bluuebunny/crossref_metadata_2025_split huggingface dataset:
Out-of-Phase and In-Phase Thermo-Mechanical Fatigue Response Simulations of Haynes 230
Authors: Ahmed Raasheduddin , Barrett Paul R. , Hassan Tasnim 
Published: December 2017

Temperatures at critical locations in propulsion turbine engine combustor components can be as high as 982°C (1800°F). High temperature thermal gradients, and start-up and shut-down operations of gas turbines, induce thermo-mechanical fatigue (TMF) failure. Dwell periods at high temperatures accompanied by repeated loading cycles, eventually lead to failure of the components through creep-fatigue processes. In an effort to decipher the complex high temperature phenomena, a large set of isothermal and thermo-mechanical fatigue experiments have been carried out on the gas turbine combustor liner material, Haynes 230. The out-of-phase strain-controlled TMF experiments with compressive peak hold result in mean stress evolution in the tensile direction, whereas the in-phase TMF experiments with tensile peak hold result in mean stress evolution in the compressive direction. Experimental results indicate that the maximum temperature in the loading cycle influences the material property evolution with cycle. A unified viscoplastic constitutive model based on the Chaboche type nonlinear kinematic hardening rule was developed, including the added features of strain range dependence, rate dependence, temperature rate dependence, static recovery, mean stress evolution, and maximum temperature influence. The new constitutive model was validated against stress-strain responses of Haynes 230 under TMF loading.
Paper published with permission.
High Temperature Testing and Analysis of a Class 300 Bolted Flange Joint
Authors: Mann J. Adin , Hilsabeck Jeremy , Mckoon Cale 
Published: December 2017

When class 300 flange bolted joints are held at temperatures in the material creep range, it is documented that the bolt loads can relax. Tests and analysis are being performed with the goal of developing a validated FEA simulation approach to predicting the impact of creep on the bolt loads. The bolt load and gasket geometry are evaluated upon bolt up and after being heated to 1100 deg F. Tests are performed with and without a gasket to separate the impact of the gasket relaxation and flange material creep. The results of the tests and analysis approaches will be presented.
Paper published with permission.
A Comparative Creep and Creep Fatigue Analysis of an Austenitic Steam Superheater
Authors: Di Lillo Patrizio , Gaetani Luca , Germoni Lorenzo , Masia Alessandro 
Published: December 2017

A steam superheater is a component which operates at elevated temperature in the creep regime.
The component is normally not working in cycling conditions. Due to operation failures, the component has been subjected to several unplanned start-ups and shutdown. The SSH has reported a very high level of damage of the tubesheet to shell junction in field (high deformation and weld cracks).
Several investigations have been performed to analyze the effect of temperature, pressure and external loads on the equipment with finite element techniques and different creep and creep-fatigue evaluation approaches: ASME III subsection NH, simplified inelastic analysis and a full creep model have been used. A comparison of the results is reported in this paper.
Paper published with permission.
Cracking in Grade 23 Weldments at Elevated Temperatures
Authors: Foulds Jude R. , Siefert John A. 
Published: December 2017

Recent, brittle fracture in Grade 23 power plant components at relatively low temperatures has increased the need to assess the cracking behavior of this material. Time-dependent cracking in the heat-affected zone of Grade 23 weldments was assessed using crack growth testing of subsize compact tension specimens at a temperature (482°C, 900°F) characteristic of the upper portion of a furnace wall in supercritical boilers. Results of additional testing at a higher, typical design temperature (566°C, 1050°F) for superheater and reheater tubing and headers will be reported later. Post-test metallurgical evaluation of the cracking morphology was conducted using traditional light microscopy and laser microscopy. Although large-scale creep deformation is absent under these lower test temperature conditions, significant weldment heat-affected zone intergranular cracking was documented. An example of application of the data to the inservice integrity and life assessment of a furnace tube is also described, providing preliminary perspective on the factors controlling lifetime and manageability of integrity.
Paper published with permission.
An R5 Based Creep-Fatigue Critical Flaw Assessment of an In-Service Reformer Piping Tee Using Finite Element Analysis
Authors: Prueter Phillip E. , Dewees David J. , Brown Robert G. 
Published: December 2017

