bluuebunny/crossref_metadata_2025_split huggingface dataset:
Estimation of Charpy Index Temperature of SA 508 Mn-Mo-Ni Low Alloy Steels at 41J Using Small Punch Tests
Authors: Kim Jongmin , Hong Seokmin , Lee Taekyoung , Kim Minchul 
Published: October 2020

Abstract
Small Punch (SP) test is one of the miniature test methods, which is a quasi-destructive testing method that evaluates the mechanical properties from limited volumes of materials. Structural materials in nuclear power plants experience degradation in the neutron irradiation environment during the operation, and this aging is assessed by measuring the Charpy index temperature shift of materials at 41J (T41J). However, there are many difficulties in deriving T41J or the fracture toughness from the SP test due to the difference in the failure mode between the SP tests and the conventional Charpy impact and fracture toughness tests. In this study, a method to predict the index temperature at 41J by using the key parameters obtained from SP load-displacement curve was proposed. T41J prediction methods using the fracture energy and displacement after the maximum load in SP test were proposed, and an analysis of the correlation between the key parameters of SP test and the fracture toughness values was performed. As a result of applying displacement after the maximum load parameter of SP curve, it showed better prediction than the fracture energy parameter, and the fracture mode in transition curve of SP was consistent with that of Charpy impact test.
Shakedown and Limit Analysis of Kinematic Hardening Piping Elbows Under Inner Pressure and Bending Moments
Authors: Peng Heng , Liu Yinghua 
Published: October 2020

Abstract
The stress compensation method (SCM) for shakedown and limit analysis was previously proposed and applied to elastic-perfectly plastic (EPP) piping elbows. In this paper, the SCM is extended to account for limited kinematic hardening (KH) material model based on the extended Melan’s static shakedown theorem using a two-surface model defined by two hardening parameters: initial yield strength and ultimate yield strength. To validate the extended SCM, a numerical test on a cylinder pipe is performed. The results agree well with ones from literature. Then the extended SCM is applied to the shakedown and limit analysis of KH piping elbows subjected to inner pressure and cyclic bending moments. Various loading combinations are investigated to create the shakedown limit and plastic limit load interaction curves. The effects of the material hardening, angle of the elbow and loading conditions on the shakedown limit and plastic limit load interaction curves are presented and analysed. The present method is incorporated in the commercial software of Abaqus and can be considered as a general computational tool for shakedown analysis of KH engineering structures. The obtained results provide a useful information for the structural design and integrity assessment of practical piping elbows.
Investigation of Chloride-Induced Stress Corrosion Cracking for AISI 304/304L Stainless Steel Using Notched Bar Specimens
Authors: Jeong Jae-Yoon , Kim Yun-Jae , Lam Poh-Sang , Duncan Andrew , Lee Myeong-Woo 
Published: October 2020

Abstract
Spent nuclear fuels are stored in dry storage canisters made of austenitic stainless steels. Canisters have a sensitivity to chloride-induced stress corrosion cracking since some storage facilities are located in coastal regions, and this environment could have a higher risk for pits and cracks to form on the surface of canisters. Hence, this paper suggests an experimental method of evaluating crack initiation for austenitic stainless steels. Notched bar specimens fabricated using AISI 304 and 304L stainless steels are designed to simulate loads corresponding to welding residual stresses on the surface of canisters. For testing to be conducted in chloride environments, the developed tester for evaluating CISCC is designed such that the notched bar specimens are immersed into the brine of 5% salinity and are loaded by tightening a spring to provide a constant load condition at 50 °C temperature. Also, initial load applied to each notched specimen is verified using finite element analysis as in the same experimental condition. As a result, the areas by pitting corrosion on notches are 1measured using the Image Analyzer program and the stress parameters obtained from the finite element analysis are used to correlate the effect of the pit formation.
The Effect of Thermal Cycling and Packing Stress on Stem Friction and Sealability
Authors: Girão Carlos D. , Veiga José C. , Cristello Josmar B. 
Published: October 2020

