The Experts below are selected from a list of 222 Experts worldwide ranked by ideXlab platform
Jin Jiang - One of the best experts on this subject based on the ideXlab platform.
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reliability evaluation of secondary support structure in hypothetical drop of reactor Core Barrel assembly
Nuclear Engineering and Design, 2019Co-Authors: Jin JiangAbstract:Abstract Dynamic responses of the secondary support structure are investigated to evaluate its structural reliability in a hypothetical drop impact accident of the reactor Core Barrel assembly in this work. Impact analyses both in still water situation (20 °C) and dynamic water situation (289 °C) are carried out by LS-DYNA. Tensile tests in both temperature cases are conducted for the secondary support structure material (Inconel 690) in order to make the simulation results more accurate with real material properties. Various aspects of structural dynamic responses during impact are analyzed, including distributions of equivalent stress, equivalent plastic strain, and maximum principal stress as well as time histories of impact force, vertical displacement, vertical velocity, and energy. In still water situation, equivalent stress of the secondary support structure exceeds the yield strength due to the higher initial impact velocity and downward base acceleration. But in dynamic water situation, there is no plastic deformation due to the lower initial impact velocity and upward base acceleration. Peak values of the maximum principal stress distributed in the secondary support structure are less than the corresponding ultimate tensile strengths respectively in both analysis cases. Results reveal that both of the secondary support structure and the reactor pressure vessel meet the design criteria of RCC-M Code with respect to stress limits, thus the structural reliability can be ensured in this drop accident.
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Structural integrity assessment of secondary support structure and reactor pressure vessel following drop impact of Core Barrel assembly
Annals of Nuclear Energy, 2019Co-Authors: Jin Jiang, Yucheng Wang, Shuang Luo, Rui YingAbstract:Abstract This work presents a study on hypothetical drop accident of the Core Barrel assembly in a pressurized water reactor, in which structural dynamic responses of the secondary support structure and the reactor pressure vessel during impact are investigated numerically. The analysis results are conservative enough by considering the drop impact accident in an air environment. And the tensile test was conducted for the Inconel 690 material of secondary support structure in order to input real material property in simulations. During the impact, the peak impact force and impact velocity of the falling Core Barrel assembly are respectively 19,881 kN and 0.627 m/s. With respect to stress distributions, peak magnitudes of effective stress and maximum principal stress are both concentrated at the top edge of connection location between the secondary support structure and the reactor pressure vessel. For both of these components, their effective stresses exceed the corresponding yield strengths while the maximum principal stresses are less than their ultimate tensile strengths, indicating that they deform plastically in this drop impact accident. The impact energy absorbed by the reactor pressure vessel is much larger than that by the secondary support structure. With respect to a single support key, the first and second peak strain energy are concentrated in the connection location and contact location respectively while the free end is almost unaffected. And three concerned through-thickness paths are defined for stress linearization in which membrane stress and bending stress are separated. Stress results meet the design criteria of the ASME Boiler and Pressure Vessel Code with respect to stress limits, indicating that both of the secondary support structure and the reactor pressure vessel maintain their structural integrity following drop impact of the Core Barrel assembly. This investigation makes a contribution to the safety analysis of the pressurized water reactor.
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Hydrodynamic buffer effect on the fall-off Core Barrel assembly in hypothetical drop accident
Nuclear Engineering and Design, 2018Co-Authors: Jin Jiang, Yanhui Li, Wensheng Zhao, Rui Ying, You FuAbstract:Abstract Hypothetical accident analysis is of great importance for design and research of nuclear reactor. The Core Barrel assembly is assumed to fall off in a hypothetical drop accident and we conducted a numerical simulation to study the hydrodynamic buffer effect with CFD dynamic mesh. An orifice plate with square holes was designed to simplify the reactor Core and two kinds of circumstances were compared in this investigation, including the normal operating condition (high-temperature dynamic water) and the still water situation (low-temperature). Distribution of flow field (e.g. pressure contour, velocity vector and stream line) at the moment of impact was analyzed. Due to the significant buffer effect of dynamic coolant flow, it has a greater hydrodynamic force and a smaller falling velocity respectively in normal operating condition than that in still water situation. The change laws of the hydrodynamic force and the falling velocity are similar in normal operating condition that they increase first and then decrease from their peak values (2410.2 kN and 0.16 m/s respectively). Whereas in still water situation, the hydrodynamic force rapidly increases to 835.5 kN (peak value) and maintains stable a level subsequently, meanwhile the falling velocity increases continuously with parabolic growth. The fall-off assembly is inferred to achieve effective hydrodynamic buffer in drop accident during normal operation. As well, the investigation also provides input parameters to subsequent research on impact load of the secondary support structure.