This study employs three dimensional (3D) non-linear, finite element analysis (FEA) to supplement a critical flaw sizing assessment of an in-service reformer piping tee. The analysis is guided by the EDF (Électricité de France) Assessment Procedure R5 (formerly from British Energy Generation LTD. or BEGL) as well as Part 10 of API 579-1/ASME FFS-1. Specifically, Volume 4/5 of R5, which addresses crack growth, is used to determine the largest permissible flaw as a function of operational cycles and time at temperature. Required stresses are generated using FEA as well as the simplified reference stress techniques of R5 as appropriate. The analysis explicitly considers thermal transients, as well as cyclic plasticity. Furthermore, modeling of steady state operating conditions considers creep in the FEA material model. Additionally, creep-fatigue flaw growth is considered for a range of initial defect sizes. The targeted inelastic, non-linear FEA is leveraged to remove significant uncertainty and conservatism, and the simplified techniques of R5 are employed wherever reasonable to give the most efficient analysis possible. This investigation provides estimates of flaw propagation rates based on historical cyclic operation and permits determination of reasonable inspection intervals for the reformer tee in question.
Paper published with permission.
Simplified Methods for Elevated Temperature Structural Design: An Overview of Some Current Activities
Authors: Jetter Robert I. , Wang Yanli , Carter Peter , Sham T.-L. (Sam) 
Published: December 2017

Elevated temperature design criteria for Class 1 nuclear components employ two fundamental approaches for evaluation of structural integrity in the temperature regime where creep effects are significant: full inelastic analysis to predict the actual stress and strain resulting from time dependent loading conditions and simplified methods which bound the actual response with, conceptually, simpler material models and analytical procedures. However, the current simplified methods have been found to be more complex for real component design applications than originally envisioned. There is an added complication that the current simplified methods are considered inappropriate in the very high temperature regime where there is no distinction between plasticity and creep. Recently, some improved, less complex methods have been proposed which would overcome these objections. One set of criteria are based on elastic-perfectly plastic (E-PP) analysis methods. Draft code cases have been prepared which address the use of the E-PP methodology to primary loading, strain limits and creep-fatigue damage evaluation. Another proposed criterion is based on the use of test specimens which include the effects of stress and strain redistribution due to plasticity and creep to develop creep-fatigue damage evaluation design curves. An overview of the key features, associated analytical and experimental verification, status and path forward are presented. Although targeted to nuclear components, these criteria also have potential application to non-nuclear components and operating temperatures below the creep regime.
Paper published with permission.
Observation of the Transformation in Microstructure and Vickers Hardness of Ferrous Model Alloy by Electromagnetic Induction Heating to Form a High Temperature Gradient
Authors: Ye Changqing , Li Qiulin , Wu Ping , Tang Guoyi , Liu Wei 
Published: December 2017

Nuclear radiation can induce hardening and embrittlement of RPV steel. Solute defect copper rich precipitate / clusters is considered harden the base steel. Reduce or delay the irradiation hardening and embrittlement of RPV steel could be effective to ensure the safe operation of nuclear power. Furthermore, the expected target of the improved RPV steel will assist consummate nuclear power evaluation to decide whether feasible to extend the service lifetime of nuclear power plant from 40 to 60 years facing at present in China nuclear industry.
In order to study the precipitate phase Cu rich clusters, but different from the traditional methods, the present work employed a new technique, electromagnetic induction heating to produce a high temperature gradient along the model rod Fe-1.2wt.%Cu alloy sample. The results exhibit that this method can speed up the research process and is convenient to study the transformation in microstructure and hard / brittle defect phase formation of copper solute rich clusters. In addition, the change of Vickers hardness along the high temperature gradient assist denote the transformation in microstructure and the Cu rich clusters in quantity roughly. And large rolling deformation can promote formation of Cu rich cluster.
Improvement of Probabilistic Fracture Mechanics Analysis Code PASCAL-SP With Regard to Primary Water Stress Corrosion Cracking
Authors: Mano Akihiro , Yamaguchi Yoshihito , Katsuyama Jinya , Li Yinsheng 
Published: December 2017

Structural integrity assessments for cracked nuclear components are currently performed on the basis of deterministic fracture mechanics in accordance with the Rules on Fitness-for-Service for Nuclear Power Plant of the Japan Society of Mechanical Engineers. On the other hand, probabilistic fracture mechanics (PFM) is expected as a rational method for a structural integrity assessment because it can account for the uncertainties and scatters of various influencing factors and can evaluate quantitative values such as the failure probabilities of the components as the solutions. In the Japan Atomic Energy Agency (JAEA), a PFM analysis code PASCAL-SP was developed in order to evaluate the failure probability of nuclear pipe by taking into account aging degradation mechanisms such as inter-granular stress corrosion cracking (IGSCC) and fatigue in the boiling water reactor (BWR) environment.
Recently, a number of cracks due to primary water stress corrosion cracking (PWSCC) have been detected in nickel-based alloy welds in the primary piping of pressurized water reactors (PWRs). Therefore, the structural integrity assessment for primary piping taking PWSCC into consideration has become important.
This paper details the improvement of PASCAL-SP to evaluate the failure probability of primary pipe taking PWSCC into consideration. We introduce several probabilistic evaluation models such as crack initiation, crack growth and crack detection models related to PWSCC into PASCAL-SP. As numerical examples, the failure probabilities for circumferential and axial cracks in nickel-based alloy weld in pipe in the PWR primary water environment are calculated. We also evaluate the influence of non-destructive inspection on failure probabilities. On the basis of the numerical results, we conclude that the improved PASCAL-SP is useful for evaluating the failure probability of primary pipe taking PWSCC into account.
The Hot Test of Primary Separation Equipment of Steam Generator for CAP1400 Nuclear Power Plant
Authors: Bingbin Ying , Xilin Li , Feixi Zhang , Shaoxuan Lin 
Published: December 2017

