight of the behavior of the flanges.
An Efficient Modelling Approach for Predicting Residual Stress in Power-Beam Welds
Authors: Horne Graeme , Thomas Danny , Collett Andrew , Clay Andrew , Cott Martin , Moffat Andrew 
Published: November 2019

Abstract
The prediction of welding residual stress in components is often an important input to structural integrity assessments. An efficient modelling approach was developed for predicting residual stress in power-beam welds, including validation against residual stress measurements. Specifically, sequentially coupled thermo-mechanical finite element analysis was conducted using a simplified heat source that was tuned to the observed fusion zone from a weld macrograph and thermocouple data for a series of electron beam welds in 316L austenitic stainless steel with a variety of geometries. The predicted residual stresses were compared with contour method and neutron diffraction residual stress measurements.
Development of GTN Model Parameters for Simulating Ductile Fracture Behavior of X 70 Carbon Steel SENT Specimens
Authors: Yoon Sung Ho , Ryu Tae-Young , Kim Moon Ki , Choi Jae-Boong , Kim Ik-Joong 
Published: November 2019

Abstract
Single-Edge-Notched-Bending (SENB) specimen is mainly used for fracture characteristics test of pipe material. However, there is also a disadvantage in that it does not sufficiently simulate the fracture characteristics of thin pipes due to the difference in boundary effect between the SENB specimen and the actual pipe. The Single-Edge-Notched-Tensile (SENT) specimen can be used as complementary test method due to its less boundary effect compared to the SENB specimen. In this study, the SENT specimen which are fabricated with API X 70 carbon steel was simulated ductile fracture behavior by using finite-element-analysis (FEA). To simulate ductile fracture behavior, Gurson-Tvergaard-Needleman (GTN) model was applied. GTN model is a kind of damage model that describes the behavior of ductile fracture through three steps of void changes : nucleation, growth, and coalescence. And GTN model is composed of three constitutive equations and nine parameters. In order to develop the GTN ductile fracture model for API X 70 carbon steel, five kinds of tensile specimen tests were simulated by FEA. In addition, we analyzed the influence of parameters of GTN model through analysis and developed optimal material parameters for API X 70 carbon steel. Finally, the ductile fracture behavior of the SENT specimen was simulated and the FEA results of SENT specimen was compared with actual SENT specimen test.
Comparison of Allowable Bending Moments for Circumferentially Flawed Pipes in ASME Boiler and Pressure Vessel Code Section XI Nonmandatory Appendices C and H
Authors: Kim Hune-Tae , Kim Ji-Su , Seo Jun-Min , Kim Yun-Jae , Lee Kuk-Hee , Huh Nam-Su 
Published: November 2019

Abstract
In this paper, allowable bending moments for circumferential inner surface cracked pipes are evaluated. ASME Boiler and Pressure Vessel code Section XI, “Rules for Inspection of Nuclear Power Plant Components” provides analytical evaluation procedures. Analytical evaluation methods based on the failure mechanism are provided in nonmandatory Appendix C and those based on failure assessment diagram are given in nonmandatory Appendix H. Allowable bending moments are evaluated using both appendices and compared with experiments. Conservativeness is compared quantitatively between both methods by normalizing allowable bending moments with experimental maximum moments.
Advanced Structural Integrity Assessment Tools for Safe Long Term Operation: ATLAS+ Project — Status of the Activities of the WP3 on Modelling
Authors: Marie Stéphane , Blouin Arnaud , Nicak Tomas , Moinereau Dominique , Dahl Anna , Le Delliou Patrick , Bourgeois Myriam 
Published: November 2019

Abstract
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.
ATLAS+ WP3 focuses mainly on ductile tearing prediction for large defect in components: Several approaches have been developed to accurately model the ductile tearing process and to take into account phenomena such as the triaxiality effect, or the ability to predict large tearing in industrial components. These advanced models include local approach coupled models or advanced energetic approaches. Unfortunately, the application of these tools is today rather limited to R&D expertise. However, because of the continuous progress in the performance of the calculation tools and accumulated knowledge, in particular by members of ATLAS+, these models can now be considered as relevant for application in the context of engineering assessments.
WP3 will therefore:
• Illustrate the implementation of these models for industrial applications through the interpretation of large scale mock-ups (with cracks in weld joints for some of them),
• Make recommendations for the implementation of the advanced models in engineering assessments,
• Correct data from the conventional engineering approach by developing a methodology to produce J-Δa curve suitable case by case, based on local approach models,
• Improve the tools, guidance and procedures for undertaking leak-before-break (LBB) assessments of piping components, particularly in relation to representing structural representative fracture toughness J-Resistance curves and the influence of weld residual stresses.
To achieve these goals, WP3 is divided into 4 sub-WPs and this paper presents the progress of the work performed in each sub-WP after 24 months of activities.
Study on Seismic Designs Controlling Locations of Failure Inside Steel Frame Structures Under Severe Ground Motions
Authors: Shiomi Kensuke 
Published: November 2019

