bluuebunny/crossref_metadata_2025_split huggingface dataset:
Non-Metallic Sealing Material for Elevated Temperature Applications
Authors: Damdar Sherwin , Drago James , Evans Wayne 
Published: December 2017

Higher and higher process temperatures in the power, hydrocarbon and chemical process industries are employed to quicken production and synthesize new materials. Ultra-high temperatures range from 850°F (454°C) to 1800°F (982°C). The choice of ultra-high temperature non-metallic sealing materials has been limited. Traditionally sealing materials for these extremes have been made with ceramics (e.g., aluminum silicate based) and mica. Recent developments utilized vermiculite and most recently talc. This paper will present the latest development of talc based material. Comparative physical properties and well as functional attributes of creep and sealing ability will be presented. Experience with the talc based material in solid oxide fuel cells, exhaust and gasification systems are highlighted.
Paper published with permission.
Experience in the Application of the Omega Creep Model in Creep Experiments and Component Analysis
Authors: Marriott Douglas L. , Sreeranganathan Arun , Carter Peter , Read Shannon 
Published: December 2017

Since the publication of ASME FFS-1/API 579 in 2007, the Omega creep model has become a virtual de facto standard for reduction of creep data and for high temperature structural analysis. Experience in both experimental and analytical applications has revealed some important features of this model. This paper presents a number of case studies illustrating these features, and discussing the range and limitations of the scope of the model, and procedures built around it. The paper includes recommendation, based on this experience, on the place considered appropriate for this model in future FFS and RLA applications.
Paper published with permission.
Serviceability Assessment of Creep Strength Enhanced Ferritic Steels
Authors: Arnold John L. 
Published: December 2017

The metallurgical advancements exhibited in creep strength enhanced ferritic steels have changed the dynamics of boiler design and operation by improving the fatigue life of thick-walled components and improving cycle efficiency by increasing operating temperatures. These advancements have a similar affect on the assessment of these materials. No longer can a serviceability evaluation be based on a general assumption of metallurgical properties. Instead, condition assessment protocols, examination methods, and flaw assessment approaches must reflect the metallurgical nuances of these high temperature alloy steels. This paper will provide guidance on new issues to be considered when developing a serviceability program.
Paper published with permission.
Constitutive Modeling of Haynes 230 for High Temperature Fatigue-Creep Interactions
Authors: Barrett Paul R. , Ahmed Raasheduddin , Hassan Tasnim 
Published: December 2017

Non-linear stress analysis for high temperature cyclic viscoplasticity is increasingly becoming an important modeling framework for many industries. Simplified analyses are found to be insufficient in accurately predicting the life of components; such as a gas turbine engine of an airplane or the intermediate-heat exchanger of a nuclear power plant. As a result, advanced material models for simulating nonlinear responses at room to high temperature are developed and experimentally validated against a broad set of low-cycle fatigue responses; such as creep, fatigue, and their interactions under uniaxial stress states. . This study will evaluate a unified viscoplastic model based on nonlinear kinematic hardening (Chaboche type) with several added features of strain-range-dependence, rate-dependence, temperature-dependence, static recovery, and mean-stress-evolution for Haynes 230database. Simulation-based model development for isothermal creep-fatigue responses are all critically evaluated for the developed model. The robustness of the constitutive model is demonstrated and weaknesses of the model to accurately predict low-cycle fatigue responses are identified.
Paper published with permission.
Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors
Authors: Asayama Tai , Nagae Yugi , Wakai Takashi , Tsukimori Kazuyuki , Morishita Masaki 
Published: December 2017

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs.
Paper published with permission.
Comparison of the Requirements of the British R5, French RCC-MRx, Proposed New Rules in ASME Section VIII, and API 579 Codes in the Design of a Cylinder Subjected to Pressure With Thermal Gradient at Elevated Creep Temperatures
Authors: Anderson David , Chithranjan Nadarajah , Jawad Maan , Martin Antoine 
Published: December 2017

The authors analyze two sample problems using four different international codes in the evaluation. The first is the British R5 code, the second is the French RCC-MRx code, third is the ASME Section VIII, Division 2, code using proposed new simplified rules taken from the ASME nuclear code section NH, and the fourth is the API 579 code. The requirements, assumptions, and limitations of each of the four codes as they pertain to the sample problems are presented.
The first sample problem is for creep-fatigue analysis of a cylindrical shell subjected to internal pressure with a linear thermal gradient through the wall. The second sample problem is evaluating the critical buckling strength of the cylindrical shell under external pressure in accordance with proposed new rules in ASME Section VIII, Division 2, API 579, and a finite element analysis.
Paper published with permission.
The Latest Progress of Containment Design in China
Authors: Shang Ziduan , Huang Xiao , Yang Yong , Chu Meng , Li Shaoping 
Published: December 2017

