scholarly journals Comparative Evaluation of Plastic Design Methods for Fatigue Assessment of a Nuclear Class 1 Piping Nozzle

2020 ◽  
Author(s):  
David M. Clarkson ◽  
Christopher D. Bell ◽  
Donald Mackenzie
Keyword(s):  
Pressure Vessels ◽  
Design Methods ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Technical Literature ◽  
Pressurized Water ◽  
Plastic Design ◽  
Class 1 ◽  
The Cost ◽  
Elastic Design

Abstract Design-by-analysis (DBA) procedures for Nuclear Class 1 pressure vessels such as those prescribed within ASME Boiler and Pressure Vessel Code (BPVC) Section III, provide rules to demonstrate assurance against fatigue failure. Two general assessment routes exist, linear finite element analysis (FEA) with stress categorization and elastic-plastic penalty factors, or nonlinear FEA with direct multiaxial strain evaluation. Whilst the elastic design route possesses many practical advantages, it is widely acknowledged to be very conservative, sometimes unacceptably so. At the cost of additional analysis effort, plastic design methods can provide a more appropriate evaluation of fatigue usage, potentially avoiding unnecessary design modifications and reducing the burden of in-service inspection requirements. This paper presents and compares various strain measures proposed for ASME III plastic fatigue analysis within the technical literature. A case study of a typical pressurized water reactor (PWR) main coolant line (MCL) piping nozzle subjected to pressure and thermal loads is presented. The influence of strain measure selection on the FE-derived strain concentration (Ke) factors is examined. Some important considerations for calculation of realistic Ke factors in ASME III are further discussed.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

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.

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.

Deterministic and Probabilistic Fracture Mechanics Analysis for Structural Integrity Assessment of Pressurized Water Reactor Pressure Vessel

2016 ◽  
Vol 138 (3) ◽  
Author(s):  
Kuan-Rong Huang ◽  
Chin-Cheng Huang ◽  
Hsiung-Wei Chou
Keyword(s):  
Structural Reliability ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Benchmark Problems ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Oak Ridge ◽  
Integrity Assessment ◽  
Reactor Pressure

Cumulative radiation embrittlement is one of the main causes for the degradation of pressurized water reactor (PWR) reactor pressure vessels (RPVs) over their long-term operations. To assess structural reliability of degraded reactor vessels, the FAVOR code from the Oak Ridge National Laboratories of the U.S. is employed to perform probabilistic fracture analysis for existing Taiwan domestic PWR reactor vessels with consideration of irradiation embrittlement effects. The plant specific parameters of the analyzed reactor vessel associated with assumed design transients are both considered as the load conditions in this work. Furthermore, two overcooling transients of steam generator tube rupture (SGTR) and pressurized thermal shock (PTS) addressed in the USNRC/EPRI benchmark problems are also taken into account. The computed low failure probabilities can demonstrate the structural reliability of the analyzed reactor vessel for its both license base and extended operations. This work is helpful for the risk assessment and aging management of operating PWR RPVs and can also be referred as its regulatory basis in Taiwan.

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.

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.

Probabilistic Structural Integrity Evaluation on a Pressurized Water Reactor Pressure Vessel Under Pressure–Temperature Limit Operations

2015 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

The normal reactor startup (heat-up) and shut-down (cool-down) operation limits are defined by the ASME Code Section XI-Appendix G, to ensure the structural integrity of the embrittled nuclear reactor pressure vessels (RPVs). In the paper, the failure risks of a Taiwan domestic pressurized water reactor (PWR) pressure vessel under various pressure-temperature limit operations are analyzed. Three types of pressure-temperature limit curves established by different methodologies, which are the current operation limits of the domestic RPV based on the KIa fracture toughness curve in 1998 or earlier editions of ASME Section XI-Appendix G, the recently proposed limits according to the KIC fracture toughness curve after the 2001 edition of ASME Section XI-Appendix G, and the risk-informed revision method proposed in MRP-250 report that provides more operational flexibility, are considered. The ORNL’s probabilistic fracture mechanics code, FAVOR, is employed to perform a series of fracture probability analyses for the RPV at multiple levels of embrittlement under such pressure-temperature limit transients. The analysis results indicate that the pressure-temperature operation limits associated with more operational flexibility will result in higher failure risks to the RPV. The shallow inner surface breaking flaw due to the clad fabrication defect is the most critical factor and dominates the failure risk of the RPV under pressure-temperature limit operations. Present work can provide a risk-informed reference for the safe operation and regulation of PWRs in Taiwan.

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