Abstract
Control valves are widely used in modern plants in order to keep process variables within a required operating range and ensure both a quality product and efficient operation. Nonlinearities, such as deadband and high static friction (stiction), are the most common cause of oscillations in process variables. Since plants usually consist of various interacting control loops, even small oscillations propagate and disrupt the process parameters out of their operating range, risking accidents, increasing product rejection rates, and decreasing profitability. Packing friction is the primary source of nonlinearities for rotating and sliding-stem valves, and while it is often tested and defined in room temperature, it is common for packings to undergo one or more thermal cycles during their life cycle. A particular packing style may have low stem friction with little to no stiction in room temperature, but present a completely different behavior either in elevated temperatures or after cycling. Failure to ensure correct packing tightness and select proper packing material can disrupt an otherwise well-designed control loop. This study correlates the resulting valve stem friction for different packing styles when subjected to multiple assembly stresses and temperatures. Finally, those same styles undergo standardized fugitive emissions testing, where packing behavior is studied while undergoing thermal and mechanical cycles.
Numerical Simulation of Thermal-Mechanical Stress Field of Coke Drum Considering Deformation
Authors: Yin Jianzhong , Wang Zengchao , Xu Junchen , Han Zhiyuan , Xie Guoshan 
Published: October 2020

Abstract
The coke drum is subjected to a combination of thermal-mechanical loads during operation, so its safety and reliability must be considered in actual production. In each production cycle, the thermal stress generated by temperature fluctuations in the coke drum is the main cause of structural failure. Based on this, a long-term service carbon steel coke drum was used as the research object and a method was proposed to evaluate the universality of the bulged coke drums in this paper. The details are as follows: firstly, the deformed coke drum was processed into a finite element geometric model by internal laser scanning detection. Then, the thermal mechanical coupled finite element analysis was performed on the two regions with large deformation based on the dynamic thermal boundary method. Next, the stresses and strains of the perfect geometry were analyzed and compared. Finally, the stress and strain distribution levels of different bulging patterns are obtained. These results indicate that the maximum stress and the maximum strain of the deformed region were increased by 35% and 77%, respectively, compared with the perfect geometry. Therefore, when evaluating the remaining life of the coke drums, bulges must be taken into account. Furthermore, this method proposed in this paper can provide guidance and reference for the life assessment of coke drums.
Case Study of the Effect of Combined Axial and Lateral Loadings on the Critical Buckling Capacity of Piping Supports
Authors: Wiseman Phillip , Mayes Alex , Karnik Shreeya 
Published: October 2020

Abstract
Snubbers are used in industry to restrain piping in dynamic events which can see significant axial loading as well as lateral acceleration. Snubbers are often employed with an extension when required to bridge gaps between the piping and building structure. As a result, they are susceptible to buckling instability issues. The pipe support and restraint design by analysis buckling criteria for supports given within the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 1, Subsection NF is investigated to determine the behavior of snubber assemblies under combined axial and lateral loadings. Four types of analyses are performed on the assemblies under the action of axial loading to demonstrate finite element and closed form solutions. These include the following: linear Eigen buckling, nonlinear second order large deformation method, energy method and Euler Bernoulli beam theory. In addition, a variety of snubber assembly sizes are subjected to combined axial and lateral loading in the form of multiple magnitudes of lateral acceleration. The behavior was analyzed by the Euler Bernoulli beam theory and nonlinear second order large deformation method. The techniques of each method are compared providing explanations of the assumptions taken, relevant limitations and recommended applications.
Materials Degradation and Strength Reduction Factors of CFRP and Their Applications in Repair of Nuclear Safety Related Piping
Authors: Uddin M. F. , Pothana S. , Krishnaswamy P. , Smith L. , Basavaraju C. 
Published: October 2020

