Fatigue Effect of RCS Branch Line by Thermal Stratification

2003 ◽  
Author(s):  
Hag-Ki Youm ◽  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Tea-Eun Jin ◽  
Sun-Ki Lee ◽  
...  
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Methodology ◽  
Element Analysis ◽  
Branch Line ◽  
Reactor Coolant ◽  
Stratification Effect ◽  
Fatigue Evaluation ◽  
Coolant System

Recent events reported at a number of nuclear power plants worldwide have shown that thermal stratification, cycling, and striping in piping can cause excessive thermal stress and fatigue on the piping material. These phenomena are diverse and complicated because of the wide variety of geometry and thermal hydraulic conditions encountered in reactor coolant system. Thermal stratification effect of re-branched lines is not yet considered in the fatigue evaluation. To evaluate the thermal load due to turbulent penetration, this paper presents a fatigue evaluation methodology for a branch line of reactor coolant system with the re-branch line. The locations of fatigue monitoring and supplemented inspections are discussed as a result of fatigue evaluations by Interim Fatigue Management Guideline (ITFMG) and detail finite element analysis. Although the revised CUF was increased less than 50 %, the CUF values for some locations was greater than the ASME Code limits.

Fatigue Evaluation of RCS Branch Line for License Renewal Application

2006 ◽  
Author(s):  
Il-Kwun Nam ◽  
Hagki Youm ◽  
Tae Eun Jin ◽  
Byung Sup Kim
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Peak Stress ◽  
Stress Intensity Range ◽  
Branch Line ◽  
Fatigue Evaluation ◽  
License Renewal ◽  
Safety Enhancement

The fatigue effects have to be revaluated in preparing the license renewal application for the continued operation of an old vintage nuclear power plant. This paper presents a complete fatigue analysis for a branch piping with the effect of thermal stratification, induced by turbulent penetration, and environmental factors on fatigue. Three-dimensional computational fluid dynamics and finite element analyses were performed for the branch line to evaluate the thermal stratification loading. Proposed is a supplementary methodology of considering the effect of environmental factor on the combined conventional peak stress intensity range, based on the NB-3600 of ASME Section III Code, with thermal stratification loading. It can be used for safety enhancement of old vintage nuclear power plants.

Assessment of a Piping System Subjected to Thermal Stratification and Thermal Striping

2010 ◽  
Author(s):  
Somnath Chattopadhyay
Keyword(s):  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Stresses ◽  
Temperature Gradients ◽  
Low Flow ◽  
Piping System ◽  
Thermal Transients ◽  
Fatigue Evaluation ◽  
Thermal Striping

Piping systems in nuclear power plants are often designed for pressure and mechanical loadings (including seismic loads) and operating thermal transients. In the last few decades a number of failures have occurred due to thermal stratification caused by the mixing of hot and cold fluids under certain low flow conditions. Such stratified temperature fluid profiles give rise to circumferential metal temperature gradients through the pipe leading to high stresses causing fatigue damage. In this work, thermal stresses due to such temperature gradients have been calculated using a finite element method. The peak stresses calculated by this method has been used for fatigue evaluation. In addition the stresses due to thermal striping associated with stratification have also been independently assessed for high cycle fatigue. The method outlined in this paper is a simplified conservative procedure to obtain stratification stresses.

A Study on Thermal Stress Evaluation of Safety Injection System for NPPs Considering Leakage Effect

2011 ◽  
Author(s):  
Hee-Dong Sung ◽  
Sun-Hye Kim ◽  
Ik-Joong Kim ◽  
Young-Jin Kim ◽  
Jeong-Soon Park ◽  
...  
Keyword(s):  
Thermal Stress ◽  
Thermal Stratification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Injection System ◽  
Stress Evaluation ◽  
Branch Line ◽  
Leakage Model ◽  
Cfd Analyses

Several piping failures caused by thermal stratification have been reported in some nuclear power plants since the early 1980s. However, this kind of thermal effect was not considered when the old vintage nuclear power plants were designed. Thermal stratification is usually generated by turbulent penetration from the RCS to branch line or leakage through damaged part of valve in branch line. In this paper, using the CFD analysis, characteristics of thermal stratification in a safety injection system of PWR plant were investigated and thermal stress evaluation was also conducted. First, CFD analyses were carried out on in-leakage model and out-leakage model according to operating condition. The case of out-leakage, the thermal stratification based on temperature distribution was generated a little at the rear of 1st valve. In contrast, significant thermal stratification was generated in front of 1st valve in in-leakage model because the effect of rapid flow velocity from RCS.

Reactor Coolant System Branch Line Pipe Break Analysis for Shin-Kori Units 3 and 4

2010 ◽  
Author(s):  
Young Jin Byun ◽  
Young Sil Sul ◽  
Kwang Won Lee ◽  
Jong Tae Seo
Keyword(s):  
Power Plants ◽  
Power Reactor ◽  
Time History ◽  
Dynamic Responses ◽  
Lumped Mass ◽  
Line Pipe ◽  
Branch Line ◽  
Reactor Coolant ◽  
Main Steam ◽  
Coolant System

Shin-Kori Units 3&4 (SKN 3&4) are the first APR1400 (Advanced Power Reactor 1400 MWe) nuclear power plants being under construction in Korea. The capacity and size of main components in the APR1400 nuclear steam supply system have been increased from the OPR1000 (Optimized Power Reactor 1000 MWe) which is the reference plant in operation. Especially, the increased pipe size of the main steam and the economizer feedwater lines cause more severe excitations such as the nozzle thrust, jet impingement, and nozzle reactions when those lines are postulated to break at terminal ends or intermediate locations. In this paper, the major mechanical effects on the Reactor Coolant System (RCS) due to the branch line pipe breaks for SKN 3&4 are investigated; the support loads of the reactor vessel, steam generators, reactor coolant pumps are evaluated. In this investigation, the RCS is modeled using 3-dimensional lumped mass and beam elements, and the nonlinear time-history structural analyses are performed. The analysis results show that the RCS dynamic responses are governed by the main steam line breaks and feedwater economizer line breaks.

scholarly journals Facts & Fission

2009 ◽  
Vol 131 (12) ◽  
pp. 32-37
Author(s):  
Theodore Rockwell
Keyword(s):  
Real World ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactors ◽  
Laws Of Nature ◽  
Spent Fuel ◽  
Reactor Coolant ◽  
System A ◽  
Chernobyl Reactor ◽  
Coolant System

This article discusses the real-world practicality of nuclear power. Neither reactors nor casks of spent fuel have the capability of going “prompt critical” like a bomb. The laws of nature prohibit it and engineers must make clear that facts of nature are not matters of opinion. The Chernobyl reactor, which failed so spectacularly in Ukraine in 1986, became for many a symbol of the dangers represented by nuclear reactors. But that is not warranted; such an accident is simply impossible for the kind of commercial reactors now being planned or built. From a public safety standpoint, the most important feature of our current reactors is that, after any event that ruptures the reactor coolant system, a large amount of water and steam would be violently swirling around inside the containment structure, even if containment structure has been ruptured. In the emerging context of realistically reexamining many long-held assumptions, engineers will find opportunities to drastically improve the way nuclear power plants are built and operated.

Sign in / Sign up

Export Citation Format

Share Document