bluuebunny/crossref_metadata_2025_split huggingface dataset:
Experimental Investigation on the Seismic Performance of a Multi-Component System for Major-Hazard Industrial Facilities
Authors: Butenweg Christoph , Bursi Oreste S. , Nardin Chiara , Lanese Igor , Pavese Alberto , Marinković Marko , Paolacci Fabrizio , Quinci Gianluca 
Published: October 2021

Abstract
Past earthquakes demonstrated the high vulnerability of industrial facilities equipped with complex process technologies leading to serious damage of the process equipment and multiple and simultaneous release of hazardous substances in industrial facilities. Nevertheless, the design of industrial plants is inadequately described in recent codes and guidelines, as they do not consider the dynamic interaction between the structure and the installations and thus the effect of seismic response of the installations on the response of the structure and vice versa. The current code-based approach for the seismic design of industrial facilities is considered not enough for ensure proper safety conditions against exceptional event entailing loss of content and related consequences. Accordingly, SPIF project (Seismic Performance of Multi-Component Systems in Special Risk Industrial Facilities) was proposed within the framework of the European H2020 - SERA funding scheme (Seismology and Earthquake Engineering Research Infrastructure Alliance for Europe). The objective of the SPIF project is the investigation of the seismic behaviour of a representative industrial structure equipped with complex process technology by means of shaking table tests. The test structure is a three-story moment resisting steel frame with vertical and horizontal vessels and cabinets, arranged on the three levels and connected by pipes. The dynamic behaviour of the test structure and of its relative several installations is investigated. Furthermore, both process components and primary structure interactions are considered and analyzed. Several PGA-scaled artificial ground motions are applied to study the seismic response at different levels. After each test, dynamic identification measurements are carried out to characterize the system condition. The contribution presents the experimental setup of the investigated structure and installations, selected measurement data and describes the obtained damage. Furthermore, important findings for the definition of performance limits, the effectiveness of floor response spectra in industrial facilities will be presented and discussed.
Study on Reliability of Seismic Capacity Analysis for Important Equipment in Nuclear Facilities
Authors: Muta Hitoshi , Ohtori Yasuki , Furuya Osamu , Fujita Satoshi 
Published: October 2021

Abstract
Since the severe accident of Fukushima Dai-ichi Nuclear Power Plants, the enhancement and improvement of the safety regulations of nuclear facilities in Japan from various perspectives have been considered the new regulations are based on defense in depth concept as well as before Fukushima Accident, and have significantly raised design basis assumptions and required countermeasures for natural phenomena to prevent the simultaneous loss of safety functions due to the dependency.
From the above background, according to raised design basis assumptions, Severe Accident (SA) countermeasures against beyond design basis external hazards including earthquakes, tsunami and so on have been considered and the plans of installation of them have been accepted by Nuclear Regulatory Authority. On the other hand, since Probabilistic Risk Assessment (PRA) and Safety Margin Analysis (SMA) are performed in the safety improvement evaluation to ensure the residual risk of severe accident and confirm the actual performance of safety functions. However, since less information could be obtained from the experienced capacity data or operation experiences, the safety insights from PRA and SMA have not been clarified so far.
This paper presents the research and examination results of the reliability of capacity data based on the very few experience data by using Statistical Technique and Bayesian Update Technique to enhance the reliability and validity of PRA and SMA and summarizes kinds of insights obtained by this research and examination.
A Simplified Method for Analyzing Pressure Vessel Response to An External Blast Overpressure Generated From an Explosion
Authors: Hundl Jacob , Millet Barry , Mosher Bryan , Kirkpatrick Kenneth 
Published: October 2021

Abstract
The widespread increase in the construction of liquefied natural gas (LNG) processing facilities and gas producing offshore platforms has produced new requirements in designing equipment, in this case pressure vessels, to withstand a blast overpressure condition resulting from the failure of an LNG train or some other form of gas-based explosion. New client requirements are aimed at determining the viability of equipment after one of these catastrophic events. Traditional methods of analyzing pressure vessel response to blast overpressure, such as finite element analysis (FEA) and computational fluid dynamic (CFD) studies, are costly and impractical when applied to large scale projects and tens or hundreds of pieces of equipment. To adequately account for and measure pressure vessel response to blast overpressure, a simplified analytical method is required. However, a simplified method that does not consider the unique conditions presented by an overpressure event produce extremely conservative results (e.g. applying the overpressure as a wind load or pressure times area). Unfortunately, there is limited methodology and literature on blast overpressure phenomena, particularly related to petrochemical refining processes.
This paper will examine the development of a simplified method for analyzing pressure vessel response to blast overpressure using methods originally intended for the study of the effects of nuclear weapons on structures as well as more traditional methodology which employs an explicit solution utilizing modern, commercially available FEA software. The paper will review the methods presented in the existing literature and examine how those methods can be efficiently and practically applied to pressure vessels. The results generated by the simplified methodology will be compared to results from a control case that uses an explicit FEA method to evaluate pressure vessel response to blast overpressure.
Statistically Analyzing the Characteristics of Gas Pipelines Accidents Based on PHMSA Database
Authors: Liu Shengli 
Published: October 2021

