Fatigue Evaluation of RCS Branch Line for License Renewal Application

Author(s):  
Il-Kwun Nam ◽  
Hagki Youm ◽  
Tae Eun Jin ◽  
Byung Sup Kim

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.

Author(s):  
Hag-Ki Youm ◽  
Kwang-Chu Kim ◽  
Man-Heung Park ◽  
Tea-Eun Jin ◽  
Sun-Ki Lee ◽  
...  

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.


Author(s):  
Somnath Chattopadhyay

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.


Author(s):  
Timothy Gilman ◽  
Jay Gillis ◽  
Jagannath Hiremagalur ◽  
Scott Rodamaker ◽  
William Weitze ◽  
...  

This paper describes the techniques utilized to perform a nonlinear, strain-based, environmentally-assisted fatigue evaluation of a pressurizer vessel in a nuclear power plant. Significant differences between the strain-based and more traditional fatigue analysis results are demonstrated. This paper concludes that leveraging today’s computer power with the use of more detailed, nonlinear analysis is an effective tool for nuclear power plants to meet license renewal commitments related to the management of environmentally-assisted fatigue.


Author(s):  
Hee-Dong Sung ◽  
Sun-Hye Kim ◽  
Ik-Joong Kim ◽  
Young-Jin Kim ◽  
Jeong-Soon Park ◽  
...  

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.


Author(s):  
Hwan Ho Lee ◽  
Joon Ho Lee ◽  
Dong Jae Lee ◽  
Seok Hwan Hur ◽  
Il Kwun Nam ◽  
...  

A numerical analysis has been performed to estimate the effect of thermal stratification in the safety injection piping system. The Direct Vessel Injection (DVI) system is used to perform the functions of Emergency Core Cooling and Residual Heat Removal for an APR1400 nuclear power plant (Korea’s Advanced Power Reactor 1400 MW-Class). The thermal stratification is anticipated in the horizontally routed piping between the DVI nozzle of the reactor vessel and the first isolation valve. Non-axisymmetric temperature distribution across the pipe diameter induced by the thermal stratification leads to differential thermal growth of the piping causing the global bending stress and local stress. Thermal hydraulic analysis has been performed to determine the temperature distribution in the DVI piping due to the thermal stratification. Piping stress analysis has also been carried out to evaluate the integrity of the DVI piping using the thermal hydraulic analysis results. This paper provides a methodology for calculating the global bending stresses and local stresses induced by the thermal stratification in the DVI piping and for performing fatigue evaluation based on Subsection NB-3600 of ASME Section III.


2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Benan Cai ◽  
Qi Zhang ◽  
Yu Weng ◽  
Hongfang Gu ◽  
Haijun Wang

Abstract Pipelines such as the surge line and main pipe are easily subjected to thermal stratification and thermal fatigue as a result of the nonuniform temperature distribution in the nuclear power plants. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen due to the existence of pipeline crack. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The flash evaporation related to pipe leak was experimentally studied in the paper. The experiment was carried out under high temperature and high pressure with low spray rate. The temperature and relative humidity (T&H) variations over time were monitored in the experiment with installing T&H detectors. The T&H variations at different measurement positions and with different spray rates were analyzed, respectively. In addition, the effect of the dimensionless parameters including the Weber number and Jakob number was also investigated. Results indicated that the response speed increased with the increase of the spray flow rate. Higher Weber number and higher Jakob number led to higher evaporation rate. The slight pipe leakage can be predicted by using the (T&H) in the hazardous areas.


Author(s):  
Akemi Nishida

It is becoming important to carry out detailed modeling procedures and analyses to better understand the actual phenomena. Because some accidents caused by high-frequency vibrations of piping have been recently reported, the clarification of the dynamic behavior of the piping structure during operation is imperative in order to avoid such accidents. The aim of our research is to develop detailed analysis tools and to determine the dynamic behavior of piping systems in nuclear power plants, which are complicated assemblages of different parts. In this study, a three-dimensional dynamic frame analysis tool for wave propagation analysis is developed by using the spectral element method (SEM) based on the Timoshenko beam theory. Further, a multi-connected structure is analyzed and compared with the experimental results. Consequently, the applicability of the SEM is shown.


Author(s):  
Omid Malekzadeh ◽  
Matthew Monid ◽  
Michael Huang

Abstract Three-Dimensional (3D) CAD models are utilized by many designers; however, they are rarely utilized to their full potential. The current mainstream method of design process and communication is through design documentation. They are limited in depth of information, compartmentalized by discipline, fragmented into various segments, communicated through numerous layers, and finally, printed onto an undersized paper by the stakeholders and end-users. Large nuclear projects, such as refurbishments and decommissioning, suffer from spatial, interface, and interreference challenges, unintentional cost and schedule overruns, and quality concerns that can be rooted to the misalignments between designed and in-situ or previously as-built conditions that tend to stem from inaccessibility and lack of adequate data resolution during the transfer of technical information. This paper will identify the technologies and the methodology used during several piping system modifications of existing nuclear power plants, and shares the lessons learned with respect to the benefits and shortcomings of the approach. Overall, it is beneficial to leverage available multi-dimensional technologies to enhance various engineering and execution phases. The utilization and superposition of various spatial models into 3D and 4D formats, enabled the modification projects to significantly reduce in-person plant walkdown efforts, provide highly accurate as-found data, and enable stakeholders of all disciplines and trades to review the as-found, as-designed, and simulated as-installed modification; including the steps in between without requiring significant plant visits. This approach will therefore reduce the field-initiated changes that tend to result in design/field variations; resulting in less reliance on Appendix T of ASME BPVC Section III, reduction in the design registration reconciliations efforts, and it aligns with the overarching goal of EPRI guideline NCIG-05. Beyond the benefits to design and execution, the multidimensional approach will provide highly accurate inputs to some of the nuclear safety’s Beyond Design Basis Assessments (BDBA) and allowed for the incorporation of actual design values as input and hence removing the unnecessary over-conservatisms within some of the inputs.


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