Mechanical Characterization for a Large Test Design of a Dissimilar Metals Welding With a Narrow Gap Nickel Alloy Weld: Experimental and Numerical Analysis on Specimens

2012 ◽  
Author(s):  
Myriam Bourgeois ◽  
Olivier Ancelet ◽  
Stéphane Marie ◽  
Stephane Chapuliot
Keyword(s):  
Nickel Alloy ◽  
Structural Integrity ◽  
Atomic Energy Commission ◽  
Dissimilar Metals ◽  
Narrow Gap ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Piping Systems ◽  
Water Reactors

Dissimilar metal welds are a common feature of light water reactors in connections between ferritic components and austenitic stainless steel piping systems. Inspection difficulties, variability of material properties and residual stresses all combine to create problems for structural integrity assessment. Within the framework of European project STYLE, a fracture test on a pipe containing a through wall crack in a narrow gap Nickel alloy Dissimilar Metals (DMWinc) is under preparation. The work is focusing on the nickel alloy - ferrite steel interface which is the weakest area of such welded pipes in front of ductile tearing. The study temperature is 300°C, which covers typical temperatures in components like hot pipes in the primary coolant system of pressurized water reactors. This paper gives an overview of the program and the first results of works which is been carried out by the French Atomic Energy Commission and Alternative Energies (CEA) in order to study the mechanical properties and integrity of component of the DMWinc pipes provided and designed by AREVA France.

The Integral PWR SIR Transients: Comparisons Between CATHARE and RELAP Codes

2002 ◽  
Author(s):  
Jean-Franc¸ois Pignatel
Keyword(s):  
Natural Circulation ◽  
Atomic Energy Commission ◽  
Light Water Reactors ◽  
Pressurized Water Reactor ◽  
Large Area ◽  
Primary Coolant ◽  
Pressurized Water ◽  
The Subject ◽  
Water Reactors ◽  
Predictive Study

Within the framework of the research program on innovative light water reactors, the SERI (Service of Studies on Innovative Reactors) of the French Atomic Energy Commission (CEA), is presenting a predictive study on the modeling of a low-power integral Pressurized Water Reactor, using the CATHARE thermalhydraulic code. The concept selected for this study is that of the SIR reactor project, developed by AEA-T and ABB consortium. This very interesting concept is no doubt that which is the most complete to this date, and on which most information in the literature can be obtained. Many safety calculations made with the RELAP code are also available and represent a highly interesting base for comparison purposes, in order to improve the approach on the results obtained with CATHARE. A comparison of the behavior of the two codes is thus presented in this article. This study therefore shows that CATHARE finely models this type of new PWR concept. The transients studied cover a large area, ranging from natural circulation to loss of primary coolant accidents. The ATWS and a power transient have also been calculated. The comparison made between the CATHARE and RELAP results shows a very good agreement between the two codes, and leads to a very positive conclusion on the pertinence of simulating an integral PWR. Moreover, even though this study is a thorough investigation on the subject, it confirms the potentially safe nature of the SIR reactor.

scholarly journals Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

2015 ◽  
Vol 19 (3) ◽  
pp. 989-1004 ◽  
Author(s):  
Ezddin Hutli ◽  
Valer Gottlasz ◽  
Dániel Tar ◽  
Gyorgy Ezsol ◽  
Gabor Baranyai
Keyword(s):  
Heat Transfer ◽  
Structural Integrity ◽  
Complex Geometry ◽  
Measurement Techniques ◽  
Coolant Flow ◽  
Temperature Fields ◽  
Head Part ◽  
Pressurized Water ◽  
Area Of Interest ◽  
Water Reactors

The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs). Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively) are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel). The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

Thermal Fatigue of Pwr Coolant System Loop Drain Lines Due to Outflow Activities

2021 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Thermal Transients ◽  
Piping Systems ◽  
Class 1 ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. In addition, cumulative effects from other thermal transients, such as outflow activities, may also contribute to the failure of the RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Integrity Assessment of Pressure Components With Local Hot Spots

2005 ◽  
Vol 127 (2) ◽  
pp. 137-142 ◽  
Author(s):  
R. Seshadri
Keyword(s):  
Hot Spots ◽  
Structural Integrity ◽  
Refractory Lining ◽  
Assessment Method ◽  
Pressure Vessels ◽  
Industrial Processes ◽  
Integrity Assessment ◽  
Piping Systems ◽  
Further Development ◽  
Level 2

Local hot spots can occur in some pressure vessels and piping systems used in industrial processes. The hot spots could be a result of, for instance, localized loss of refractory lining on the inside of pressure components or due to a maldistribution of process flow within vessels containing catalysts. The consequences of these hot spots on the structural integrity of pressure components are of considerable importance to plant operators. The paper addresses structural integrity issues in the context of codes and standards design framework. Interaction of hot spots, as is the case when multiple hot spots occur, is addressed. An assessment method, suitable for further development of a Level 2 “Fitness-for-Service” methodology, is discussed and applied to a commonly used pressure component configuration.