Abstract
Carbon Fiber Reinforced Polymer (CFRP) composites have been used for decades in various industries such as aerospace, oil and gas, mainly due to their high strength-to-weight-ratio and excellent corrosion resistance. However, the use of CFRP in nuclear industry is very limited. Recently, a new ASME Boiler and Pressure Vessel (BPV) Code Case N-871 has been accepted for internal repair of buried Class 2 and 3 nuclear safety related piping using CFRP for Service Levels A, B, C and D. The ASME BPV Code Case N-871 allows the use of both Allowable Stress Design (ASD) and Load and Resistance Factor Design (LRFD) methods to design the CFRP repair system. The USNRC has not yet accepted this code case. In 2016, a relief request was submitted to US Nuclear Regulatory Commission (US-NRC) by Surry Power Station, Virginia Electric and Power Company to perform internal repair of degraded ASME Class 2 and 3 safety related circulating and service water buried piping using LRFD method and CFRP, which was approved by US Nuclear Regulatory Commission (NRC).
It is documented in the literature that CFRP materials experience property degradation when exposed to certain environments and creep behavior under sustained loading. From a limited number of tests, it has also been found that the overall strength of CFRP repair decreases with an increase in the number of plies, however, the extent of the strength reduction is not well-supported with test data. The statistical values of CFRP strength can be determined in three ways. The ASME BPV Code Case N-871 recommends using the characteristic value of CFRP strength at a 5th percentile value with 80% confidence, whereas the A-basis (1st percentile with 95% confidence) and B-basis (10th percentile with 95% confidence) values of strength are used in other industries such as in aerospace applications. For a safety related nuclear application, it is therefore important to evaluate how these values compare to one another. The property degradation due to environmental exposure, creep behavior, multi-ply laminate as well as difference in various statistical analysis can be given a general term as strength reduction factor and These factors will have a direct effect on the safety margin of any CFRP repair using the ASD method and hence, it is essential to confirm these factors with an adequate number of experiments.
In this work, mechanical testing and statistical analyses have been conducted to identify specific degradation mechanisms in CFRP as well as other design considerations that may affect the effective factor of safety of the repair system. These include the effect of multi-ply laminates, misalignment angle, width of the specimen as well as the characteristic values of the ultimate strength. Finally, the strength reduction factors are determined from test results followed by a discussion of how these factors affect the overall effective factor of safety of the repair system.
Review of Life Assessment and Repair Strategies for Hydrogen Reformer Furnace Outlet Header Castings
Authors: Nelson Jeremy , Silfies Trace , Crandall Brian , Penso Jorge 
Published: October 2020

Abstract
Steam methane reforming is the most common method of hydrogen production relevant for plants in the petroleum upgrading, downstream refining, methanol, and ammonia industries. Owner-operators of steam methane reformer furnaces continue to make repair and replacement decisions that involve the cast outlet manifold fittings. One key part of these plans is assessment of the weldability and remaining life of the cast components. The 20Cr-32Ni-1Nb alloy casting materials typically used in the outlet manifolds are usually operated in the low end of their creep temperature range but are subject to metallurgical aging mechanisms which reduce their ductility, weldability, homogeneity, and fracture toughness. This paper covers the practices employed by several owner-users to optimize the lifecycle costs of the outlet manifold castings. These practices include but are not limited to controlled materials specifications, in-situ weldability tests, non-destructive testing in-situ and destructive testing post service, and repair practices such as annealing heat treatments.
This paper also includes a limited survey of several owner-users and their fleets of reformer heaters. The details in the survey include the population of affected cast manifold components, alloy grades for the castings and welds, operating temperature ranges, number of startup and shutdown cycles, ranges of time in service, generic design details, and repair case studies.
Also discussed are recent improvements in the state of the art for high temperature materials property data-gathering, as well as the structural modeling via Finite Element Methods. These new technologies are opportunities for future work to develop better strategies in the areas of condition assessment, repair planning, and remaining life prediction, taking into account the relevant parameters of installed manifold components, including: specific aging behavior of the casting chemistry, component mechanical design details, as well as the welding and heat treatment parameters during initial fabrication and subsequent maintenance activities.
On the Study of Packed Catalyst Bed Stresses for Outward Radial Flow Reactors
Authors: Zhao David , Zhao Mingxin 
Published: October 2020