Since first Nuclear Power Plant (NPP) in operation in early 1990s, China has built dozen of Light Water Reactor (LWR). Originated from Pressure Water Reactor (PWR) technology, almost all these plants, such as Qinshan, Dayawan, and Tianwan, were designed with a Pre-stressed Concrete Containment Vessel (PCCV). With the application of Third Generation (GEN III) reactor technology in recent years China is designing and building more Gen III NPPs. As the first kind of passive safety GEN III + technology, AP1000 uses double-layered containment structures comprising steel containment and shield building. GEN III EPR use similar double-layered containment structures, the outer layer is reinforced concrete shield structure, the inner layer is PCCV. In the traditional containment design, the design load and load combinations are determined based on different NPP operation conditions. This means the basic design principles of defense-in-depth and single failure criteria are incorporated into containment structure design through the consideration of all operation-related design basis loading conditions. However The Indian Ocean tsunami and the Fukushima earthquake and tsunami followed by the Fukushima Daiichi accident illustrated the hidden potential vulnerability in current NPP design when subjecting to Beyond Design Basis Event (BDBE) loading conditions. Containment, as the final physical shielding barrier to protect plant from potential radioactive releasing in accidental events, shall be designed to be not only functional during design basis accident conditions, but also be able to perform its containment and shielding functions, maintain the integrity of pressure boundary during BDBE and their loading conditions.
Aiming on containment and shield structure design for GEN III NPPs, this paper summarized preliminary research results on BDBE design. Considering the occurrence frequency and impact severity to containment structure, Beyond Design Basis Seismic (BDBS) is taken as the investigation focus in this paper. The summary is based on an investigation of the BDBS loadings and their impact to safety-related nuclear containment and shielding structures of the NPPs. The purpose of this investigation is to understand the failure pattern and modes under BDBS loads; and how such failure impact the safety features of the plant and the safety functions of the systems (e.g. the impact to passive safety functions). This paper also covers ways to reduce the impact and mitigate the consequences of BDBS impact by pre-incorporated measures in the structural design.
The Electro Magnetic Compatibility Tests for the AMB System on HTR-10GT
Authors: Yan Zhou , Xunshi Yan , Ni Mo , Guojun Yang , Zhengang Shi 
Published: December 2017

With more and more widely used in the industrial field, Active Magnetic Bearings (AMB) have bigger challenge, such as the size and the load capacity are larger than before. In order to increase the reliability of the whole system, a lot of aspects have to be improved to adapt some special industrial requirements. Just like the 10MW high temperature gas-cooled test module reactor (HTR-10GT) with the core made of spherical fuel elements, which was designed and constructed by the Institute of Nuclear and New Energy Technology of Tsinghua University in China, take the AMB as the support bearings. HTR-10GT is the reactor with 3.5 meter long rotor, 230mm diameter, and 15000rpm Rated speed. Since there are some difficulties in HTR-10GT including the blade stages, small gap and so on, the high precision control on magnetic bearing is required very much. Once any part in the system is out of work, the rotor would happen to vibrate, which will lead to turbine blades scraping. The consequences could be disastrous especially when the system is in full power at the rated speed [1].
As you know, the electronic control system is the weakest point in AMB, both in operating site and the control room. Therefore, before engineering applications, the electronic control system must be tested for Electromagnetic Compatibility (EMC). Based on the experimental results, some design would be optimized, to improve anti-interference capability to meet the requirements of HTR-10GT. This paper describes the testing process and gives some solutions for EMC problems during the Electro Magnetic Susceptibility (EMS) test, such as the grounding, cable, wiring, filtering etc. These methods are the fundamental basis for reducing interference and improving the reliability of AMB in the HTR-10GT.
Shaking Table Test for Base-Isolated Nuclear Island Structure Model
Authors: Jie Yang , ShaoPing Li , Fang Yuan , ZuFeng Xia , XiaoLin Huang 
Published: December 2017