Evaluation of Residual Stresses in Dissimilar Weld Joints

2011 ◽  
Vol 681 ◽  
pp. 182-187 ◽  
Author(s):  
Alix Bonaventur ◽  
Danièle Ayrault ◽  
Guillaume Montay ◽  
Vincent Klosek
Keyword(s):  
Residual Stresses ◽  
Weld Joint ◽  
Safety Issue ◽  
Austenitic Steels ◽  
Hole Drilling ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Dissimilar Weld ◽  
Dissimilar Welds ◽  
Water Reactors

Dissimilar metal joints between pipes of ferritic and austenitic steels are present in primary coolant circuit of pressurized water reactors. Over the last years in particular in USA and Japan, stress corrosion cracks, often associated with weld repairs, have been observed for some dissimilar welds made with an Inconel filler metal. The integrity of this type of components is thus a major safety issue. In this context, the goal of this work is to evaluate the welding residual stresses field for a dissimilar weld joint. A representative bi-metallic tubular weld joint was fabricated and residual stresses profiles in the different weld zones were evaluated by means of the hole drilling and neutron diffraction methods.

Modeling of Porous Crud Layer Based on Effects of Diffusion and Water Radiolysis on Anomalous Porous Crud Deposition on Fuel Pin Surfaces in PWRs

2002 ◽  
Author(s):  
Hyun-Jong Joe ◽  
Barclay G. Jones
Keyword(s):  
Negative Impact ◽  
Nucleate Boiling ◽  
Water Radiolysis ◽  
Trace Impurities ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Fuel Pin ◽  
Negative Effect ◽  
Water Reactors ◽  
Coolant System

Many studies have been undertaken to understand crud formation on the upper spans of fuel pin clad surfaces, which is called axial offset anomaly (AOA), is observed in pressurized water reactors (PWR) as a result of sub-cooled nucleate boiling. Separately, researchers have considered the effect of water radiolysis in the primary coolant of PWR. This study examines the effects of radiolysis of liquid water, which aggressively participate in general cladding corrosion and solutes within the primary coolant system, in the terms of pH, temperature, and Linear Energy Transfer (LET). It also discusses the effect of mass transfer, especially diffusion, on the concentration distribution of the radiolytic products, H2 and O2, in the porous crud layer. Finally it covers the effects of chemical reactions of boric acid (H3BO3), which has a negative impact on the mechanisms of water recombination with hydrogen, lithium hydroxide (LiOH), which has a negative effect on water decomposition, dissolved hydrogen (DH), and some trace impurities.

The Sensitivity of Risk-Informed Reactor Structural Integrity Analysis Results to Various Interpretations of Warm Pre-Stress

2009 ◽  
Author(s):  
Mark T. EricksonKirk ◽  
Terry L. Dickson
Keyword(s):  
Structural Integrity ◽  
The United States ◽  
Integrated Analysis ◽  
Total Risk ◽  
Pressurized Water ◽  
Best Estimate ◽  
Water Reactors ◽  
Using Data ◽  
Accident Scenarios ◽  
Reactor Operating

Warm pre-stress, or WPS, is a phenomenon by which the apparent fracture toughness of ferritic steel can be elevated in the fracture mode transition if crack is first “pre-stressed” at an elevated temperature. Taking proper account of WPS is important to the accurate modeling of the postulated accident scenarios that, collectively, are referred to as pressurized thermal shock, and to the accurate modeling of routine cool-down transients. For both accident and routine cool-downs the transients begin at the reactor operating temperature (approximately 290°C for pressurized water reactors in the United States) and proceed to colder temperatures as time advances. The probabilistic fracture mechanics code FAVOR, which is being used by the NRC to provide the technical basis for risk-informed revisions of 10 CFR 50.61 and 10 CFR 50 Appendix G, adopts a model of WPS as part of its fracture driving force module. In this paper we assess the conservatism inherent to the FAVOR WPS model relative to a best-estimate WPS model constructed using data recently produced by the European Commission “SMILE” project and published by Moinereau and colleagues. Assessments of the conservatisms inherent to the so-called “conservative principle” WPS model, and also to a classic LEFM model that does not credit WPS are also made. The data presented herein demonstrate that, for an integrated analysis of PTS risk, the FAVOR and conservative principle WPS models both over-estimate the vessel failure risk by a factor of between 2 and 3× relative to the best estimate model. Our examination of the effect of WPS models on the predictions of individual transients reveals that for the severe transients that dominate risk there is little difference (usually less than 4×) between the conditional probabilities of crack initiation and of through wall cracking predicted by the different WPS models. There are considerable differences in the predictions of the various WPS and non-WPS models for low severity transients, however, the contribution of these transients to the total risk of vessel failure is small.

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