Abstract
For steel frame infrastructure facilities like thermal power plants, storage facilities or port facilities, the more advanced seismic performance is needed which not only prevent major damages against assumed design ground motions but also result in the “desirable failure mode” that concerns the recovery works or prevent from resulting in catastrophic failure mode, even under severe ground motions beyond design assumptions in which occurrence of some damages in structures are inevitable.
“Seismic structures which can control the locations of failure of structural members inside structures” is one of the examples of this seismic performance. By adding this performance to steel frame structures at the stage of seismic design, the high resilience structures which concern recovery works after earthquakes can be realized.
In this research, a basic study on the seismic performance which controls the locations of fractures of steel frame members by adjusting the cross sections of each structural member was carried out. The analytical studies about the design procedure to realize this seismic performance were conducted. Then, by conducting the shaking table tests for simple steel frame structures and confirming the location of fractures under dynamic loads, the possibility of this seismic performance was discussed experimentally.
Addressing NRC Concerns Regarding Proposed CC N-830: Direct Use of Fracture Toughness for Flaw Evaluations of Pressure Boundary Materials in Class 1 Ferritic Steel Components
Authors: Erickson Marjorie , Kirk Mark 
Published: November 2019

Abstract
Section XI of the ASME Boiler and Pressure Vessel Code provides KIc and KIa fracture toughness models for ferritic steels. These models are based on linear elastic fracture mechanics methods and were initially developed in the 1970s; they remain largely unchanged since that time. Recently, a modification to Code Case (CC) N-830 has been proposed to provide alternative fracture toughness models for use in the flaw evaluation methodologies of ASME Section XI Nonmandatory Appendices A and K. The integrated models contained in proposed Code Case revision predict the mean trends and scatter of the fracture toughness behavior of ferritic steels throughout the temperature range from the lower shelf to the upper shelf. These models include the transition fracture toughness Master Curve and crack arrest master curve approaches that describe the temperature dependence and scatter in KJc and KIa, respectively in the lower transition temperature region. Also included is a model describing the temperature dependence and scatter of JIc on the upper shelf. Finally, linkage models quantify the inter-relationships between these toughness metrics and how they change due to the irradiation-induced hardening. Together, these models describe the temperature dependence and scatter of fracture toughness initiation and arrest behavior for all ferritic reactor pressure vessel (RPV) steels from lower shelf through transition to the upper shelf, all indexed to a single parameter: T0.
In late 2017 the Electric Power Research Institute (EPRI) published a report, MRP-418, providing the technical basis for these revisions to CC N-830. Nuclear Regulatory Commission (NRC) staff review of the revised Code Case and MRP-418 resulted in substantive questions regarding validation and range of applicability of the various toughness models. An on-going effort addresses these concerns, and a revision to MRP-418 scheduled for publication later in 2019 will summarize that work. This paper describes the efforts of the WGFE CC-N-830 group to respond to the NRC’s comments, and summarizes responses to some of the comments.
Material Properties of Non-Irradiated Zircaloy 4 in Support of ASME Code Acceptance for Pressure Vessel Design
Authors: Nanstad R. K. , Server W. L. , Kombaiah B. , Geringer J. W. 
Published: November 2019

Abstract
A program underway at Oak Ridge National Laboratory (ORNL) to support private industry design of a system for production of 99Mo includes review of Zircaloy-4 (Zry-4) properties as a potential material for a primary vessel in the system. Because Zry-4 is not a material specified in the ASME Boiler and Pressure Vessel Code, a review has been performed to obtain the necessary data in the non-irradiated condition from the literature and from additional testing of Zry-4 to enable creation of tables of material properties required for ASME Code Section III Article ND-3000 design. A strategy for an in-vessel surveillance program is also recommended for early detection of potential corrosion and irradiation degradation issues during operational life of the vessel. Additionally, the ASME Code Section II, Part D was reviewed to ascertain the requirements for proposing a new material for Code approval. A plate of Zry-4 procured by ORNL was used to obtain tensile properties that were used as the basis for determining the maximum allowable stresses applicable to the vessel design. The materials properties of primary interest for ASME Code application are tensile properties, fracture toughness, fatigue, and five specific physical properties required by the Code for approval of a new material. The literature review revealed that all reported experimental data showed tensile strengths to substantially exceed the minimum specifications in ASTM B352. Sufficient fracture toughness and fatigue data are available to conclude that Zry-4 has sufficient fracture and fatigue resistance in the unirradiated condition for vessel design and operation. Of the physical properties reviewed, only thermal conductivity revealed a discrepancy that is recommended for further review. A surveillance program plan has been developed that is probably the maximum that would need to be implemented for the specific vessel design considered. Once adequate non-irradiated and irradiated data are available from preproduction testing of the Zry-4 and weld metal and HAZ materials, the program could be reduced to reflect that either initial properties are very good and/or the effects of embrittlement appear to be minimal. However, the overall design of a surveillance program is, of course, also dependent on the ASME Code design class for the pressure vessel.
Importance of Accurate and Detailed Data Processing of Laser Mapping in Coke Drum
Authors: Rutt Daryl , Park Stephen , Love Darren , Araque Egler , Clark Rick 
Published: November 2019