A new type of Steam separator for CAP1400 Steam Generator is introduced. The result of characteristic experiment, the variables of which contain steam load, saturated water flow and water level, shows its separation characteristics are suitable for CAP1400 Steam Generator. Steam generator is a major equipment of pressurized water reactor power plant, one function of which is to transfer heat from reactor to feedwater and to generate qualified steam. Steam separator, located in upper shell of Steam generator, is used to separate steam from saturated water/steam mixture which exits from the tube bundle region.
Steam separating process usually consists of three stages. Centrifugal separation first happens in Primary separator, where most of water is separated. Gravity separation is achieved in the space formed between Primary separator and the dryer. Finally, as the relatively low moisture content mixture flows through the chevron-type dryer, the qualified steam is obtained.
The new developed primary separator, named SP3 type, consist of raise cylinder, swirl impeller, downcomer and vane. As the steam/water mixture enters the primary separator, a centrifugal force is induced by the swirl. Thus the greater dense water in the steam-water mixture is forced toward the outside diameter of the separator, while the less dense steam stays in the central portion of the separator. The separated water leaved the separator through the downcomer. Relatively high quality steam exits the separator, passing upward through the center of raise cylinder.
The paper introduced the SP3 primary separator developed by hot test, which includes test device, device design, test method, and result analysis. Separation characteristics of SP3 primary separator are obtained by various tests. The result of varying steam load test shows the outlet moisture is 13.3% under the condition of 100% steam load and 4.3 recirculation ration. The result of varying saturated water flow test shows the outlet moisture increases as the saturated mixture flow rate increases. The outlet moisture increased slowly as saturated water flow rate increases when steam load is at 50% and 100%, while the outlet moisture increases fast when steam load reaches at 120%. The result of varying water level test shows the water level is not a major factor for the SP3 primary separator performance.
By analyzing experimental data, a fitting formula of the primary separator performance is obtained. the steam flow rate is the most sensitive parameter affecting primary separator performance in that formula. The water flow is the subordinate parameter for the primary separator performance. the resistance characteristics of separator is also obtained, which supports the research and development of the CAP1400 steam generator..
Large Scale PWR Passive Containment Cooling System Wind Tunnel Test
Authors: Xinxin Pan , Jingyu Huang , Chunjing Song 
Published: December 2017

CAP1400 is a large passive pressurized water reactor nuclear power plant, which relies on engineering safety features such as passive containment cooling system (PCS) to remove the decay heat in the containment and mitigate accident consequences. PCS is designed to perform passive containment cooling which is mainly dependent on natural convection inside the containment and inner wall condensation heat transfer, outer containment surface water film coverage and evaporation heat transfer and external air flow path cooling performance, etc. Among them, the key factors that affect the performance of the external air flow path include the flow resistance characteristics of the air flow path and the wind-direction neutrality characteristics. The relevant performance will be the important design input of the accident analysis, which will directly affect the safety of the power plant.
During the normal operation of power plant, the PCS air flow path is influenced by the external environment, and its internal flow is very complicated. Designers are often lack of data support, and can’t fully consider the impact of environmental flow. In order to fully study the performance of PCS air flow path, it is necessary to perform PCS integrated scaled wind tunnel test.
According to the original design of CAP1400 PCS system, the model scale research is developed and CAP1400 PCS wind tunnel test scaled model is established and the scale is 1:100. The test model includes shield building model and the surrounding plant model, which contain pressure measuring points uniformly distributed in 6 horizontal cross sections of the shield building. The pressure measuring point arrangement does not affect air flow in the air flow path. The following wind tunnel tests are simulated in different wind speed including 15m/s, 20m/s, 10m/s, 25m/s. The air flow pressure, wind velocity at the inlet and outlet of air flow path and the pressure distribution of inner annulus and outer annulus are measured in order to study the air flow pressure drop and wind-direction neutrality characteristics, and the wind tunnel test also considers the different wind direction angle, with and without the surrounding buildings and the effects of different landforms.
The test results show that the flow rate of inlet and outlet of air flow path is balanced and the wind velocity at the upwind and central area of the flow path outlet is larger than other area, and a large vortex comes on the leeward side near the wall. The local uneven flow phenomenon exists in the outer annulus of the air flow path, but the wind pressure distribution of inner annulus is not affected by environment wind speed, wind direction angle, landforms and the surrounding buildings. So CAP1400 PCS air flow path has the characteristics of wind direction neutrality, and the natural convection of the air flow path will not be adversely affected by the environment wind.
Seismic Performance of PCS Water Storage Tank Considering Fluid-Structure Interaction
Authors: Gou Hongliang , Shang ZiDuan , Sun Yugang , Chu Meng , Ge Honghui 
Published: December 2017

