Proposal of a Guideline for Thickness Management of Wall Thinning T-Joints

2012 ◽  
Author(s):  
Mayumi Ochi ◽  
Katsuhiko Yamakami ◽  
Yoshinobu Hamaguchi ◽  
Katsumasa Miyazaki ◽  
Keita Naito ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Japan Society ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Study Results ◽  
Simulation Results ◽  
Design Pressure ◽  
Design Construction

Although the required minimum thickness (tsr) of T-joints (tees) of class 2, 3 and lower classes of nuclear power plants in Japan is calculated from the design pressure and temperature for design, there are neither any rules nor standards for thickness management of wall thinning T-joints for facilities maintenance. This paper describes additional parametric study results and proposes a guideline for thickness management of wall thinning T-joints. In other papers related to this project, the experiment and numerical simulation results are reported. This paper refers to these results and performs further investigation under the consideration of JSME (The Japan Society of Mechanical Engineers) design, construction and maintenance codes and standards.

Pressure Tests for Thickness Management of Wall Thinning Tees

2012 ◽  
Author(s):  
Kaina Teshima ◽  
Yoichi Iwamoto ◽  
Kiminobu Hojo ◽  
Tomoyuki Oka ◽  
Kunihiro Kobayashi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Outer Diameter ◽  
Test Results ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Pressure Tests ◽  
Design Pressure

Although the minimum thickness of pipe wall required (tsr) of T-joints (tees) of class 2, 3 and lower classes of nuclear power plants in Japan is calculated from the design pressure and temperature, there is no rule or standard of wall thinning T-joints for thickness management. This paper describes the pressure tests procedure and six test results with parameters of T-joint geometry such as outer diameter D, thickness T and T/D to establish structural integrity of wall thinning T-joints. Based on the fracture surface observation, a ductile crack initiation of each test mock-ups was confirmed.

The Proposal of Wall Thickness Management Criteria of T-Pipes

2012 ◽  
Author(s):  
Kiyoharu Tsunokawa ◽  
Taku Ohira ◽  
Naoki Miura ◽  
Yasumi Kitajima ◽  
Daisuke Yoshimura
Keyword(s):  
Numerical Simulation ◽  
Wall Thickness ◽  
Parametric Study ◽  
The Other ◽  
Wall Thinning ◽  
Study Results ◽  
Simulation Results ◽  
Design Construction

Although the reinforcement for openings is checked in accordance with design / construction standard when thinning was observed in T-pipes, this evaluation becomes too conservative or requires much time and effort. This paper describes additional parametric study results and proposes a guideline for thickness management of wall thinning T-pipes. On the other papers related to this project, the experiment and numerical simulation results are reported. This paper referred these results and performed further investigation.

Comparison Between Pressure Tests and Simulations for Thickness Management of Wall Thinning T-Joints

2012 ◽  
Author(s):  
Kaina Teshima ◽  
Mayumi Ochi ◽  
Seiji Asada ◽  
Kiminobu Hojo ◽  
Takahiro Suzuki ◽  
...  
Keyword(s):  
Experimental Tests ◽  
Ultimate Tensile Stress ◽  
Fe Analysis ◽  
Test Results ◽  
Minimum Thickness ◽  
T Joints ◽  
Wall Thinning ◽  
Pressure Tests ◽  
Pressure Test ◽  
Design Pressure

Although the required minimum thickness (tsr) of T-joints (tees) is calculated from the design pressure and temperature for design, there are no rules or standards for thickness management of wall thinning T-joints for facilities maintenance. This paper describes the comparison between the five pressure test results of T-joints and their numerical simulations using FE analysis and confirms the failure criterion. The investigation for the numerical simulation and the experimental tests showed that ultimate tensile stress (σu) is the most suitable criterion for the burst of T-joints.

Load Factor for Evaluating Non-Planar Flaws in Carbon Steel Piping Using Limit Load Methodology

2016 ◽  
Author(s):  
Phuong H. Hoang
Keyword(s):  
Carbon Steel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Local Strain ◽  
Limit Load ◽  
Evaluation Methodology ◽  
Load Factor ◽  
Wall Thinning ◽  
Steel Piping

Non-planar flaw such as local wall thinning flaw is a major piping degradation in nuclear power plants. Hundreds of piping components are inspected and evaluated for pipe wall loss due to flow accelerated corrosion and microbiological corrosion during a typical scheduled refueling outage. The evaluation is typically based on the original code rules for design and construction, and so often that uniformly thin pipe cross section is conservatively assumed. Code Case N-597-2 of ASME B&PV, Section XI Code provides a simplified methodology for local pipe wall thinning evaluation to meet the construction Code requirements for pressure and moment loading. However, it is desirable to develop a methodology for evaluating non-planar flaws that consistent with the Section XI flaw evaluation methodology for operating plants. From the results of recent studies and experimental data, it is reasonable to suggest that the Section XI, Appendix C net section collapse load approach can be used for non-planar flaws in carbon steel piping with an appropriate load multiplier factor. Local strain at non-planar flaws in carbon steel piping may reach a strain instability prior to net section collapse. As load increase, necking starting at onset strain instability leads to crack initiation, coalescence and fracture. Thus, by limiting local strain to material onset strain instability, a load multiplier factor can be developed for evaluating non-planar flaws in carbon steel piping using limit load methodology. In this paper, onset strain instability, which is material strain at the ultimate stress from available tensile test data, is correlated with the material minimum specified elongation for developing a load factor of non-planar flaws in various carbon steel piping subjected to multiaxial loading.

