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.

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.

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.

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.

Elastic-Plastic Analysis of Thick-Walled Toroidal Pressure Vessels

2008 ◽  
Author(s):  
R. Adibi-Asl
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Three Dimensional ◽  
Pressure Vessels ◽  
Toroidal Shell ◽  
Piping System ◽  
Straight Pipe ◽  
Element Analysis ◽  
Process Industries ◽  
Piping Systems

Piping systems in process industries and nuclear power plants include straight pipe runs and various fittings such as elbows, miter bends etc. Elbows and bends in piping systems provide additional flexibility to the piping system along with performing the primary function of changing the direction of fluid flow. Distinctive geometry of these toroidal shell components result in a structural behavior different from straight pipe. Hence, it would be useful to predict the behavior of these components with acceptable accuracy for design purposes. Analytical expressions are derived for stresses set up during loading and unloading in a toroidal shell subjected to internal pressure. Residual stresses in the component are also evaluated. The proposed solutions are then compared with three-dimensional finite element analysis at different locations including intrados, extrados and flanks.

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.

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.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

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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