Water sloshing of PCS water storage tank (PCSWST) can cause a significant effect on the dynamic response of Shield Building under seismic loads. It is complicated to perform the calculation of water sloshing especially for the tanks with irregular shapes. Consequently, it is important to establish an appropriate equivalent mechanical model for simulation[1], [2].
In this paper, the water sloshing is firstly investigated based on the potential flow theory, which including the seismic modal analysis. Based on the theoretical research, a highly efficient (simplified) calculation formula is derived, which mainly considering the impulse mass, convective mass, position function and spring stiffness etc., through this way the equivalent model for PCSWST is established by applying mass-spring element. The equivalent models based on Housner & Graham theory[3], [4] are also established. Additionally, the 3-D finite element model of water sloshing considering fluid-structure interaction is established by using the software of Ansys.
Total of four models are built as shown in the paper, then modal analysis and dynamic response under earthquake excitation are performed using ANSYS. The results are compared to justify the equivalent model in this paper. The results indicate that Graham formula did not provide the correct location expressions for the convective masses. The expressions for the impulsive mass and its position given by Housner are not satisfactory. As a comparison, the results from the equivalent model, which is recommended in this paper, can best fit the data from finite model. From above results and comparisons, a more reasonable and refined equivalent model for PCSWST design is provided.
Based on the equivalent model established, the influence on structure caused by the increase of water mass is analyzed. The results from the seismic analysis are compared, including member force, shear strain and shear force. Based on the research, the feasibility of the design is analyzed, which can provide important support for the structural design.
Finally, the seismic reduction of water tank is studied using the finite element model established in this paper. The horizontal and vertical anti-sloshing baffles are designed. The maximum acceleration and displacement corresponding to different baffle length are compared to study the effect of the seismic reduction.
Study on the Design Methodology of Swirl Nozzle for Nuclear Reactor Pressurizer
Authors: Cao Xiong , Ding Zhiwei 
Published: December 2017

Pressurizer is one of the most important components in reactor coolant system of a nuclear power plant, which operates normally at pressure of 15.4 MPa and temperature of 345°C[1]. The main function of pressurizer is to regulate the pressure in the reactor coolant system by either cooling the steam or heating the saturated water in its upper zone. When the pressure in the reactor coolant system increases, it will distribute cold water to decrease its temperature and pressure through atomizing the reactor coolant with swirl spray nozzle in pressurizer. Swirl nozzle is the key part of pressurizer with swirl structure of full cone spray pattern, and the atomization performance include drop size, spray angle and distribution, also it is characterized by huge flow rate and low pressure drop, and its atomization performance decides the quality of pressure control of the reactor coolant system. To enhance the independent design level of both pressurizer and cooling system, it’s necessary to study the atomization performance of swirl nozzle for nuclear reactor pressurizer.
Aimed at improving atomization performance of swirl spray nozzle, the structure design methodology of nuclear reactor pressurizer was studied systematically in three aspects including theory design, numerical simulation and test confirm in this thesis. Through designing the swirl nozzle structure according to similar design formula of spray nozzle in theory, especially studying the influence of different structures that mainly include internal swirl structure on internal flow field of swirl nozzles, the primary structure parameters of swirl nozzle were confirmed. Then, through numerical simulation of the internal flow field, flow rate and pressure drop, and swirl core structure of the swirl nozzle (by building physical model and mathematic model according to the spray nozzle structure), the atomization performance of the nozzle was analyzed. On this basis, the typical swirl nozzle was designed and tested, which included spray angle, flow rate as well as pressure drop tests, and spray drop tests, and the applicability of the computational fluid dynamics (CFD) method was verified when it was applied in swirl nozzle design. Finally, the design method of swirl nozzle with deep groove of swirl core for pressurizer was put forward.
Through this studying of theoretical calculation, numerical simulating and test, the correlation between the structural parameters of swirl nozzle and atomization performance was achieved, meanwhile design, analysis and test methods of spray nozzle with low pressure drop and huge flow rate were established. It is helpful to realize the independent design of pressurizer’s swirl nozzle and even to put forward the design methodology of pressurizer’s swirl nozzle with our own characteristic.