scholarly journals Quantitative Limit State Assessment of a 3-Inch Carbon Steel Pipe Tee in a Nuclear Power Plant Using a Damage Index

Energies ◽  
2020 ◽  
Vol 13 (23) ◽  
pp. 6395
Author(s):  
Sung-Wan Kim ◽  
Da-Woon Yun ◽  
Sung-Jin Chang ◽  
Dong-Uk Park ◽  
Bub-Gyu Jeon
Keyword(s):  
Carbon Steel ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Damage Index ◽  
Failure Criteria ◽  
Relative Displacement ◽  
Steel Pipe ◽  
Piping System ◽  
Piping Systems

Seismic motions are likely to cause large displacements in nuclear power plants because the main mode of their piping systems is dominated by the low-frequency region. Additionally, large relative displacement may occur in the piping systems because their supports are installed in several places, and each support is subjected to different seismic motions. Therefore, to assess the seismic performance of a piping system, the relative displacement repeated by seismic motions must be considered. In this study, in-plane cyclic loading tests were conducted under various constant amplitudes using test specimens composed of SCH 40 3-inch pipes and a tee in the piping system of a nuclear power plant. Additionally, an attempt was made to quantitatively express the failure criteria using a damage index based on the dissipated energy that used the force–displacement and moment–deformation angle relationships. The failure mode was defined as the leakage caused by a through-wall crack, and the failure criteria were compared and analyzed using the damage index of Park and Ang and that of Banon. Additionally, the method of defining the yield point required to calculate the damage index was examined. It was confirmed that the failure criteria of the SCH 40 3-inch carbon steel pipe tee can be effectively expressed using the damage index.

Lessons Learned for Nuclear Piping Integrity in New Reactors

2011 ◽  
Author(s):  
G. Wilkowski ◽  
F. Brust ◽  
P. Krishnaswamy ◽  
K. Wichman ◽  
D.-J. Shim
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Lessons Learned ◽  
Dynamic Strain ◽  
Piping System ◽  
Aging Effects ◽  
Piping Systems ◽  
Loading Effects

From the early 1980’s to the present time, there has been a significant amount of research and development on the structural integrity of nuclear power plant piping. From those efforts, there are a number of lessons that could be applied to design and fabrication of new nuclear power plant piping systems. In this paper, the various aspects evaluated in NRC-funded efforts for understanding degraded piping were reviewed and implications on how to avoid detrimental aspects were discussed, as well as some more recent efforts. Some of these aspects include; (1) materials aspects (variability of wrought stainless steel base metal toughness with composition, dynamic strain aging effects on toughness of ferritic steels, fracture toughness in HAZ/fusion lines, material anisotropy effects on toughness, effects of static versus dynamic loading on material toughness, cyclic loading effects during seismic loading on toughness, thermal aging effects on strength and toughness), (2) designing weld sequencing to avoid SCC cracking; (3) crack morphology effects on leak-rate evaluations, (4) system effects that can significantly affect the structural integrity analysis of the piping system (secondary stresses, restraint of pressure induced bending, system displacement and rotation constraints, and margins associated from full dynamic analyses).

Application of Probabilistic Fracture Mechanics Analysis on BWR Recirculation Piping Systems

2016 ◽  
Author(s):  
Yu-Yu Shen ◽  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Ru-Feng Liu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fracture Mechanics ◽  
Nuclear Power ◽  
Fracture Probability ◽  
Probability Of Detection ◽  
Piping System ◽  
Probabilistic Fracture Mechanics ◽  
Piping Systems ◽  
Inspection And Maintenance

In recent years, the probabilistic fracture mechanics (PFM) approach has been widely applied to estimate the fracture risk of nuclear power plant piping systems. In the paper, the probabilistic fracture mechanics code, PRO-LOCA, developed by the Probabilistic Analysis as a Regulatory Tool for Risk Informed Decision Guidance (PARTRIDGE) project, is employed to practically evaluate the fracture probability of the recirculation piping system welds in a Taiwan domestic boiling water reactor (BWR) nuclear power plant. To begin with, the models based on the real situation of the recirculation piping welds are built. Then, the probabilities of through-wall cracking, leak with different rates, and rupture on the welds considering both in-service inspection and leak detection are analyzed. Meanwhile, the effects of probability of detection curves of ISI on the piping are simulated. Further, the efficiencies of performing the induction heating stress improvement and weld overlay are also studied and discussed. The present work could provide a reference of operation, inspection and maintenance for BWR plants in Taiwan.

Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities: Part 2 — Application of CCFS Method to the Multiply Supported Systems

2014 ◽  
Author(s):  
Koichi Tai ◽  
Keisuke Sasajima ◽  
Shunsuke Fukushima ◽  
Noriyuki Takamura ◽  
Shigenobu Onishi
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Seismic Isolation ◽  
Evaluation Method ◽  
Time History ◽  
Piping System ◽  
History Analysis ◽  
Piping Systems ◽  
Isolation Systems ◽  
Seismic Isolation Systems

This paper provides a part of series of “Development of an Evaluation Method for Seismic Isolation Systems of Nuclear Power Facilities”. Paper is focused on the seismic evaluation method of the multiply supported systems, as the one of the design methodology adopted in the equipment and piping system of the seismic isolated nuclear power plant in Japan. Many of the piping systems are multiply supported over different floor levels in the reactor building, and some of the piping systems are carried over to the adjacent building. Although Independent Support Motion (ISM) method has been widely applied in such a multiply supported seismic design of nuclear power plant, it is noted that the shortcoming of ignoring correlations between each excitations is frequently misleaded to the over-estimated design. Application of Cross-oscillator, Cross-Floor response Spectrum (CCFS) method, proposed by A. Asfura and A. D. Kiureghian[1] shall be considered to be the excellent solution to the problems as mentioned above. So, we have introduced the algorithm of CCFS method to the FEM program. The seismic responses of the benchmark model of multiply supported piping system are evaluated under various combination methods of ISM and CCFS, comparing to the exact solutions of Time History analysis method. As the result, it is demonstrated that the CCFS method shows excellent agreement to the responses of Time History analysis, and the CCFS method shall be one of the effective and practical design method of multiply supported systems.

Current Wall Thinning Measured on Piping System of Main or Auxiliary Boiler Plant in Ships

2008 ◽  
Author(s):  
H. Shiihara ◽  
H. Matsushita ◽  
Y. Nagayama
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Wall Thickness ◽  
Nuclear Power ◽  
Feed Water ◽  
Piping System ◽  
Thickness Gauge ◽  
Corrosion Failure ◽  
Boiler Plant ◽  
Secondary Line

A disaster happened in a nuclear power plant in Japan in August 2004, which was caused by failure of condensation water pipe in the secondary line. Shipping industries were concerned for possibility of occurrence of such a disaster in ships due to its construction similarity to marine boiler plant in steam, feed water and condensation piping for main or auxiliary boilers. Nippon Kaiji Kyokai has therefore investigated and gathered data of piping lines corrosion in ships collaborated with major Japanese ship owners right after the disaster. The results show that similar corrosion failure as in the nuclear power plant has occurred in shipboard steam/feed water/condensation water pipes for main and auxiliary boiler plants without causing severe consequences. The wall thickness measurements on actual pipe lines of steam, feed water and condensation water at bend parts, at T-junction, behind orifices, behind valves and at diffusers/reducers with a ultrasonic thickness gauge show a very definite evidence of a reduction in wall thickness of carbone steel pipes. It was confirmed that the amount of actual reduction in wall thickness could be well predicted by Kastner Equation [2–3].

A Frequency Approach to Mathematical Modeling of a Nuclear Power Plant Piping System

1985 ◽  
Vol 107 (1) ◽  
pp. 106-111 ◽  
Author(s):  
V. Skormin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Model Updating ◽  
Current Model ◽  
Statistical Technique ◽  
Parameter Estimates ◽  
Piping System ◽  
Model Configuration ◽  
Function Matrix

A methodology is presented for identification of a nuclear power plant piping system, which employs mathematical description in the form of transfer function matrix, frequency domain technique for estimation of system dynamic parameters, statistical technique for verification of model configuration and evaluation of parameter estimates, adaptive approach for current model updating. Model applications for estimation and monitoring of forcing functions, displacements, and stresses due to transient processes and steady state vibrations in the piping system are proposed. Methodology is illustrated by numerical examples.

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