Abstract
Deeply analyzing the temporal characteristics of gas pipeline accidents is crucial to the safety management of gas pipeline. In this study, gas distribution pipeline accidents recorded from 1986 to 2020 provided by PHMSA have been chosen to explore the characteristics of gas pipeline failure frequency in the United States. The result indicates that the failure frequency of gas pipelines has been reduced with the improvement of safety management. The analysis results show that the gas pipeline accidents occurred in the U.S. from 1986 to 2020 are temporal connected. This is due to the fact that the leading failure causes are related with human factors, which have strong temporal correlations. Improving the pipeline safety management can reduce the temporal dependence. However, there are essential temporal pattern differences among gas pipeline accident sequences of different severity. Accurate prediction of the gas pipeline accident occurrence can improve the quantitative risk assessment of gas pipeline accidents, and also help the pipeline management personnel to develop an efficient emergency response plan. Therefore, as for the quantitative risk assessment of gas pipeline accidents, it is necessary to carry out in-depth analysis of the occurrence probability according to the severity of the accidents.
A Numerical Approach to Correlate Compression Stress Relaxation and Compression Set of Elastomer O-Rings With Tightness
Authors: Weber Mike , Zaghdoudi Maha , Kömmling Anja , Jaunich Matthias , Wolff Dietmar 
Published: October 2021

Abstract
The excellent mechanical properties of elastomer seals at a wide range of temperatures as well as their high versatility and recovery potential under several load conditions make these materials well suitable for the application in containers designed for transport and disposal of negligible heat generating radioactive waste. While a seal exchange at defined intervals is typical in many conventional applications, it is impossible or at least hard to perform when principles of minimization of radiation exposure have to be considered which prohibit an avoidable cask handling. An extensive knowledge of the change of the elastomer properties during aging and the availability of reliable end-of-lifetime criteria to guarantee the safe enclosure of the radioactive material for the required time are mandatory. As BAM is involved in most of the national cask licensing procedures and in the evaluation of cask-related long-term safety issues, great efforts have already been made and are still ongoing to scientifically support this task.
Among other representative types of elastomers, specimen made from ethylene propylene diene rubber (EPDM) were tested before, during and after aging to capture the with respect to application most important of their complex mechanical properties.
Exemplary results of these investigations were used to calibrate material models implemented in the commercial finite element software ABAQUS/Standard®. The finite element model already presented in previous works uses a sequential temperature displacement coupling. The calculated compression stress relaxation (CSR) and compression set (CS) values do satisfactorily match the experimental results.
In many investigations performed at BAM both values (CSR and CS) were identified as key indicators of elastomer’s long-term performance.
However, the possibility to correlate these equivalent indicators with performance values such as tightness and leakage rate, measurable in the mounted state, is an important goal of our future work. In the presented study the ABAQUS® feature of “pressure penetration” is introduced in the suggested finite element model for this purpose. It provides the possibility to simulate the penetration of a gas into a possible gap between flange and O-ring causing an opening of a leakage path. Three dimensional and axis-symmetric finite element models were generated to represent flat and grooved flanges of different dimensions.
The sensitivity of the feature to several input parameters is investigated and the observed behavior of the O-ring is correlated with the results of performed leakage tests.
Applicability of Design Rules for Openings in Shells, ASME B&PV Code, Section VIII, Division 2 - A Case Study
Authors: Kagita Gurumurthy , Pudipeddi Krishnakant V. , Sripada Subramanyam V. R. 
Published: October 2021

Abstract
The Pressure-Area method is recently introduced in the ASME Boiler and Pressure Vessel (B&PV) Code, Section VIII, Division 2 to reduce the excessive conservatism of the traditional area-replacement method. The Pressure-Area method is based on ensuring that the resistive internal force provided by the material is greater than or equal to the reactive load from the applied internal pressure. A comparative study is undertaken to study the applicability of design rules for certain nozzles in shells using finite element analysis (FEA). From the results of linear elastic FEA, it is found that in some cases the local stresses at the nozzle to shell junctions exceed the allowable stress limits even though the code requirements of Pressure-Area method are met. It is also found that there is reduction in local stresses when the requirement of nozzle to shell thickness ratio is maintained as per EN 13445 Part 3. The study also suggests that the reinforcement of nozzles satisfy the requirements of elastic-plastic stress analysis procedures even though it fails to satisfy the requirements of elastic stress analysis procedures. However, the reinforcement should be chosen judiciously to reduce the local stresses at the nozzle to shell junction and to satisfy other governing failure modes such as fatigue.
Comparative Research on Nonlinear Analysis Method Technique for the Failure Modes Based on RCC-M Annex ZC
Authors: Shao Xuejiao , Xie Hai , Xiong Furui , Fu Xiaolong , Tian Jun , Zhang Ying , Shi Kaikai , Yu Mingda 
Published: October 2021

