Evaluation of the Environmental Fatigue at a Long Radius Elbow of a Piping System under Transient Thermo Mechanical Stresses

2021 ◽  
Vol 412 ◽  
pp. 197-206
Author(s):  
Lenin Ramos-Cantú ◽  
Luis Héctor Hernández-Gómez ◽  
Rafael García-Illescas ◽  
Tanya Nerina Arreola-Valles ◽  
José Javier Moctezuma-Reyes ◽  
...  
Keyword(s):  
Nuclear Power ◽  
High Stress ◽  
Thermal Fluctuations ◽  
Mechanical Analysis ◽  
Piping System ◽  
Transient Effects ◽  
Stress Concentrations ◽  
Pipe System ◽  
Thermal Transients ◽  
Piping Systems

Thermal fatigue widely takes place in light water reactor (LWR) piping systems. It is an important aging mechanism of a nuclear reactor. Thermal transient effects occur at the startup and shutdown of a nuclear power plant. During the thermal transients, local and global cyclic stresses are induced in the piping systems. They are exacerbated by local geometric imperfections and environmental factors, which may lead to crack initiation. The elbows of such piping systems are subject to various combinations of loads (internal pressure, bending, and torsion, as well as thermal fluctuations) during their service life. As can be seen, high-stress concentrations are developed in these piping elements. Therefore, it is important to make a failure evaluation. In this paper, a 12” pipe system segment, which was made with SA 106 Gr C steel, has been considered. It was composed by two straight sections joined by a long radius elbow. Typical start-up and shutdown transient effects of a BWR-5 were considered. A computer-aided thermo-mechanical analysis was carried out using the finite element method. The linearization of the stresses was considered, based on the ASME B & PVC Code Section III, subsection NB. Under these conditions, environmental fatigue was analyzed after 40-and 60-years operation.

Non-Linear Design Verification of Nuclear Power Piping According to ASME III NB/NC

2008 ◽  
Author(s):  
Lingfu Zeng ◽  
Lennart G. Jansson
Keyword(s):  
Nuclear Power ◽  
Design Evaluation ◽  
Piping System ◽  
Design Verification ◽  
Intensive Study ◽  
Non Linear ◽  
Piping Systems ◽  
Design Requirements

A nuclear piping system which is found to be disqualified, i.e. overstressed, in design evaluation in accordance with ASME III, can still be qualified if further non-linear design requirements can be satisfied in refined non-linear analyses in which material plasticity and other non-linear conditions are taken into account. This paper attempts first to categorize the design verification according to ASME III into the linear design and non-linear design verifications. Thereafter, the corresponding design requirements, in particular, those non-linear design requirements, are reviewed and examined in detail. The emphasis is placed on our view on several formulations and design requirements in ASME III when applied to nuclear power piping systems that are currently under intensive study in Sweden.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.

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.

Investigation of Thermal Fatigue Characteristics of Charging Line of RCV in PWR

2017 ◽  
Author(s):  
Naeem Ahmad ◽  
XiangBin Li ◽  
Iftikhar Ahmad ◽  
Nan Li ◽  
Shahroze Ahmed ◽  
...  
Keyword(s):  
Fatigue Damage ◽  
Thermal Fatigue ◽  
Nuclear Power ◽  
Thermal Loading ◽  
Structural Model ◽  
Fatigue Cracks ◽  
Volume Control ◽  
Piping System ◽  
Thermal Transients ◽  
Fatigue Evaluation

Nuclear Power Plant (NPP) components need to tolerate thermal constraints, internal pressure and thermal transients. These thermal transients being repeated again and again can lead to thermal fatigue of the component. It has significant effect on the degradation of the NPP components in long term. Studies of thermal fatigue on different NPP components such as mixing tees and valves have been carried out before but the charging line in the chemical and volume control system (RCV) of the NPP seems to have been ignored for thermal fatigue analysis. Charging Line is the connection from RCV towards Reactor Coolant System (RCP). To enhance the safety of the charging line, thermal fatigue evaluation of piping system was performed using the Fluid Structure Interaction (FSI) analysis. Temperature distributions in the pipes were determined via thermal hydraulic analysis (CFX) and the results were applied to the structural model of the piping system to determine the thermal stress (Transient Structural). Results revealed the location of fatigue cracks. Types of stress were identified that caused the fatigue damage. The CFD analysis enabled us to clarify the role of turbulence with respect to the thermal loading of the structure. The study will provide valuable information for establishing a permanent methodology to help minimize thermal fatigue damage in NPP components.

scholarly journals Nonlinear Seismic Correlation Analysis of the JNES/NUPEC Large-Scale Piping System Tests

2008 ◽  
Author(s):  
Jinsuo Nie ◽  
Giuliano DeGrassi ◽  
Charles H. Hofmayer ◽  
Syed A. Ali
Keyword(s):  
Correlation Analysis ◽  
Test Data ◽  
Nuclear Power ◽  
Large Scale ◽  
Failure Modes ◽  
Plastic Behavior ◽  
Piping System ◽  
Highly Nonlinear ◽  
Piping Systems ◽  
The U.S

The Japan Nuclear Energy Safety Organization/Nuclear Power Engineering Corporation (JNES/NUPEC) large-scale piping test program has provided valuable new test data on high level seismic elasto-plastic behavior and failure modes for typical nuclear power plant piping systems. The component and piping system tests demonstrated the strain ratcheting behavior that is expected to occur when a pressurized pipe is subjected to cyclic seismic loading. Under a collaboration agreement between the U.S. and Japan on seismic issues, the U.S. Nuclear Regulatory Commission (NRC)/ Brookhaven National Laboratory (BNL) performed a correlation analysis of the large-scale piping system tests using detailed state-of-the-art nonlinear finite element models. Techniques are introduced to develop material models that can closely match the test data. The shaking table motions are examined. The analytical results are assessed in terms of the overall system responses and the strain ratcheting behavior at an elbow. The paper concludes with the insights about the accuracy of the analytical methods for use in performance assessments of highly nonlinear piping systems under large seismic motions.