Abstract
Pressure drop in a radial flow reactor occurs when process flow crosses the packed catalyst bed installed between the two concentric perforated screens during operation. This pressure drop generates the lateral bed stress against the reactor’s perforated screens to shift. The pressure drop will further grow as catalyst attrition increases in production. For an outward radial flow, the pressure drop may exert higher stresses to the outer screen as the packed bed is pushed toward it. An extreme case is when the entire catalyst bed could be pinned to the outer screen of the reactor by enough pressure drop. This could cause the internal components to be overly stressed on the excessive bed load, for which the components might not have been designed adequately. Predicting how radial pressure drop impacts the bed stress and shifts the load distribution is important in preventing mechanical failure during operation. In this study, an analytical model is derived based on Janssen’s theory, a classical semi-empirical granular solid material model, to examine a generic packed catalyst bed in an outward radical flow reactor. A modification to Janssen’s theory is introduced to include pressure drop in order to explore its effects on bed stress and load. The critical condition is derived.
Fabrication and Assembly of the First Accident Tolerant Fuel Concept for TREAT Testing
Authors: Woolum Connor , Imholte D. Devin , Fleming Austin , Kamerman David , Tritthart Korbin 
Published: October 2020

Abstract
Following the tragic events at the Fukushima Daiichi power plant in 2011, priority was given to increasing the accident tolerance of fuel systems for the current fleet of nuclear reactors. These enhanced Accident Tolerant Fuel (ATF) concepts include a wide variety of fuel and cladding materials, both as variants of the current Zircaloy-UO2 system and also as novel fuel and cladding concepts. In addition to testing at steady-state, prototypic, conditions within a nuclear reactor, performance of these ATF concepts in off-normal and transient conditions must be evaluated. The Transient Reactor Test (TREAT) facility at Idaho National Laboratory’s (INLs) Materials and Fuels Complex (MFC) was restarted in the Fall of 2017 and is well-suited to serve this purpose. September of 2018 marked the first fueled specimen to be tested in TREAT since its restart; testing of fuel specimens has been ongoing since then. Initial fuel tests focused on the traditional Zircaloy-UO2 fuel system in order to gain a more thorough understanding of operating characteristics of both the test vehicle system and also the interactions between the reactor and the experiment itself. These tests also served to commission new test vehicles using the well-characterized Zircaloy-UO2 system.
The Separate Effects Test Holder, SETH Capsule, is a modular capsule designed such that it can support a wide variety of specimen geometries ranging from prototypic pressurized water reactor (PWR) fuel samples, heat sink based experiments, and more. The capsule itself is an additively manufactured titanium capsule, within which the experimental specimen is loaded. The SETH Phase I series of tests included five individual SETH capsules, each with a single fuel rodlet and instrumentation to measure temperature during irradiation in TREAT. Each fuel rodlet is representative of a fuel rod in a PWR, with UO2 in Zircaloy-4 cladding.
In August of 2019, TREAT irradiated the first ATF candidate fuel, U3Si2. This marked the first transient test of an ATF concept and is part of a larger campaign that will irradiate a total of four capsules containing ATF concepts. This test campaign, SETH Phase II, built upon the previous SETH Phase I campaign with a nearly identical design except for the fuel rodlet itself. Two of the four SETH capsules contained U3Si2 fuel within Zircaloy-4 cladding, and the other two capsules contained U3Si2 fuel within SiC cladding.
This paper reviews the design, fabrication, and assembly efforts resulting in the four qualified SETH capsules for TREAT irradiation of these ATF concepts.
Case Study on Beam-Mode Flow Induced Vibration Against Random Force Fluctuation due to Turbulent Flow at Bend
Authors: Kataoka Shunji , Matsuura Kota 
Published: October 2020

