Fracture Estimation Method for Pipe With Multiple Circumferential Surface Flaws

2009 ◽  
Author(s):  
Yinsheng Li ◽  
Kunio Hasegawa ◽  
Kunio Onizawa ◽  
Masayoshi Shimomoto
Keyword(s):  
Stainless Steel ◽  
Nuclear Power ◽  
Numerical Solutions ◽  
Estimation Method ◽  
Arbitrary Distribution ◽  
Piping System ◽  
Piping Systems ◽  
Asme Code ◽  
As Stress ◽  
Steel Piping

When a flaw is detected in a stainless steel piping system of a nuclear power plant during in-service inspection, the fracture estimation method provided in the codes such as the ASME Code Section XI or the JSME S NA-1-2004 can be applied to evaluate the integrity of the pipe. However, in these current codes, the fracture estimation method is only provided for the pipe containing a single flaw, although independent multiple flaws such as stress corrosion cracks have actually been detected in the same circumference of stainless steel piping systems. In this paper, a fracture estimation method is proposed by formula for multiple independent circumferential flaws with any number and arbitrary distribution in the same circumference of the pipe. Using the proposed method, the numerical solutions are compared with the experimental results to verify its validity, and several numerical examples are provided to show its effectiveness.

Prediction of Collapse Stress for Pipes With Arbitrary Multiple Circumferential Surface Flaws

2010 ◽  
Vol 132 (6) ◽  
Author(s):  
Yinsheng Li ◽  
Kunio Hasegawa ◽  
Kunio Onizawa ◽  
Nathaniel G. Cofie
Keyword(s):  
Stainless Steel ◽  
Nuclear Power ◽  
Numerical Solutions ◽  
Estimation Method ◽  
Limit Load ◽  
Arbitrary Distribution ◽  
Piping System ◽  
Load Estimation ◽  
As Stress ◽  
Steel Piping

When a flaw is detected in a stainless steel piping system of a nuclear power plant during in-service inspection, the limit load estimation method provided in codes such as ASME Section XI or JSME S NA-1-2008 can be applied to evaluate the integrity of the pipe. However, in the current editions of these codes, a limit load estimation method is only provided for pipes containing a single flaw. Independent multiple flaws, such as stress corrosion cracks, have actually been detected in the same plane of stainless steel piping systems. In this paper, a failure estimation method by formula is proposed for any number and arbitrary distribution of multiple independent circumferential flaws in the same plane of a pipe. Using the proposed method, numerical solutions are compared with experimental results to validate the model, and several numerical examples are provided to show its effectiveness.

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.

Fracture Assessment of Austenitic Stainless Steel Piping With Multiple Flaws in Heat-Affected Zone

2010 ◽  
Author(s):  
Chihiro Narazaki ◽  
Toshiyuki Saito ◽  
Masao Itatani ◽  
Takuya Ogawa ◽  
Takao Sasayama
Keyword(s):  
Stainless Steel ◽  
Heat Affected Zone ◽  
Nuclear Power ◽  
Power Plants ◽  
Estimation Method ◽  
Limit Load ◽  
Low Carbon ◽  
Fracture Assessment ◽  
Steel Piping ◽  
Weld Heat

Stress corrosion cracking (SCC) has been observed as circumferential multiple flaws in the weld heat-affected zone of primary loop recirculation system piping and core shrouds made of low carbon stainless steel. In the Japan Society of Mechanical Engineers code, Rules on Fitness-for-Service for Nuclear Power Plants, there is no fracture assessment of piping with multiple flaws which are not subject to flaw combination rule criteria. Through fracture testing of piping with two circumferential flaws in the weld heat-affected zone, the limit load estimation method was used for fracture assessment of stainless steel piping.

The Role of Displacement-Controlled Stresses in Critical Flaw Size Determination for Piping Systems

2008 ◽  
Author(s):  
Peter C. Riccardella ◽  
Paul Hirschberg ◽  
Ted Anderson ◽  
Greg Thorwald ◽  
Eric Scheibler
Keyword(s):  
Stainless Steel ◽  
Thermal Expansion ◽  
Nuclear Power ◽  
Power Plants ◽  
Bending Moment ◽  
Flaw Size ◽  
Piping System ◽  
Nuclear Plants ◽  
Piping Systems ◽  
Critical Flaw