In this paper, the base-isolated design of Nuclear Island structure will be introduced, including the general requirement and the goal of the base-isolated design. Integrated assessment has been performed for the base-isolated design of Nuclear Island structure in the earthquake 0.6g. A series of nonlinear time-history analysis were performed to predict the maximum displacement and acceleration of the isolation layer, the maximum stress of the isolation units, and the floor response spectra of each story of the superstructure in the earthquake 0.6g, considering the realistic mechanical properties and the layout of the isolators.
In order to provide realistic data to validate the numerical method, a reduced-scale earthquake simulation of base-isolated nuclear structure on a shaking table was carried out. The study was primarily focused on the response of superstructure and the isolation unit. The dynamic characteristic was examined, together with the vibration acceleration and displacement under different levels of seismic wave. The test results of a reduced-scale nuclear island model previously tested on a shaking table were compared with three-dimensional finite element simulation results.
The results of this study provide the technical basis for the base-isolated design of Nuclear Island structure.
Study on Load Capacity of Main Steam Superpipe Nozzle Used in Nuclear Power Plant
Authors: Gengyu Zhou , Shuhua Liang , Lin Sun , Feng Lv 
Published: December 2017

The main steam super pipe used in nuclear power plant is an important safety class2 component. There are several nozzles located on it and linked with main steam safety valves. In the past two decades, the hot extrusion forming technology has been widely used to manufacture the super pipe nozzles. Comparing with traditional insert weldolet, the wall thickness of the extruded nozzle is relative small, and the nozzle inner radius is hard to control precisely in the fabrication process. Due to high temperature working condition and complicated loading conditions, the load capacity of the super pipe extruded nozzle has become an issue of concern for manufacturers and users.
This paper presents a structural integrity assessment of a super pipe extruded nozzle. The nozzle stresses due to internal pressure and external loads for different operating conditions are obtained by the three-dimensional finite element analysis. The extruded nozzle is evaluated against the RCCM code Subsection C3200 Service Levels O, B and D stress limits for design, upset and faulted conditions. A parametric sensitivity analysis of the extruded nozzle inner radius size is also carried out. In addition, in order to reduce the calculation effort, an efficient calculation method is developed by using the commercial finite element program ANSYS.
Fluid-Structure Interaction Numerical Analysis of the Spent Fuel Pool and Storage Racks Under Earthquake
Authors: Yun Ling , Lei Li , Rongjun Xue , Hao Qian , Honghui Ge , Ziduan Shang 
Published: December 2017

Spent fuel pool and storage racks are important nuclear security structures and components. In order to prevent it from structural failure, which includes the loss of the structural integrity of the spent fuel pool and stability of the spent fuel storage racks, also includes the possibility of fallen down of storage racks under seismic loading. Besides the necessary static analysis of structures, the influence of seismic loading on the interaction between water and structure should be fully considered, Especially concerned the analysis of the shaking effect of water sloshing on the storage racks, the displacement and the possibility of fallen down of the storage racks.
The present paper is concerned with the problem of modeling the fluid-structure interaction (FSI) in filled liquid and filled with spent fuel pool. The study focuses on the sloshing phenomena and on the coupling computational fluid dynamic (CFD) analysis with the finite element stress analysis (FEA) code LS-DYNA. By the results of the response of seismic, such as the displacement of the storage racks, pressure exerted on the plate of racks and the walls of the pool. This paper also evaluates the seismic performance of the structure and the safety margin.
Various numerical methods can be used for analysis of liquid storage pools, among these we mention explicit finite element, implicit Lagrangian-Eulerian, hybrid finite element, Smoothed Particle Hydrodynamics volume of fluid. In this article the coupled sloshing dynamics in a rectangle pool were studied using a model developed in LS-DYNA environment. The main solution methodology is based on explicit time integration. In order to demonstrate the FSI results of the FEA models of the spent fuel pool on seismic analysis, a 3D FEA models were developed. The Finite element model composed of the spent fuel pool (steel plate concrete), spent fuel storage racks, cushion block, water and air. Solid element modeling is used in concrete, cushion block, water and air. Steel plate and storage racks employ the shell element. The constitutive model of solid element is linear elastic. And the constitutive model of fluid element is described by the Gruneisen equation. Arbitrary Lagrangian-Eulerian (ALE) formulation is thought of as algorithms that perform automatic rezoning. It realized the advection of water and air in the ALE multi-material group.
Overview of the CAP1400 Reactor Hydraulic Simulation Tests
Authors: Shaoxuan Lin , Ming Zhang , Ying Fang , Zonghua Ding , Ping Yang , Sheng Wang , Wei Zhang , Yejia Liang 
Published: December 2017