Abstract
Laser mapping is a well-accepted technique for obtaining surface profiles of coke drum walls to identify bulges. The resulting data is used to track and trend vessel distortions and mapped to illustrate the shape of the vessel from a base radius and or previous inspection. Monitoring the development and evolution of these distortions over time in an accurate and consistent manner has been demonstrated to be an effective tool for predicting bulging induced crack in a coke drum. In this paper, the authors discuss several aspects of the laser mapping technique such as scanner positioning, data noise, laser range accuracy, missing data and the repercussions in the assessment of bulging. The effects on second derivative-based analysis are covered in detail and the use of techniques to reduce the effects of data noise and sensor motion are discussed.
Development of the Containment and Confinement System for Hazardous Material Shock Physics Experiments at Los Alamos National Laboratory
Authors: Spernjak Dusan , Valdiviez Robert , Fehlmann Kevin , Hill Dallas , Gibson Joshem , Bustos Gerald , Tafoya Jose , Yost Nathan , Cardon Devin , Bernardin John , Llobet Megias Anna , McNeil Wendy V. 
Published: November 2019

Abstract
A unique containment and confinement system is under development to conduct small explosively driven physics experiments containing hazardous materials at the Proton Radiography facility at Los Alamos National Laboratory (LANL). In these experiments, the detonation of high explosives (HE) is used to drive materials to extreme loading conditions, where some of the materials tested can be extremely hazardous (e.g. nuclear materials). The main components of the system are the Inner Pressure Confinement Vessel (IPCV, which hosts the physics experiment), the Outer Pressure Containment Vessel (OPCV) and Beam Pipes and Auxiliary Hardware (BPAH).
This paper describes the design and preliminary analyses of the IPCV. The body of the IPCV, also referred to as the Inner Vessel, is being designed to the criteria of the ASME Boiler and Pressure Vessel Code, Section VIII, Division 3, Code Case 2564, with the exception of the materials of construction. The closure covers have different devices mounted on them, such as feedthrough devices for sending or receiving electrical and optical signals across the pressure boundary, and valves for venting the vessel interior. The unique feature in the vessel design are the radiographic windows, tentatively made of Beryllium, which need to be strong enough to maintain the pressure boundary during dynamic events, while being radiographically low-attenuating for the purpose of proton imaging.
Effect of Hydrogen Isotopes on the Fracture Toughness Properties of Types 316L and 304L Stainless Steel Forgings
Authors: Morgan Michael J. 
Published: November 2019

Abstract
Forged stainless steels are commonly used for the containment of hydrogen isotopes and fracture toughness properties are needed for structural integrity assessments. In this study, the effects of hydrogen and tritium precharging on the fracture-toughness properties of Types 316L and 304L stainless steel forgings were measured. The purpose of the study was to evaluate hydrogen and tritium effects on fracture toughness properties of: (1) Type 316 stainless steel stem-shaped and cup shaped forgings; and (2) Type 304L cylindrical block forgings with two different yield strengths. Arc-shaped fracture toughness specimens were cut from the forgings and precharged by exposing the specimens to hydrogen or tritium gas at 623K and 34.5 MPa. Tritium precharged specimens were aged at 193 K for 45 months prior to testing to build-in helium-3 from tritium decay. In the as-received condition, the J-Integral fracture toughness of the stem, cup, and block forgings were very high and exceeded 1200 kJ/m2 on average. The fracture toughness of specimens cut from the low yield strength Type 304L stainless steel block forging had the highest fracture toughness values and Type 316L stainless steel cup forging had the lowest. The reduced fracture toughness values were attributed to the large strain required to produce the cup forging and its high yield strength. Hydrogen precharging reduced the fracture toughness of the stem, cup, and block forgings to values between 34%–51% of a baseline value which was taken to be the fracture toughness value of the low yield strength block forging. Tritium precharging reduced the fracture-toughness values more than hydrogen precharging because of the effects of helium from radioactive decay of tritium. The fracture-toughness properties of tritium-precharged forgings ranged from 12% to 23% of the baseline values. In general, Type 316L stainless steel was more resistant to toughness reductions by hydrogen or tritium (and decay helium) than Type 304L stainless steel. Yield strength had only minor effects on fracture toughness for the precharged steels.
Use of Stainless Steel Tubing in Boiling Applications for Shell and Tube Heat Exchangers
Authors: Radovcich Sarah , Shargay Cathleen , Daru Kuntak 
Published: November 2019