A debate has long ensued in ASME Subcommittee XI regarding the need to include displacement-controlled (secondary) stresses in critical flaw size calculations for austenitic weldments. There is general agreement that inclusion of secondary stresses is not necessary for highly ductile piping materials such as wrought stainless steel and high nickel alloys. However, some stainless steel weldments are classified as “low-toughness” because, although not considered brittle, they exhibit lower toughness than wrought stainless steel. The Code requires the inclusion of global secondary stresses, such as piping thermal expansion loads, in critical flaw size calculations for such weldments, albeit at reduced safety factors. The Code requirements are less clear for dissimilar metal weldments, such as Alloy 82/182, which were often used for ferritic nozzle to safe-end welds in nuclear power plants, and which have proven in service to be susceptible to a form of stress corrosion cracking. Analyses are presented in this paper that shed additional light on the subject. Finite element analyses (FEA) of a straight pipe with a through-thickness crack were used to determine the effect on bending moment and crack driving force due to an imposed end rotation. Moment and J-integral knock-down factors are computed for a range of crack sizes for two different pipe lengths. Piping analyses are also presented for two typical PWR surge lines, which are among the highest secondary stress locations in U.S. nuclear plants. These analyses predict the maximum rotation at the surge nozzle that could be produced by the secondary loads (anchor movement + thermal expansion + stratification), and compare that to rotations that were sustained in full scale pipe tests containing large complex cracks. The analyses demonstrate that secondary loads would be substantially reduced prior to fracture of a cracked weldment, and that they are therefore of reduced significance in critical flaw size calculations. A general method for estimating the effect of secondary loads on pipe fracture as a function of relative piping system and crack section stiffness is suggested.

Analysis of ASME Class 3 Buried HDPE Piping Systems Related to Code Case N-755-1

2013 ◽  
Author(s):  
Young Seok Kim ◽  
Jung Kwang Yoon ◽  
Young Ho Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Procedure ◽  
Piping System ◽  
Service Water ◽  
Hdpe Pipe ◽  
Division 1 ◽  
Piping Systems ◽  
Asme Code ◽  
Essential Service

This paper proposes an analysis method for Section III, Division 1, Class 3 buried High Density polyethylene (HDPE) piping system in the nuclear power plants (NPP). Although HDPE pipe would yield at high temperature (limited to 140°F), it may be suitable for the areas prone to earthquakes; owing to its comparable ductility and flexibility. Thus, the buried HDPE piping may be applicable for the safety related Essential Service Water (ESW) system in the NPPs. Despite some limitations to buried HDPE piping, the piping could be designed based on ASME Code Case [1]. Generally, codes and standards including ASME Code Case [1] do not provide load combinations for the design of both buried steel piping and HDPE piping. Meanwhile, EPRI Report [4] provides load combinations including thermal expansion effects and seismic loads with detailed seismic criteria for polyethylene pipe. In this paper, load cases and load combinations for buried HDPE piping are suggested for implementation of reference documents and a buried HDPE piping system is analyzed referring to EPRI Report [4] to evaluate stress, force, and moment using a piping stress analysis program. Additionally, this paper will recommend the design procedure in accordance with ASME Code Case [1] using an example of buried HDPE piping analysis. An investigation of soil spring coefficients and the design considerations for hydrostatic tests are suggested for the enhanced analysis of buried HDPE piping.

Recent Operational Experience of Pressurized Water Reactor Safety Injection and Drain Line Cracking and Supporting Flaw Evaluations

2019 ◽  
Author(s):  
Greg Imbrogno ◽  
Stephen Marlette ◽  
Alexandria Carolan ◽  
Anees Udyawar ◽  
Mark Gray
Keyword(s):  
Nuclear Power ◽  
Operating Experience ◽  
Piping System ◽  
Operational Experience ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Piping Systems ◽  
Class 1 ◽  
Asme Code

Abstract A recent increase in operating experience (OE) related to pipe cracking in non-isolable auxiliary piping systems has been realized in the Pressurized Water Reactor (PWR) nuclear power industry. The majority of PWR auxiliary piping systems are comprised of welded stainless steel pipe and piping components. The susceptible piping systems are Class 1 pressure boundary and typically non-isolable from the primary loop. Since they are non-isolable, when a pipe crack or crack indication is identified, an emergent flaw evaluation and/or repair is required. Typically, the evaluations begin with an ASME Section XI IWB-3640 flaw evaluation to determine acceptability of the as-found flaw at the time of shutdown. Subsequent flaw evaluations are performed to demonstrate the possibility of continued operation of the piping component by leaving the flaw as-is without repair. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth evaluations, and the detection capabilities of the non-destructive examination technique. If evaluation of the as-found indication does not produce acceptable results, then a repair/replacement activity per ASME Section XI is considered. Possible repair scenarios include replacement of the piping section or component, or structural weld overlay. The results of the flaw evaluations or repairs must ensure that the auxiliary piping system will continue to operate safely. This paper will discuss the recent experiences of two different piping systems (boron injection tank line and drain line) that experienced cracking, the potential causes for the cracking in the absence of evidence, and the ASME Code flaw evaluations and/or repairs performed to support continued operation of the plant.

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.

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