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Investigation of hydraulic force of a 1000-MW pressurized water reactor under hypothetical drop accident of Core Barrel assembly
Advances in Mechanical Engineering, 2016Co-Authors: Jin Jiang, Lisheng ZhangAbstract:In this study, we conducted a simulation investigation of a hypothetical Core Barrel assembly drop accident based on the dynamic mesh model. The investigation included three separate circumstances: high-temperature dynamic water, high-temperature still water, and low-temperature still water. Results showed that dynamic parameters (e.g. velocity and resultant force) of the drop assembly in high-temperature dynamic water are smaller than those in still water due to the buffer effect of the flowing water. The velocity and hydraulic force in the dynamic water situation increased quickly at first and then remained stable throughout the remainder of the experiment; due to turbulence effect, the hydraulic force fluctuated slightly, first increased, then decreased, and then reached as stable a level in still water as that in dynamic water. The velocity of the drop assembly showed parabolic growth
Imre Pazsit - One of the best experts on this subject based on the ideXlab platform.
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first evidence of the pivotal motion tilting mode of the Core Barrel in the ringhals 4 pwr
Physics of Reactors 2016 PHYSOR 2016: Unifying Theory and Experiments in the 21st Century, 2016Co-Authors: C Montalvo, Imre Pazsit, Henrik Nylen, Victor DykinAbstract:The Division of Subatomic Physics and Plasma Physics (formerly Division of Nuclear Engineering) in Chalmers, Goteborg, and the Ringhals Nuclear Plant have investigated the Core Barrel vibrations in the Ringhals PWRs over the last 20 years. Based on the different symmetry properties of the vibration modes, a mode separation technique was developed to enhance the contributions from the different modes. Recent observations of wear at both the lower and upper Core-Barrel-support structures in the Ringhals PWRs indicated that vibration modes of the Core Barrel other than pendular (beam mode) and shell mode are likely to occur. A beam mode type movement alone is not able to explain such a wear, and therefore, it is fair to assume that the vibration mode in question is a small amplitude periodic tilting movement of the Core Barrel around a horizontal, diagonal pivot at the half height of the Core. In this work, ex-Core data taken in the Ringhals-4 PWR were analyzed in order to find evidence of such a tilting movement. First, cross spectra between various ex-Core detectors were calculated and analyzed to locate the frequency range of the new vibrational mode. Then, a model based on symmetry considerations was derived in order to extract the sought mode from the spectra. The measurements were evaluated by the new mode enhancement technique. The results show that it is possible to enhance such a mode and find it in the spectra by properly combining the signals in the time domain.
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Developments in Core-Barrel Motion Monitoring and Applications to the Ringhals PWR Units
Nuclear Science and Engineering, 2016Co-Authors: Imre Pazsit, Henrik Nylen, C Montalvo, Tell Andersson, Augusto Hernández-solís, Bernitt CartemoAbstract:Core-Barrel motion (CBM) surveillance and diagnostics, based on the amplitude of the peaks of the normalized auto power spectral densities (APSDs) of the ex-Core neutron detectors, have been performed and continuously developed in Sweden and were applied for monitoring of the three PWR units, Ringhals 2 to 4. From 2005, multiple measurements were taken during each fuel cycle, and these revealed a periodic behavior of the 8-Hz peak of the beam-mode motion: the amplitude increases within the cycle and returns to a lower value at the beginning of the next cycle. The work reported in this paper aims to clarify the physical reason for this behavior. A combination of a mode separation method in the time domain and a nonlinear curve fitting procedure of the frequency spectra revealed that two types of vibration phenomena contribute to the beam-mode peak. The lower frequency peak around 7 Hz in the ex-Core detector APSDs corresponds to the CBM, whose amplitude does not change during the cycle. The higher frequency peak around 8 Hz arises from the individual vibrations of the fuel assemblies, and its amplitude increases monotonically during the cycle. This paper gives an account of the work that has been made to veri,b, the above hypothesis.
