Simulation of Acoustic-Structure Interaction by Using Finite Element Analysis in Reactor Pressure Vessel of PWR

2018 ◽  
Author(s):  
Koji Maeta ◽  
Keisuke Matsuyama ◽  
Hirokazu Sugiura ◽  
Shigeyuki Watanabe ◽  
Hideyuki Morita ◽  
...  
Keyword(s):  
Structural Integrity ◽  
Acoustic Analysis ◽  
Structural Response ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Pressure Pulsations ◽  
Element Analysis ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Pressure Distributions

The reactor coolant pump (RCP) in a pressurized water reactor (PWR) plant generates pressure pulsations at multiple frequencies. These pressure pulsations excite the acoustic modes inside the reactor vessel (RV), and cause significant acoustic loads on the reactor internals (RIs). For verifying the structural integrity of the RIs, it is important to predict the acoustic loads, which is used for vibration analysis of the RIs. Traditionally, an analytical method, assuming that structures are rigid, has been used in order to predict the acoustic pressure distributions inside the RV [1]. However, water in actual PWR plant is heavy enough to influence structural response, so that it is required to use methods for a coupled structural acoustic system. In this article, the coupled structural-acoustic analysis using the commercial software ANSYS is proposed in order to predict the acoustic loads, and the applicability of this method is discussed. The structural-acoustic interactions inside the RV are investigated by element tests and the scale model test. The acoustic pressures measured by these tests are compared with the calculated results.

Predictive Study of CAP1400 Core Barrel Vibration: Part 2 — Structural Responses

2017 ◽  
Author(s):  
Li Yuan ◽  
Zhang Wei ◽  
Zhang Ming ◽  
Yu Qing
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Response ◽  
Mass Effect ◽  
Scale Model ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Pressurized Water ◽  
Structural Responses ◽  
Predictive Study

As described in Part 1 of this paper, [1], CAP1400 is a 1400 MWe pressurized water reactor developed by SNERDI to be the next series of nuclear power plants in China. As a part of the feasibility study, a 1/6 scale model test, conducted in Chengdo, China, of the CAP1400 pressure vessel and its internals was carried out to study the flow induced vibration (FIV) characteristics. This paper describes the predictive study on the structural responses of the core barrel vibration with particular emphasis on using the “Fourier node” method in modeling the hydrodynamic mass effect. It is noted that this is the second part of a two-part series and is focusing on the structural response calculation using the forcing functions described in Part 1, [1], and the comparison with the measured data.

Experimental Research on the Fluid Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2018 ◽  
Author(s):  
Rui Xu ◽  
Yaoyu Hu ◽  
Yun Long ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Canned Motor

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.

Applicability of Fracture Toughness Curves Developed for Japanese Pressure Vessel Steels to Structural Integrity Evaluation

2015 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Compact Tension ◽  
Transition Curve ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Reactor Pressure

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014. In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.

Dynamic Assessment of the Core Barrel During Loss of Coolant Accident

2018 ◽  
Author(s):  
Yaroslav Dubyk ◽  
Vladislav Filonov ◽  
Oleksii Ishchenko ◽  
Igor Orynyak ◽  
Yuliia Filonova
Keyword(s):  
Dynamic Assessment ◽  
Pressurized Water Reactor ◽  
Acoustic Response ◽  
Element Analysis ◽  
Pressurized Water ◽  
The Core ◽  
Loss Of Coolant Accident ◽  
The Finite Element Analysis ◽  
Loss Of Coolant ◽  
Duhamel Integral

This article focuses on the dynamic behavior of the Pressurized Water Reactor (PWR) during the Loss Of Coolant Accident (LOCA) which cause the significant acoustic loads on the Core Shrouds. The finite element analysis of a PWR was performed to obtain the acoustic response to the LOCA event. We have performed dynamic stress and strain calculations in the frequency domain for the Core Barrel, according to classical shell theories. The Duhamel integral was used to calculate the transient response of a shell to the transient load caused by the water hammer event. The results obtained were used for fracture mechanics evaluations for flaws, which may occur between inservice inspections.

Effect of Uneven Crack Front on Crack Tip Mechanics and the Implication to Stress Corrosion Crack Growth

2009 ◽  
Author(s):  
He Xue ◽  
Zhanpeng Lu ◽  
Hiroyoshi Murakami ◽  
Tetsuo Shoji
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Crack Front ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Nickel Base

Uneven crack fronts have been observed in laboratory stress corrosion cracking tests. For example, cracking fronts of nickel-base alloys tested in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments could exhibit uneven crack front. Analyzing the effect of an uneven crack front on further crack growth is important for quantification of crack growth. Finite-Element analysis shows that the local KI distribution can be significantly affected by the shape and size of the uneven crack front. Stress intensity factor at the locally extended crack front can be significantly reduced. Since generally there is a nonlinear CGR versus KI relationship, it is expected that crack growth rate at the locally extended crack front can be significantly different from those in the neighboring areas. There could be several patterns for the growth of an uneven crack front. For example, once initiated, the crack growth rate in areas other than the locally protruded front would become higher and then the whole crack front would tend to become uniform. On the other hand, if the crack growth in other areas is still low, there is a possibility that the crack growth rate at the front tip would slow down.

