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.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Line Down Horizontal Elbow

2020 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Reactor Coolant ◽  
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. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from various thermal transients such as the reactor coolant system (RCS) sampling and excess letdown may also contribute to the failure of 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 Section III Class 1 piping stress formula, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Recent Operational Experience of Pressurized Water Reactor Safety Injection and Drain Line Cracking and Supporting Flaw Evaluations

2019 ◽  
Author(s):  
Greg Imbrogno ◽  
Stephen Marlette ◽  
Alexandria Carolan ◽  
Anees Udyawar ◽  
Mark Gray
Keyword(s):  
Nuclear Power ◽  
Operating Experience ◽  
Piping System ◽  
Operational Experience ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Piping Systems ◽  
Class 1 ◽  
Asme Code

Abstract A recent increase in operating experience (OE) related to pipe cracking in non-isolable auxiliary piping systems has been realized in the Pressurized Water Reactor (PWR) nuclear power industry. The majority of PWR auxiliary piping systems are comprised of welded stainless steel pipe and piping components. The susceptible piping systems are Class 1 pressure boundary and typically non-isolable from the primary loop. Since they are non-isolable, when a pipe crack or crack indication is identified, an emergent flaw evaluation and/or repair is required. Typically, the evaluations begin with an ASME Section XI IWB-3640 flaw evaluation to determine acceptability of the as-found flaw at the time of shutdown. Subsequent flaw evaluations are performed to demonstrate the possibility of continued operation of the piping component by leaving the flaw as-is without repair. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth evaluations, and the detection capabilities of the non-destructive examination technique. If evaluation of the as-found indication does not produce acceptable results, then a repair/replacement activity per ASME Section XI is considered. Possible repair scenarios include replacement of the piping section or component, or structural weld overlay. The results of the flaw evaluations or repairs must ensure that the auxiliary piping system will continue to operate safely. This paper will discuss the recent experiences of two different piping systems (boron injection tank line and drain line) that experienced cracking, the potential causes for the cracking in the absence of evidence, and the ASME Code flaw evaluations and/or repairs performed to support continued operation of the plant.

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

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.

Method for Predicting Mixing Tee Thermal Fatigue in Carbon Steel Based on Austenitic Stainless Steel Operating Experience

2009 ◽  
Author(s):  
Robert O. McGill ◽  
David O. Harris ◽  
Ken Wolfe
Keyword(s):  
Stainless Steel ◽  
Carbon Steel ◽  
Austenitic Stainless Steel ◽  
Thermal Fatigue ◽  
Operating Experience ◽  
The United States ◽  
Pressurized Water ◽  
Short Period ◽  
Water Reactors ◽  
Steel Piping

Over the past ten years, a significant amount of research has been conducted regarding mixing tee thermal fatigue in pressurized water reactors prompted by the leakage event at Civaux Unit 1 in 1998. The plant experienced a leak in the reactor residual heat removal system piping after a short period of operation during plant start-up. An evaluation as to the cause of the leakage concluded that mixing of hot and cold fluid upstream from the failed austenitic stainless steel elbow resulted in thermal fatigue cracking. Recently, an assessment of susceptibility to this thermal fatigue mechanism in boiling water reactors in the United States was completed. The piping systems in these reactors where the potential for thermal mixing exists are predominantly constructed of carbon steel. Thus, an analytical model was developed for predicting mixing tee thermal fatigue in carbon steel piping based on the austenitic stainless steel piping operating experience at Civaux. This paper describes how the model was developed and presents some general findings.

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.

scholarly journals Operating experience of desalination unit coupled to primary coolant system of CIRUS

2010 ◽  
Vol 4 (2) ◽  
pp. 134 ◽  
Author(s):  
Rakesh Ranjan ◽  
Surinder Kumar ◽  
D. Mathur ◽  
N. Ramesh ◽  
R.C. Sharma
Keyword(s):  
Operating Experience ◽  
Primary Coolant ◽  
Coolant System
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