Abstract
The use of stainless steel (SS) tubes in boiling applications must consider the potential risk of chloride stress corrosion cracking (SCC). Most steam generating tube bundles are carbon steel or low alloy steels, but occasionally, higher alloys are needed for the process side corrosion resistance. The use of SS tubes for these cases has had both successes and failures. SS has performed very well in other water and steam services, such as condensing, steam superheating, and boiler feed water (BFW) preheating applications, but for steam generating (i.e. boiling services) the experience has been mixed. Similar failures have also occurred in various process services which are being heated and contain water. The boiling of the water can lead to SCC. Some of the variables that can affect the risk of SCC for SS tube bundles in boiling services include: chloride concentration, tube wall temperature, exchanger design (i.e. kettle, thermosiphons, etc.), vertical vs. horizontal tubes, full vaporization vs. partial vaporization, recirculation rate, and BFW blow down rate. If SS materials are being considered, the risk of SCC can be determined by analyzing these variables as described in this paper. Where the risk of SCC cannot be avoided, an alternate, resistant tube material should be selected. The material options for various services are presented herein.
Computation of Boiling Water Reactor Annulus Flow Loads Using 2D Potential Flow Methodology and 3D Finite Element Analysis: A Comparative Study
Authors: Day Shari , Fong Minji , Walter Matthew 
Published: November 2019

Abstract
During a Loss of Coolant Accident (LOCA) in a Boiling Water Reactor (BWR), subcooled water flows past jet pump assemblies located in the annular region between the Reactor Pressure Vessel (RPV) and the shroud as it moves toward the break location and is subsequently discharged from the RPV. Flow loads caused by such an event are required design basis loads that must be considered for BWR internal components. A comparison is presented between BWR annulus flow loads calculated by the 2D potential flow methodology using complex variables and the 3D finite element method using the heat transfer analogy described in PVP2016-63091 [1]. The comparison demonstrates that greater solution fidelity is available using the FEA method. The results also demonstrate the importance of applying the appropriate velocity correction factor when flow loads are computed using complex variable techniques.
Effect of Bending Load on Burst Pressure of Nuclear Power Plant Steam Generator Tubes With Uniform Wall Thinning
Authors: Liu Michael C. , Gialdini Robert J. , Cipolla Russell C. , Ha Chang-Hoon , Cho Min-Ki , Park Tae-Jung 
Published: November 2019

Abstract
Tube integrity is an important aspect for safe and reliable operation of nuclear power plant steam generators. As a U.S. industry and licensing requirement, all in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions and design basis accidents by meeting the structural integrity performance criterion (SIPC) as given in NEI 97-06. The SIPC margin shall be maintained during plant operation between tube examinations.
The burst strength of tubes subjected to wall thinning will depend on the extent and mode of degradation, and the magnitude of design loads to include pressure differential across the tube wall during normal operation and postulated accident conditions. In addition, non-pressure loads that can occur during postulated accident events shall be evaluated and included in the assessment of tube integrity if determined to significantly reduce the tube burst strength. The EPRI Flaw Handbook provides burst pressure relationships for flaws which include a reduction factor that accounts for the effect of applied bending stress on circumferential degradation. However, this previous industry work was only for planar crack-like flaws and did not directly address uniform volumetric wall loss which can have both axial and circumferential extent.
This paper describes a test program to determine the effect of bending loads on the burst pressure of a tube with uniform thinning over a given axial length. The uniform thinning geometry was selected since it represented a bounding case of general wall loss and is conservative for calculating a tube repair limit for volumetric degradation for a given steam generator design. Tube repair limits are required for defining an upper limit on in-service degradation for which a tube is to be removed from service. Tube repair limits are cited in the Plant Technical Specifications, which is an important part of the licensing basis.