Abstract
The flow induced vibration is the important phenomena when designing piping system and rigorous analysis method without excessive conservatism is highly demanded. Recently guidelines published by Energy Institute is often applied in design phase of piping system to consider the vibration risk in design; however, the safety factor for the method is not fully investigated, especially, for example, the effect of the piping layout, flexibility factor of bend, extrapolation to large bore headers, etc ....
In previous paper, Authors investigated the forcing function due to turbulence flow at the bend by using LES-CFD analysis and quantified the random force fluctuation at the 90 degrees miter bend and two smooth bends and showed that random vibration analysis with these forcing functions can be used to assess the risk of piping vibration.
In this paper, the authors conducted a series of random vibration analysis on the beam mode vibration of typical piping systems. The design variable, such as piping diameter, diameter to thickness ratio, number of flow direction change in each span, and support span were varied as a parameter of the design and evaluated the magnitude of stress and vibration level. The results were compared with the LOF score provided by Energy Institute Guidelines and clarified the difference in design margin of the methods. In conclusions, provides some guidance to reasonably apply the screening method in the design phase of the piping system.
Impact of the Outer Surface Air Cooling and WPS Approaches on the Brittle Fracture Margin of WWER RPV
Authors: Zarazovskii Maksym , Dubyk Yaroslav , Filonov Vladyslav , Antonchenko Vitalii , Ishchenko Oleksii 
Published: October 2020

Abstract
This paper presents a results of brittle fracture (BF) assessment of WWER-1000 reactor pressure vessel (RPV) subjected to pressurized thermal shook (PTS).
Work is focused on the two factors affected on the BF safety margin. The first one is the definition of the heat transfer conditions from outer surface of RPV, which is commonly not considered due to the assumption of conservatism. The second one is the warm pre-stressing (WPS) of RPV metal during PTS.
The procedure for the formation of thermal boundary conditions (BC) for the RPV (i.e. lower nozzle forging, cylindrical part and elliptical bottom), reactor supports, as well as for the reactor thermal insulation, is developed. This procedure takes into account the heat transfer due to radiation in the GA-301 room and the presence of forced supply of “cold” air by the ventilation system.
A brief history of WPS implementation in Ukraine is presented. Also, the comparative review of the WPS experience in PWR countries and WPS approaches is made.
The results of fracture mechanics calculations for the three most severe PTS scenarios of the Unit 3 RPV of South-Ukrainian NPP are presented. Wherein, both type of results is demonstrated: with and without consideration of the RPV outer surface heat transfer due to the air-cooling ventilation system.
It is shown that the RPV air cooling consideration: leads to decreasing of the stress intensity factor; provides to satisfaction of the cladding integrity criterion according to VERLIFE-2008 methodology; can justify lifetime of the RPV nozzle region in terms of not to exceed the upper shelf of fracture toughness 200 MPa·m0.5.
At the same time, consideration of the RPV air cooling alongside with the WPS approach, depending on WPS approach considered, can lead as to the decreasing the conservatism of RPV BF assessment, as to the increasing it.
Also, concluded that Ukrainian WPS approach needs to be refined more carefully with consideration of experimental nature of pre-stressing or harmonized with some of the modern ones.
Recent Argonne National Laboratory Estimated Thermal Expansion Coefficients for Reactor Pressure Boundary Base Metal, Similar and Dissimilar Metal Welds
Authors: Mohanty Subhasish , Listwan Joseph 
Published: October 2020

Abstract
The paper presents estimated thermal expansion coefficients of 316SS and 508LAS base, 316SS filler or similar metal weld (SMW), and 316SS-508LAS dissimilar metal weld (DMW) filler and butter weld metals. These base and weld metals are typically used in nuclear reactor pressure boundary components. Accurate estimation of the expansion coefficients of these metals is essential for accurate estimation of thermal-mechanical stress in reactor pressure boundary components. In this paper we present the expansion coefficients of 316SS and 508LAS base, 316SS-SMW filler, and 316SS-508LAS DMW filler and butter weld metals. The coefficients were estimated based on our own experimental data. The corresponding expansion coefficient results and the FE validation results are presented in this paper. We anticipate that these types of results can be used as guidelines for choosing appropriate expansion coefficients for thermal-mechanical stress analysis of safety critical nuclear reactor components.