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.

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.

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.

Stable and Unstable Crack Growth of A508 Class 3 Plates Subjected to Combined Force of Thermal Shock and Tension

1989 ◽  
Vol 111 (3) ◽  
pp. 234-240 ◽  
Author(s):  
G. Yagawa ◽  
Y. Ando ◽  
K. Ishihara ◽  
T. Iwadate ◽  
Y. Tanaka
Keyword(s):  
Thermal Shock ◽  
Nuclear Power ◽  
Power Plants ◽  
Fracture Behavior ◽  
Structural Integrity ◽  
Three Dimensional ◽  
Test Results ◽  
Unstable Crack ◽  
Pressurized Thermal Shock ◽  
Series Of Experiments

An urgent problem for nuclear power plants is to assess the structural integrity of the reactor pressure vessel under pressurized thermal shock. In order to estimate crack behavior under combined force of thermal shock and tension simulating pressurized thermal shock, two series of experiments are demonstrated: one to study the effect of material deterioration due to neutron irradiation on the fracture behavior, and the other to study the effect of system compliance on fracture behavior. The test results are discussed with the three-dimensional elastic-plastic fracture parameters, J and Jˆ integrals.

FSI-Based Assessment of FAC-Caused Wall Thinned Piping

2009 ◽  
Author(s):  
Sun-Hye Kim ◽  
Yoon-Suk Chang ◽  
Young-Jin Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Flow Characteristics ◽  
Limit Load ◽  
Bend Angle ◽  
Piping System ◽  
Integrity Assessment ◽  
Wall Thinning ◽  
Analysis Scheme

Lots of investigations on failures of wall thinned piping have been carried out since the accident of Surry unit 2 in USA. From these preceding efforts, flow accelerated corrosion (FAC) which is a kind of wall thinning phenomenon is revealed main factor of failure of pipes in nuclear power plants. However, there are a few researches which directly take into account of flow characteristics and geometric changes for stress assessment of FAC-caused wall thinned piping. In this paper, structural integrity assessment employing a fluid-structure interaction (FSI) analysis scheme is performed on pipes representing secondary piping system of PWR which consists of straight pipes and elbows of various bend angles. Prior to the assessment, CFD analyses are conducted to predict plausible wall thinning location by considering flow and geometric parameters such as bend angle and radius of elbow. Then, for typical pipe geometry, detailed limit load analyses are performed to calculate maximum stress caused by turbulence and velocity of flow near the wall thinned part. Through these kinds of detailed parametric analyses, effects of FSI were observed, which should be considered for assessment of FAC-caused wall thinned piping.

Limit Analysis of a Pipe Section With Non-Uniform Wall Thinning

2002 ◽  
Author(s):  
Keshab K. Dwivedy
Keyword(s):  
Limit Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Pipe Wall ◽  
Limit Load ◽  
Pipe Section ◽  
Wall Thinning ◽  
Uniform Wall

Certain process piping in nuclear and non-nuclear power plants undergo pipe wall thinning due to flow assisted corrosion (FAC). This localized mechanism of corrosion combined with erosion is complex. The potential degradation of the pipe wall depends upon the water chemistry, operating temperature and pressure, flow velocity, piping material and piping configuration. The management of FAC in a power plant is performed in the following basic steps: Identification of potential locations, UT inspection of locations and characterization of pipe wall thinning, and evaluation of wall thinning to establish structural integrity and/or repair/replacement. The section of the pipe is repaired or replaced if the structural integrity cannot be established until next scheduled inspection. In the past 15 years, FAC programs have been established in nuclear power plants. Structural integrity evaluation is a part of the program. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. Pressure design methods are formalized for uniform and non-uniform wall thinning. However, the limit analysis methods for moment loading in the code rules are formulated for uniform thinning of the wall for simplicity. FAC related wall thinning is truly non-uniform, and treating it as non-uniform in the analysis can show additional structural margin compared to analysis conservatively assuming a uniformly thinned wall. This paper has developed simple analytical formulation of limit load carrying capability of a pipe section with non-uniform thinning. The method of analysis is illustrated with examples of actual plant situations. The formulation developed here can be used with the ASME code method to extend remaining life of FAC degraded components until the plant can plan for repair or replacement. Thus the analytical tool can help the plant owners to save resources by performing repair and replacement in a planned manner.