scholarly journals A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

2013 ◽  
Vol 135 (5) ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Nuclear Industry ◽  
Ignition Source ◽  
Piping System ◽  
Nuclear Facilities ◽  
Piping Systems ◽  
Fluid Transients ◽  
Hydrogen Ignition ◽  
Explosion Mechanism

Hydrogen explosions may occur simultaneously with fluid transients' accidents in nuclear facilities, and a theoretical mechanism to relate fluid transients to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in piping systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the piping system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany (Fig. 1Fig. 1Hydrogen explosion damage in nuclear facilities Antaki, et al. [9,10–12] (ASME, Task Group on Impulsively Loaded Vessels, 2009, Bob Nickell)). Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism. In fact, this explosion mechanism may be pertinent to explosions in major nuclear accidents, and a similar explosion mechanism is also possible in oil pipelines during off-shore drilling.

Improvement of the Damping Constants for Seismic Design of Piping System for NPP

2002 ◽  
Author(s):  
Kei Kobayashi ◽  
Takashi Satoh ◽  
Nobuyuki Kojima ◽  
Kiyoshi Hattori ◽  
Masaki Nakagawa ◽  
...  
Keyword(s):  
Power Plant ◽  
Seismic Design ◽  
Nuclear Power ◽  
Power Plants ◽  
Estimation Method ◽  
Piping System ◽  
Piping Systems ◽  
Support Element ◽  
Present Design ◽  
Bolt Support

The present design damping constants for nuclear power plant (NPP)’s piping system in Japan were developed through discussion among expert researchers, electric utilities and power plant manufactures. They are standardized in “Technical guidelines for seismic design of Nuclear Power Plants” (JEAG 4601-1991 Supplemental Edition). But some of the damping constants are too conservative because of a lack of experimental data. To improve this excessive conservatism, piping systems supported by U-bolts were chosen and U-bolt support element test and piping model excitation test were performed to obtain proper damping constants. The damping mechanism consists of damping due to piping materials, damping due to fluid interaction, damping due to plastic deformation of piping and supports, and damping due to friction and collision between piping and supports. Because the damping due to friction and collision was considered to be dominant, we focused our effort on formulating these phenomena by a physical model. The validity of damping estimation method was confirmed by comparing data that was obtained from the elemental tests and the actual scale piping model test. New design damping constants were decided from the damping estimations for piping systems in an actual plant. From now on, we will use the new design damping constants for U-bolt support piping systems, which were proposed from this study, as a standard in the Japanese piping seismic design.

Use of Markov Piping Reliability Models to Evaluate Time Dependent Frequencies of Loss of Coolant Accidents

2004 ◽  
Author(s):  
Karl N. Fleming ◽  
Bengt O. Y. Lydell
Keyword(s):  
Markov Model ◽  
Nuclear Power ◽  
Interesting Property ◽  
Time Dependent ◽  
Piping System ◽  
Accelerated Corrosion ◽  
Reliability Models ◽  
Age Related ◽  
Loss Of Coolant ◽  
Piping Systems

Markov model theory has been applied to develop a method to evaluate the influence of alternate strategies for in-service inspection and leak detection on the frequency of leaks and ruptures in nuclear power plant piping systems [1–4]. This approach to quantification of pipe rupture frequency was originally based on a Bayes’ uncertainty analysis approach to derive piping system failure rates from a combination of service experience data and some simple reliability models [5–7]. More recently the Markov model approach has been used in conjunction with probabilistic fracture mechanics methods in the study of flow accelerated corrosion [8]. One interesting property of the Markov model is its capability to evaluate time dependent rupture frequencies via the model hazard rate. In this paper this time dependent modeling capability is used to investigate the age related and time dependent frequencies of loss of coolant accident (LOCA) initiating event frequencies. A case is presented that plant age dependent LOCA frequencies should be used in lieu of other metrics commonly used in probabilistic risk assessments and in risk informed inservice inspection evaluations. Such more commonly used metrics include the assumed constant failure rate method and the lifetime average rupture probability. Both of these methods are shown to provide optimistic estimates of LOCA frequencies for plants in the latter part of their design lifetimes, which most operating plants are approaching.

Development of Elastic-Plastic Constitutive Models for the Evaluation of Nuclear Piping Systems Under Severe Cyclic Loading

2012 ◽  
Author(s):  
Yukio Takahashi ◽  
Yoshihiko Tanaka
Keyword(s):  
Cyclic Loading ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Kinematic Hardening ◽  
Constitutive Models ◽  
Seismic Loading ◽  
Piping System ◽  
Elastic Plastic ◽  
Piping Systems

It is essential to predict the behavior of nuclear piping system under seismic loading to evaluate the structural integrity of nuclear power plants. Relatively large stress cycles may be applied to the piping systems under severe seismic loading and plastic deformation may occur cyclically in some portion of the systems. Accurate description of inelastic deformation under cyclic loading is indispensable for the precise estimation of strain cycles and accumulation potentially leading to the failure due to fatigue-ratcheting interaction. Elastic-plastic constitutive models based on the nonlinear kinematic hardening rule proposed by Ohno and Wang were developed for type 316 austenitic stainless steel and carbon steel JIS STPT410 (similar to ASTM A106 Gr.B), both of which are used in piping systems in nuclear power plants. Different deformation characteristics under cyclic loading in terms of memory of prior hardening were observed on these two materials and they were reflected in the modeling. Results of simulations under various loading conditions were compared with the test data to demonstrate the high capability of the constitutive models.

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