Efficient Intensity Measures for Probabilistic Seismic Response Analysis of Anchored Above-Ground Liquid Steel Storage Tanks

2016 ◽  
Author(s):  
Hoang Nam Phan ◽  
Fabrizio Paolacci
Keyword(s):  
Ground Motion ◽  
Earthquake Engineering ◽  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Time History ◽  
Response Analysis ◽  
Storage Tanks ◽  
Intensity Measures ◽  
Study Results

Liquid storage tanks are vital lifeline structures and have been widely used in industries and nuclear power plants. In performance-based earthquake engineering, the assessment of probabilistic seismic risk of structural components at a site is significantly affected by the choice of ground motion intensity measures (IMs). However, at present there is no specific widely accepted procedure to evaluate the efficiency of IMs used in assessing the seismic performance of steel storage tanks. The study presented herein concerns the probabilistic seismic analysis of anchored above-ground steel storage tanks subjected to several sets of ground motion records. The engineering demand parameters for the analysis are the compressive meridional stress in the tank wall and the sloshing wave height of the liquid free surface. The efficiency and sufficiency of each alternative IM are quantified by results of time history analyses for the structural response and a proper regression analysis. According to the comparative study results, this paper proposes the most efficient and sufficient IMs with respect to the above demand parameters for a portfolio of anchored steel storage tanks.

Effect of Opening on the Response of Containment Model to Pressure Loading

2006 ◽  
Author(s):  
M. K. Agrawal ◽  
A. Ravi Kiran ◽  
A. K. Ghosh ◽  
H. S. Kushwaha
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Dead Weight ◽  
Modeling Methodology ◽  
Containment Model ◽  
Scale Ratio ◽  
Heavy Water Reactor ◽  
Set Up ◽  
Accident Conditions ◽  
Design Pressure

The Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other phenomena which occur during simulated accident conditions in Nuclear power Plants. The facility consists of a concrete containment model having a volumetric scale ratio of 200:1 with respect to the actual containment of Indian Pressurized Heavy Water Reactor. The structure is designed for pressure of 1.73 Kg/cm2 for specified leak tightness. Adequacy to withstand design pressure is checked by test as well as numerical analysis before commissioning of the facility. Accordingly Containment building model has been analyzed by finite element method for internal design pressure and dead weight. Analysis has been carried out for the structure with and without the opening in the containment. Effect of opening on the response of containment has been studied. The paper includes the modeling methodology, maximum deflection and stress amplification around the opening for various models.

Failure Probability Assessment of Piping Systems Affected by SCC and Pipe Wall Thinning for LBB Analysis

2011 ◽  
Author(s):  
Hiromasa Chitose ◽  
Hideo Machida ◽  
Itaru Saito
Keyword(s):  
Failure Probability ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Probability Assessment ◽  
Wall Thinning ◽  
Piping Systems ◽  
Leak Before Break ◽  
Thinning Rate ◽  
Service Inspection

This paper provides failure probability assessment results for piping systems affected by stress corrosion cracking (SCC) and pipe wall thinning in nuclear power plants. On the basis of the results, considerations for applying the leak-before-break (LBB) concept in actual plants are presented. The failure probability for SCC satisfies the target failure probability even if conservative conditions are assumed. Moreover, for pipe wall thinning analysis, pre-service inspection is important for satisfying the target failure probability because the initial wall thickness affects the accuracy of the wall thinning rate. The pipe wall thinning analysis revealed that the failure probability is higher than the target probability if the bending stress in the pipe is large.

scholarly journals Analyses of the Control System Strategies and Methodology for Part Power Control of the Simple and Intercooled Recuperated Brayton Helium Gas Turbine Cycles for Generation IV Nuclear Power Plants

2017 ◽  
Vol 3 (4) ◽  
Author(s):  
A. Gad-Briggs ◽  
P. Pilidis ◽  
T. Nikolaidis
Keyword(s):  
Control System ◽  
Inventory Control ◽  
Nuclear Power ◽  
Power Plants ◽  
Control Method ◽  
Control Strategies ◽  
Limiting Factors ◽  
Cycle Performance ◽  
Study Results ◽  
And Performance

An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. First, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances are also documented. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modeling and performance simulation tool designed for this study. Results show that the ICR takes ∼102% longer than the SCR to reduce the load to 50% in design point (DP) performance conditions for similar valve flows, which correlates with the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically gas cooled fast reactors (GFRs) and very high-temperature reactors (VHTRs), where helium is the coolant.

Effects of Flow Accelerated Corrosion on Piping Steady-State Vibration Evaluations

2003 ◽  
Author(s):  
Brian J. Voll
Keyword(s):  
Steady State ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Vibration Monitoring ◽  
Accelerated Corrosion ◽  
Monitoring Programs ◽  
Wall Thinning ◽  
Piping Systems ◽  
Flow Accelerated Corrosion

Piping steady-state vibration monitoring programs were implemented during preoperational testing and initial plant startup at most nuclear power plants. Evaluations of piping steady-state vibrations are also performed as piping and component failures attributable to excessive vibration are detected or other potential vibration problems are detected during plant operation. Additionally, as a result of increased flow rates in some piping systems due to extended power uprate (EPU) programs at several plants, new piping steady-state vibration monitoring programs are in various stages of implementation. As plants have aged, pipe wall thinning resulting from flow accelerated corrosion (FAC) has become a recognized industry problem and programs have been established to detect, evaluate and monitor pipe wall thinning. Typically, the piping vibration monitoring and FAC programs have existed separately without interaction. Thus, the potential impact of wall thinning due to FAC on piping vibration evaluations may not be recognized. The potential effects of wall thinning due to FAC on piping vibration evaluations are reviewed. Piping susceptible to FAC and piping susceptible to significant steady-state vibrations, based on industry experience, are identified and compared. Possible methods for establishing links between the FAC and vibration monitoring programs and for accounting for the effects of FAC on both historical and future piping vibration evaluations are discussed.

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