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refined method for surveillance and diagnostics of the Core Barrel vibrations of the ringhals pwrs
PHYSOR 2014 – The Role of Reactor Physics Toward a Sustainable Future, 2014Co-Authors: Imre Pazsit, Henrik Nylen, Cristina Montalvo MartinAbstract:Surveillance and diagnostics of Core Barrel vibrations has been performed in the Swedish Ringhals PWRs for several years, with main focus on the pendular motion (beam mode). The monitoring of the beam mode showed that the amplitude of the corresponding peak in the ex-Core neutron spectra increases along the cycle, and decreases after refueling. Previous investigations on the reason of this behaviour, i.e. whether it is due to the increase of the Core Barrel vibration amplitude or to the increase of the neutron physics coupling between vibrations and neutron noise, were not decisive. The objective of the work reported here is to clarify this question. From frequency analysis, two modes of vibration have been identi-fied in the frequency range of the beam mode. Several results coming from the trend analysis performed during recent years indicate that one of the modes is due to the Core Barrel motion itself and remains constant during cycle, and the other is due to the indi-vidual flow induced vibrations of the fuel elements, showing an increasing trend during the cycle. In this work, the method to separate the contributions from the two modes has been refined, and the results of this approach to the latest measurements are presented. The results confirm the origin of the two vibration modes and show constant amplitude of the Core Barrel motion throughout the cycle.
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SURVEILLANCE AND DIAGNOSTICS OF THE BEAM MODE VIBRATIONS OF THE RINGHALS PWRs
2012Co-Authors: Cristina Montalvo Martin, Imre Pazsit, Henrik NylenAbstract:Surveillance of Core Barrel vibrations has been performed in the Swedish Ringhals PWRs for several years. This surveillance is focused mainly on the pendular motion of the Core Barrel, which is known as the beam mode. The monitoring of the beam mode has suggested that its amplitude increases along the cycle and decreases after refuelling. In the last 5 years several measurements have been taken in order to understand this behaviour. Besides, a non-linear fitting procedure has been implemented in order to better distinguish the different components of vibration. By using this fitting procedure, two modes of vibration have been identified in the frequency range of the beam mode. Several results coming from the trend analysis performed during these years indicate that one of the modes is due to the Core Barrel motion itself and the other is due to the individual flow induced vibrations of the fuel elements. In this work, the latest results of this monitoring are presented.
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Beam mode Core-Barrel vibrations in the PWRs Ringhals 2-4
2006Co-Authors: M. Pazsit, Carl Sunde, Imre PazsitAbstract:Analysis of Core-Barrel vibrations in the Swedish Ringhals PWRs has been performed by Chalmers since the early 1990's. In the first phase of this work, between 1991 and 1998, the evaluation method has been developed such that it made a consistent comparison between different measurements possible. A trend analysis showed that the beam mode amplitudes have steadily increased between 1991 and 1998 in all three plants. This paper is to report on the second phase of the work, performed during 2005, on measurements made both before and after the summer outage 2005 in all three plants. During the summer outage, the hold-down spring in Ringhals-3 was replaced. The analysis shows that the vibration amplitudes increased in an accelerated rate between 1998 and 2005 in all three plants. In Ringhals 3, however, after the change of the hold-down spring, the beam mode amplitude has reverted to close its original level of 1991. It became also clear that the extraction of the information from the vibration peaks needs to be refined and made less subjective. A new method of algorithmic peak separation was elaborated, which supplies more information than the previous analysis; it gives also the peak width in addition to peak more » amplitude and peak frequency, while also supplying more accurate estimates for the latter two. (authors) « less