Abstract
In the evaluation of nuclear equipment, if the elastic analysis cannot meet the corresponding design criteria, the nonlinear analysis should be adopted to assess the safety of equipment. A new nonlinear calculations appendix is supplied in RCC-M 2017 edition annex ZC to secure the nonlinear F.E. calculation performance.
In this paper, the failure modes of nuclear engineering structure are discussed. In order to ensure the safe operation of the structure, the corresponding evaluation criteria formulated by RCC-M code are proposed. Based on the annex ZC of RCC-M 2017 edition, three kinds of nonlinear analysis methods, including limit analysis, plastic analysis and cycle analysis, are studied. The austenitic stainless steel typical nozzle model is used to carry out the finite element calculation. The effect on the limit load is compared between the limit load method of perfect elastoplastic analysis and the secant method of the elastoplastic analysis. The influence of the strain range method and the modified Ke method on the fatigue calculation results is also discussed. The strain range can be determined by monotonic calculation or cyclic calculation. The two kinds of approach are implemented in the elastic plastic fatigue calculation. Compared with the simplified elastic plastic fatigue results, using Ke method to calculate elastic-plastic fatigue can reduce the conservatism of the results.
Fatigue Crack Growth Rate Data Assessment
Authors: Wang Enyang D. , Tiku Sanjay , Dinovitzer Aaron 
Published: October 2021

Abstract
A comprehensive fatigue test database was developed in PRCI SIA-1-1 and IM-3-2 projects, where there are 185 sets of fatigue crack growth rate parameters (i.e., C and m fitting coefficients in the Paris law) calculated from compact tension (CT) fatigue test data. Four sets of recommended values of C and m were presented for two flaw locations (base metal (BM) and Electric Resistance Weld seam (ERW)) and two R ratios (R &lt; 0.5 and R ≥ 0.5). The past works did not further assess the fatigue crack growth rate parameter relation (i.e., C and m), although intuitively, two parameters should be correlated for a material type (i.e., line pipe). To this end, a systematic study was carried out to reassess the fatigue crack growth rate database. It is found that parameters C and m are strongly correlated in the form of m = A * log(C) + B, where A and B are fitting coefficients. In addition, C and m (mresidual) values follow lognormal and normal distributions, respectively, where mresidual refers to the residuals associated with fitting the predicted values. The values of A and B for BM or ERW at R = 0.1 or 0.6 (four scenarios) are reported. Note that the distributions of C and m for ERW cases have relatively heavier tails and shorter peaks.
Based upon the four newly developed sets of A and B values, a series of probability studies were carried to calculate the most-likely fatigue crack growth life in the integrity analysis. Note that two ERW pipe segments reported in PRCI IM-3-2 projects were considered in the full-scale test, where fifty-one (51) initial notches were prepared to evaluate the fatigue crack growth rate. The initial notches were cycled to develop different crack lengths, depths, materials, and flaw locations (e.g., BM and ERW seam). The fatigue crack growth lives calculated from the PRCI approach were used as the baseline for comparisons. In total, 30,000 pairs of C* and m* data points were generated from random sampling for each scenario, and there are over 1.5 million random simulations carried out. The probability densities for each full-scale test case were simulated using a Monte-Carlo approach. The fatigue crack growth life distributions for BM are relatively narrow because of the fatigue crack growth rate parameters’ lower standard deviation. The most-likely (median) fatigue life from the probabilistic study is almost identical to the fatigue life from deterministic calculations for the BM cracks. The probabilistic simulation results of ERW seam crackfatigue lives are significantly different from those for BM cracks.
Inservice Inspection of Extended Dry Storage of Spent Nuclear Fuel, Part II: NDE/Sensor Technology Development and Codification
Authors: Meyer Ryan M. , Renshaw Jeremy , Hunter Kenn , Orihuela Mike , Stadler Jim , Mauskar Karan , Wolf Uwe , Klein Marvin , Weidmann Max , Bobbs Bradley , Dunn Darrell , Lin Bruce 
Published: October 2021