Mechanistic Formulation of PWSCC Growth Rates of Ni-Base Alloys and Weld Metals

2011 ◽  
Author(s):  
Zhanpeng Lu ◽  
Tetsuo Shoji ◽  
He Xue ◽  
Chaoyang Fu
Keyword(s):  
Thermal Activation ◽  
Structural Integrity ◽  
Thermal Activation Energy ◽  
Evaluation Procedure ◽  
Alloy 600 ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Weld Metals ◽  
Primary Water ◽  
Integrity Analysis

Several Ni-base alloys and their weld metals such as Alloy 600 and Alloy 82/182 suffered from stress corrosion cracking in pressurized water reactor primary water environments. Materials Reliability Program (MRP) proposed a CGR disposition curve in a report MRP 55 for PWSCC of thick-section Alloy 600 materials. This deterministic CGR equation has been adopted by Section XI Nonmandatory Appendix O of the ASME Boiler and Pressure Code for flaw evaluation. MRP also proposed a CGR disposition curve in MRP report 115 for PWSCC of Alloy 82/182/132 weld metals. In the same fashion, JSME and JNES also provided CGR disposition curves in the flaw evaluation procedure in structural integrity analysis. Stress intensity factor (K), temperature and thermal activation energy are included in both MRP 55 and MRP 115 reports. Both MRP 55 and MRP 115 are engineering-based rather than mechanism-based. The fundamental correlations such as crack growth rate vs. K are quantified based on the theoretical model and screened experimental data, which are compared to the reported disposition curves and used for improving the prediction.

Development of Structural Integrity Assessment Method for Flow-Induced Vibration of Reactor Internals in PWR

2018 ◽  
Author(s):  
Hirokazu Sugiura ◽  
Shigeyuki Watanabe ◽  
Akihisa Iwasaki ◽  
Hideyuki Morita ◽  
Hideyuki Sakata ◽  
...  
Keyword(s):  
Model Test ◽  
Structural Integrity ◽  
Pressure Fluctuations ◽  
Vibration Response ◽  
Scale Model ◽  
Empirical Equations ◽  
Cfd Analysis ◽  
Pressurized Water ◽  
Forcing Function ◽  
Reactor Internals

For verifying the structural integrity of the reactor internals (RIs) in a pressurized water reactor (PWR) plant, it is important to estimate the vibration response of the core barrel (CB) due to flow turbulence. Instead of scale model test, the computational fluid dynamics (CFD) has been expected as a method to predict the turbulence forcing function for the response analysis of the CB. In this article, a hybrid approach combining empirical equations based on flow test and CFD analysis is proposed in order to predict the turbulence forcing function. The scale model test of new RIs, which were developed by Mitsubishi Heavy Industries, Ltd., was conducted, and the pressure fluctuations for the turbulence forcing function and the vibration response of the CB were measured. The pressure fluctuations were calculated by CFD analysis, and the vibration analysis using the turbulence forcing function determined from the calculated pressure fluctuations was performed. This article provides the scale model test data and the empirical equations of the turbulence forcing functions, and validation results of the proposed method to predict the turbulence forcing function using CFD.

Analysis of Residual Stresses in a Pressurizer Surge–Nozzle Weldment

2011 ◽  
Author(s):  
Florian Obermeier ◽  
Tomas Nicak ◽  
Gottfried Meier ◽  
Elisabeth Keim
Keyword(s):  
Residual Stresses ◽  
Material Model ◽  
Round Robin ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Piping Systems ◽  
Primary Water ◽  
Abaqus Code ◽  
Transient Thermal

Residual stresses and distortion of welded structures have a significant influence on their functionality and their lifetime. In Pressurized Water Reactor (PWR) piping systems, residual stresses in dissimilar metal welds extensively increase their susceptibility to primary water stress corrosion cracking (PWSCC). An accurate crack initiation and growth assessment is essential to assure that no severe component failure will occur due to SCC or any other stress induced mechanism. Therefore it is necessary to develop methods for a proper residual stress and distortion prediction. Numerical welding simulations have developed fast during the recent years but no universally accepted guidelines for a reliable prediction of residual stresses have been established so far. In order to support further validation of developed methods the U.S. NRC launched an international round robin program. Its main intention is to benchmark different numerical approaches by direct comparison with experimental data obtained from a provided mock-up. This paper shows the contribution of AREVA NP Germany to this round robin. A two dimensional axis-symmetric analysis of a pressurizer surge nozzle – safe end – pipe weldment will be presented. The residual stresses are calculated by means of an uncoupled transient thermal and mechanical finite element analysis using the ABAQUS Code. In addition some comments on the selection of the material model and its effect on the resulting residual stresses will be given.

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