V. Yu. Gonchar - One of the best experts on this subject based on the ideXlab platform.
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A simple nonlinear model for pendular chaotic oscillations of reactor Core Barrel
Progress in Nuclear Energy, 1999Co-Authors: Yu. L. Bolotin, V. V. Bulavin, A. V. Chechkin, V. Yu. GoncharAbstract:Abstract We propose a non-linear model reproducing the unidirectional (quasi)pendular motion of pressurized water reactor (PWR) Core Barrel (CB), which is in intermittent contact with reactor pressure vessel (RPV) in places of lower fastenings. Analytic formulae are obtained, which allow one to elucidate the influence of reactor parameters on the character of motion. We show by the example of the WWER-1000 reactor that, in the framework of our model, there exists a possibility of chaotic pendular oscillations in the certain region of three parameters, namely, (i) the pendular oscillation frequency, (ii) the resulting force acting on the CB at the main circulation pump (MCP) frequency from the coolant flows of opposite legs, and (iii) the clearance between the CB and RPV. Vibratory monitoring, which is realized in modern diagnostic systems, in principle, allows one to distinguish between regular and chaotic regimes, thus giving information about the state of fastenings in the lower part of the CB and RPV. We also demonstrate the possibility of stabilizing chaotic oscillations by applying recently proposed control technique.
Rui Ying - One of the best experts on this subject based on the ideXlab platform.
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Structural integrity assessment of secondary support structure and reactor pressure vessel following drop impact of Core Barrel assembly
Annals of Nuclear Energy, 2019Co-Authors: Jin Jiang, Yucheng Wang, Shuang Luo, Rui YingAbstract:Abstract This work presents a study on hypothetical drop accident of the Core Barrel assembly in a pressurized water reactor, in which structural dynamic responses of the secondary support structure and the reactor pressure vessel during impact are investigated numerically. The analysis results are conservative enough by considering the drop impact accident in an air environment. And the tensile test was conducted for the Inconel 690 material of secondary support structure in order to input real material property in simulations. During the impact, the peak impact force and impact velocity of the falling Core Barrel assembly are respectively 19,881 kN and 0.627 m/s. With respect to stress distributions, peak magnitudes of effective stress and maximum principal stress are both concentrated at the top edge of connection location between the secondary support structure and the reactor pressure vessel. For both of these components, their effective stresses exceed the corresponding yield strengths while the maximum principal stresses are less than their ultimate tensile strengths, indicating that they deform plastically in this drop impact accident. The impact energy absorbed by the reactor pressure vessel is much larger than that by the secondary support structure. With respect to a single support key, the first and second peak strain energy are concentrated in the connection location and contact location respectively while the free end is almost unaffected. And three concerned through-thickness paths are defined for stress linearization in which membrane stress and bending stress are separated. Stress results meet the design criteria of the ASME Boiler and Pressure Vessel Code with respect to stress limits, indicating that both of the secondary support structure and the reactor pressure vessel maintain their structural integrity following drop impact of the Core Barrel assembly. This investigation makes a contribution to the safety analysis of the pressurized water reactor.
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Hydrodynamic buffer effect on the fall-off Core Barrel assembly in hypothetical drop accident
Nuclear Engineering and Design, 2018Co-Authors: Jin Jiang, Yanhui Li, Wensheng Zhao, Rui Ying, You FuAbstract:Abstract Hypothetical accident analysis is of great importance for design and research of nuclear reactor. The Core Barrel assembly is assumed to fall off in a hypothetical drop accident and we conducted a numerical simulation to study the hydrodynamic buffer effect with CFD dynamic mesh. An orifice plate with square holes was designed to simplify the reactor Core and two kinds of circumstances were compared in this investigation, including the normal operating condition (high-temperature dynamic water) and the still water situation (low-temperature). Distribution of flow field (e.g. pressure contour, velocity vector and stream line) at the moment of impact was analyzed. Due to the significant buffer effect of dynamic coolant flow, it has a greater hydrodynamic force and a smaller falling velocity respectively in normal operating condition than that in still water situation. The change laws of the hydrodynamic force and the falling velocity are similar in normal operating condition that they increase first and then decrease from their peak values (2410.2 kN and 0.16 m/s respectively). Whereas in still water situation, the hydrodynamic force rapidly increases to 835.5 kN (peak value) and maintains stable a level subsequently, meanwhile the falling velocity increases continuously with parabolic growth. The fall-off assembly is inferred to achieve effective hydrodynamic buffer in drop accident during normal operation. As well, the investigation also provides input parameters to subsequent research on impact load of the secondary support structure.