Abstract
This paper describes development and demonstration of nondestructive examination (NDE) technologies to support periodic examinations of interim dry storage system (DSS) canisters for spent nuclear fuel in the USA to verify continued safe operation and that the canister confinement is intact and performing its intended safety function. Specifically, this work relates to NDE technology development for “canister” based DSS systems, which form the majority population of DSSs in the USA for interim storage of spent nuclear fuel. Consideration of potential degradation of the welded stainless-steel canister in these systems is required for continued usage in the period of extended operation (PEO) beyond the initial license or certified term. Physical access to the canister surface is constrained due to narrow annulus spaces between the canister and the overpack, tortuous entry pathways, and high temperatures and radiation doses that can be damaging to materials and electronics related to inspections. Several activities to demonstrate NDE technologies for the inspections of different DSS systems are summarized.
Case Study to Evaluate Stresses in Welded Endplates in Structural Steel
Authors: Gawande Kshitij P. , Mayes Alex , Subedi Raju 
Published: October 2021

Abstract
Endplates are widely used in the industry to attach supplementary steel structures to main building frames. These endplates can be attached to the building steel using a bolted connection or a welded connection. Industry often favors bolted connections due to ease of installation and availability of qualification methods per AISC 360 Design Guides. However, there are some applications where a welded connection is preferable, such as, cases requiring reduction of number of parts supplied or applications with higher chance of vibration causing loosening of bolts. The present case study discusses evaluation of stresses in welded endplates due to forces and moments from the attaching supplementary steel members. The study considers various welded connection scenarios including an endplate welded on two opposite sides and an endplate welded on all four sides. The stress distribution in the plate is studied using finite element analysis with wide flange and tube steel members attaching to it. ANSYS mechanical is used to perform the finite element analysis. Multiple combinations of plate sizes, weld patterns, and attaching member sizes are analyzed to provide a well-rounded solution. An analytical model is developed for the stress evaluation as well and the results are compared with the finite element model. The study is intended to provide an efficient methodology for plate evaluation and qualification.
Alternative Nozzle Reinforcement Rules for Gasketed Plate Heat Exchangers
Authors: Nikic Milan , Srnic Djordje 
Published: October 2021

Abstract
This paper aims to provide an alternative design approach to the current nozzle reinforcement rules for gasketed plate heat exchangers (GPHE) designed in accordance with Appendix 45 of ASME BPVC Section VIII, Div. 1. The current ASME BPVC Section VIII, Div. 1 nozzle reinforcement rules require the use of bending stress calculated at the center of the GPHE end plate where the bending stress has the highest value. Typical GPHE’s have inlet and outlet nozzles located near the top or bottom of the GPHE end plates where the bending stresses are lower as compared to the bending stresses at the center of the plates. For nozzle reinforcement analysis in GPHE end plates, this alternative design approach suggests utilizing the bending stress values calculated in vicinity of nozzles in lieu of bending stress value at center of the end plate. It also suggests utilizing the modified UG-39 rules for GPHE end plate nozzle reinforcement. In order to optimize the design thickness of the GPHE end plates, this paper will analyze the impact of nozzle size and location on bending stresses in the end plates, propose modification of UG-39 rules for GPHE end plates, and to define the limits of the design approach.
Coupling of Neutron Noise and Dynamic Finite Element Analyses to Perform Remote Condition Monitoring for Reactor Vessel Internals
Authors: Banyay Gregory A. , Palamara Matthew J. , Smith Stephen D. 
Published: October 2021

Abstract
During a Spring 2019 refueling outage, Plant A, a pressurized water reactor (PWR), experienced difficulty removing the reactor vessel upper internals because of degraded thermal shield (TS) support block bolts (TSSBBs), TS flexures, and baffle-former bolts (BFBs). None of the degradation was expected based on prior industry operating experience. Dynamic finite element analyses (FEA) and neutron noise monitoring were performed to diagnose the degraded condition. The FEA simulated the indications from the 2019 outage to assess the effect that degradation would have on the dynamic response of the lower reactor vessel internals. In parallel, neutron noise data was collected and evaluated to characterize the vibratory signature of the Plant A lower internals. The results of the FEA and neutron noise monitoring were coupled to compare degradation scenarios with the measured data, and it was thus predicted that multiple non-functional (fully failed) flexures existed at Plant A. Those predictions were subsequently validated during the Fall 2020 outage.
This use of neutron noise to detect and diagnose degradation represents the practice of proactive structural health monitoring (SHM) for reactor vessel internals. This remote condition monitoring and diagnostic computational framework is currently being enhanced by quantifying the sensitivity of the structural dynamics to different degradation scenarios. This methodology leverages benchmarked FEA models and machine learning methods to enhance interpretability of neutron noise results and proactively diagnose different degradation scenarios, thus enabling prognostic asset management for reactor structures.