Canadian Nuclear Regulatory Perspective on Degradation of CANDU Feeder Piping

2005 ◽  
Author(s):  
John C. Jin ◽  
Tom Viglasky ◽  
Andrei Blahoianu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Degradation Mechanisms ◽  
Accelerated Corrosion ◽  
Wall Thinning ◽  
Structural Failures ◽  
Level Of Understanding ◽  
Periodic Inspections

As some CANDU plants in Canada are approaching the end of their design lives, various degradation mechanisms which were not anticipated during the design phase have been identified in the CANDU feeder piping and resulted in either actual structural failures or the early replacement of components. In particular, inter-granular stress corrosion cracking (IGSCC) and pipe wall thinning due to flow accelerated corrosion (FAC) are the most prominent degradation mechanisms in the CANDU feeder piping. The Canadian CANDU industry has developed and implemented programs to monitor and manage those unanticipated service-related degradations. Fitness for service guidelines are also developed to justify the structural integrity of the components until the next inspection. These programs include augmented periodic inspections that are targeted to specific components, thereby ensuring the early detection of cracks or excessive wall thinning. The inspection scope and frequency adopted in the degradation management programs exceed the requirements of ASME Section XI, “Inservice Inspection of Nuclear Reactor Coolant Systems” and CSA N285.4, “Periodic Inspection of CANDU Nuclear Power Plants Components”. However, those currently effective codes and standards do not specify requirements which are developed based on the consideration of the specific degradation mechanisms such as IGSCC and FAC wall thinning. Accordingly, it has been an issue for the nuclear regulator as well as in the industry to develop criteria for inspection and replacement/repair, which are based on the current level of understanding of degradation mechanisms and the inspection capability. Presented in this paper are the Canadian regulator’s perspective on the assurance of the safe operation of the CANDU feeder piping which endures degradations of IGSCC and FAC.

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.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant

scholarly journals Interaction of Cyclic Loading (Low-Cyclic Fatigue) with Stress Corrosion Cracking (SCC) Growth Rate

2020 ◽  
Vol 2020 ◽  
pp. 1-10
Author(s):  
Rehmat Bashir ◽  
He Xue ◽  
Rui Guo ◽  
Yueqi Bi ◽  
Muhammad Usman
Keyword(s):  
Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Cyclic Fatigue ◽  
Joint Effect ◽  
Corrosion Cracking

The structural integrity analysis of nuclear power plants (NPPs) is an essential procedure since the age of NPPs is increasing constantly while the number of new NPPs is still limited. Low-cyclic fatigue (LCF) and stress corrosion cracking (SSC) are the two main causes of failure in light-water reactors (LWRs). In the last few decades, many types of research studies have been conducted on these two phenomena separately, but the joint effect of these two mechanisms on the same crack has not been discussed yet though these two loads exist simultaneously in the LWRs. SCC is mainly a combination of the loading, the corrosive medium, and the susceptibility of materials while the LCF depends upon the elements such as compression, moisture, contact, and weld. As it is an attempt to combine SCC and LCF, this research focuses on the joint effect of SCC and LCF loading on crack propagation. The simulations are carried out using extended finite element method (XFEM) separately, for the SCC and LCF, on an identical crack. In the case of SCC, da/dt(mm/sec) is converted into da/dNScc (mm/cycle), and results are combined at the end. It has been observed that the separately calculated results for SCC da/dNScc and LCF da/dNm of crack growth rate are different from those of joint/overall effect,  da/dNom. By applying different SCC loads, the overall crack growth is measured as SCC load becomes the main cause of failure in LWRs in some cases particularly in the presence of residual stresses.

Fiber-Optic Transmitter for Nuclear Power Plants

2004 ◽  
Author(s):  
Junichi Higashi ◽  
Shinichi Murakawa
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Fiber Optic ◽  
Optic Fiber ◽  
Test Results ◽  
Maintenance System ◽  
Fiber Technology ◽  
Fabry Perot ◽  
Validation Tests

A promising Fiber-Optic Differential Pressure (DP) Transmitter is under development in Flexible Maintenance System (FMS) Projects that supported by Ministry of Economic, Trade, and Industries of Japan. The object of FMS projects is to improve maintenance works at nuclear power plants with latest technology. The new DP Transmitter uses optic-fiber technology of Extrinsic Fabry-Perot Sensor and Fizeau White-Light Cross-Correlator. Validation tests were performed to evaluate the tolerance of the DP transmitter in Nuclear Power Plant conditions. General requirements of PWR are accuracy (repeatability and linearity) of within +/−0.5%, pressure-proof of maximum 17.16MPa, Irradiation of 100Gy, and temperature range of 10–50 degrees centigrade at normal condition. The test results show the new DP transmitter can be expected as the next generation instrumentation in Nuclear Power Plants.

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