Anne Demma - One of the best experts on this subject based on the ideXlab platform.
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Materials Aging Degradation of Reactor Vessel Internals: Part II—Structural Evaluation and Global Finite Element Models
Volume 7: Operations Applications and Components, 2009Co-Authors: Matthew Snyder, Tama´s R. Liszkai, Anne DemmaAbstract:Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), developed PWR Internals Inspection and Evaluation (I&E) Guidelines mainly to support license renewal of U.S. plants [2]. The functionality analysis of reactor internals components and assemblies was one of the tools used to develop these guidelines. The purpose of the functionality analysis is to provide a best estimate evaluation of the reactor internals Core Barrel assembly for materials degradation and to assess whether the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement [3], 4, [5]. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the Core Barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the Core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major analytical calculation, which facilitates the development of operating stress fields within the RV internals. Structural analysis was performed as two parts. First, a global structural model of the Core Barrel assembly was used to represent the interaction of components of the Core Barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections, therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [6] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, detailed in this paper, describes global structural finite element models. Part III, to be also presented at PVP-2009 [7], presents a description of local models of bolted connections, results, and conclusions.Copyright © 2009 by ASME
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Materials Aging Degradation of Reactor Vessel Internals: Part II—Structural Evaluation and Global Finite Element Models
Volume 7: Operations Applications and Components, 2009Co-Authors: Tama´s R. Liszkai, Matthew Snyder, Anne DemmaAbstract:Pressurized water reactor (PWR) vessel internals components can experience material aging and degradation due to irradiation [1]. The Electric Power Research Institute (EPRI), under sponsorship of the Materials Reliability Program (MRP), developed PWR Internals Inspection and Evaluation (I&E) Guidelines mainly to support license renewal of U.S. plants [2]. The functionality analysis of reactor internals components and assemblies was one of the tools used to develop these guidelines. The purpose of the functionality analysis is to provide a best estimate evaluation of the reactor internals Core Barrel assembly for materials degradation and to assess whether the components retain their function. The evaluation uses an irradiated material-specific constitutive model for use in a finite element analysis representing the current state of knowledge for plasticity, creep, stress relaxation, void swelling, and embrittlement [3], 4, [5]. This constitutive model is a function of temperature and fluence. The analysis focuses on finding the integrated effects of material aging combined with steady-state operational characteristics of the reactor vessel (RV) internals. In order to evaluate the potential failure mechanisms of the Core Barrel assembly, finite element models were developed capable of representing the complex interactions between the components. The goal of this specific analysis is to characterize the potential failure modes, spatial and chronological distribution of potential component failures for a representative model of the Babcock & Wilcox (B&W) designed plants. Evaluation of the reactor vessel internals for materials aging degradation involves three analytical calculations. Radiation calculations of the Core provide essential information on radiation dose and heat rates, due to gamma-heating, of the RV internals. The computational fluid dynamics domain (CFD) allows the evaluation of the RV internals temperatures through conjugate heat transfer (CHT) analysis coupled with coolant flow. Detailed structural analysis of the RV internals components and bolted connections is the third major analytical calculation, which facilitates the development of operating stress fields within the RV internals. Structural analysis was performed as two parts. First, a global structural model of the Core Barrel assembly was used to represent the interaction of components of the Core Barrel assembly during 60 years of operation. The global model does not include detail of the areas of stress concentration within bolted connections, therefore local models of selected bolts were developed. Results of both the global and local models were used as a basis for evaluating age-related effects. The description of the functionality analysis for the B&W designed RV internals is divided into three papers. Part I was presented in PVP-2008 [6] and included a description of the overall methodology with special attention to CFD-CHT evaluations. Part II, detailed in this paper, describes global structural finite element models. Part III, to be also presented at PVP-2009 [7], presents a description of local models of bolted connections, results, and conclusions.