CAP1400 is the trademark of a newly designed, by Shanghai Nuclear Engineering Research and Design Institute, third generation passive PWR. To verify the hydraulic functions and to optimize the design of the CAP1400 reactor-internals, a series of simulation tests was conducted including the use a 1/6 scale replica of the CAP1400 reactor-internal model and several special-purpose models. This paper describes the design of the test models, the test setups, the results and the major conclusions from the various simulation tests.
There are four major simulation tests, namely, a) the reactor-internal hydraulic test and b) the mixing test c) the flow path visualization test and d) the bypass flow tests. From these tests, the various hydraulic parameters of the CAP1400 reactor-internal design were obtained — including the pressure drop across the various flow paths inside the pressure vessel, the comparison of core inlet flow distribution with different flow distribution devices, the mixing factor between the two primary loops, the flow paths inside the pressure vessel and the flow characteristics, such as the resistance, of the bypass channels.
The results of these tests were not only useful in understanding the hydraulic characteristics of the CAP1400 design and provided the bases for design optimization and future safety evaluations, but also the evidence that CAP1400 is a superior design.
Investigation of Thermal Fatigue Characteristics of Charging Line of RCV in PWR
Authors: Ahmad Naeem , Li XiangBin , Ahmad Iftikhar , Li Nan , Ahmed Shahroze , Abdul Gayoum AlFadl Ahmed Abdul Ghani 
Published: December 2017

Nuclear Power Plant (NPP) components need to tolerate thermal constraints, internal pressure and thermal transients. These thermal transients being repeated again and again can lead to thermal fatigue of the component. It has significant effect on the degradation of the NPP components in long term. Studies of thermal fatigue on different NPP components such as mixing tees and valves have been carried out before but the charging line in the chemical and volume control system (RCV) of the NPP seems to have been ignored for thermal fatigue analysis.
Charging Line is the connection from RCV towards Reactor Coolant System (RCP). To enhance the safety of the charging line, thermal fatigue evaluation of piping system was performed using the Fluid Structure Interaction (FSI) analysis. Temperature distributions in the pipes were determined via thermal hydraulic analysis (CFX) and the results were applied to the structural model of the piping system to determine the thermal stress (Transient Structural).
Results revealed the location of fatigue cracks. Types of stress were identified that caused the fatigue damage. The CFD analysis enabled us to clarify the role of turbulence with respect to the thermal loading of the structure. The study will provide valuable information for establishing a permanent methodology to help minimize thermal fatigue damage in NPP components.
Simulation of Behaviors of a One-Third Scale Containment Model Test
Authors: Ren Guopeng , Pan Rong , Sun Feng 
Published: December 2017

Reactor containment of a nuclear power plant is a structure to ensure the safety of nuclear power plant. It acts as the last barrier to prevent the release of radioactive materials from NPP during accidents. Finite element models were established to simulate a 1/3 scale model of a reactor containment building under leakage test pressure. General finite element software ANSYS were applied. The nonlinear behavior of containment materials, geometric were taken into account in the analysis. The reliability of the finite element model was verified through the comparison of theoretical analysis results with experimental results. In the ANSYS finite element model, the concrete, steel bars and prestress tendons were separated and the prestress tendons were considered by the method of cooling method on the prestress tendon elements. The mechanical properties of the finite element model in the prestress tension process and the absolute internal pressure of 0.52MPa were analyzed.
Transient and time dependent losses were taken into account at the same time during the calculation of prestress of tendons, so as to calculate effective prestress at different locations of tendons. Calculation results of prestress losses show that the prestress losses at the hole of equipment hatch are larger than the other areas.
The results show that, the deformation of over-all structure of the containment is shrink inward under the action of prestress. And the simulation can achieve the consistent deformation effect between tendons and concrete. The maximum radial displacement of the whole containment structure is located at of 10 ° ∼ 20 °area on the right of the hole of the gate.
The effect of expansion deformation of the containment caused by design internal pressure is insufficient to offset the inward shrink effect generated by tendons, and the over-all structure of the concrete containment scale model is mainly under compressive stress. The containment test model is still with a large safety margin under the action of design internal pressure. The largest tensile stress is on the up and down areas of the internal sides of the equipment hatch, dome area close to ring beam, and bottom of perimeter wall close to the base slab. There is possibility of cracking on the concrete in limited local zones.
This benchmark can provide a reference for engineering design of containment.
Doctor won’t see you now: changing paradigms in mountain medicine
Authors: Szawarski Piotr , Hillebrandt David 
Published: December 2017

AbstractThe evolution in communication and digital technologies is revolutionising the practice of medicine. A physician is now able to oversee provision of healthcare at a distance. In this paper, we argue that practice of telemedicine is an essential and evolving aspect of high altitude and expedition medicine. We acknowledge the lack of international consensus, limited legislation as well as possible reservations towards telemedical practice. Given some unique social and medical aspects of participation in remote, high altitude expeditions we propose a number of guiding principles for those involved in provision of telemedical services for such endeavours.