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.

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.

A Feasibility Study of Noncontact Ultrasonic Sensor for Nuclear Power Plant Inspection

2017 ◽  
Vol 3 (2) ◽  
Author(s):  
Akinori Tamura ◽  
Chenghuan Zhong ◽  
Anthony J. Croxford ◽  
Paul D. Wilcox
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Pipe Wall ◽  
Thickness Measurement ◽  
Piping System ◽  
Wall Thinning ◽  
Plant Inspection ◽  
Ultrasonic Thickness

A pipe-wall thinning measurement is a key inspection to ensure the integrity of the piping system in nuclear power plants. To monitor the integrity of the piping system, a number of ultrasonic thickness measurements are manually performed during the outage of the nuclear power plant. Since most of the pipes are covered with an insulator, removing the insulator is necessary for the ultrasonic thickness measurement. Noncontact ultrasonic sensors enable ultrasonic thickness inspection without removing the insulator. This leads to reduction of the inspection time and reduced radiation exposure of the inspector. The inductively-coupled transducer system (ICTS) is a noncontact ultrasonic sensor system which uses electromagnetic induction between coils to drive an installed transducer. In this study, we investigated the applicability of an innovative ICTS developed at the University of Bristol to nuclear power plant inspection, particularly pipe-wall thinning inspection. The following experiments were performed using ICTS: thickness measurement performance, the effect of the coil separation, the effect of the insulator, the effect of different inspection materials, the radiation tolerance, and the measurement accuracy of wastage defects. These initial experimental results showed that the ICTS has the possibility to enable wall-thinning inspection in nuclear power plants without removing the insulator. Future work will address the issue of measuring wall-thinning in more complex pipework geometries and at elevated temperatures.

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.

Experimental Investigation on Failure Estimation Method for Circumferentially Cracked Pipes Subjected to Combined Bending and Torsion Loads

2013 ◽  
Author(s):  
Yinsheng Li ◽  
Kunio Hasegawa ◽  
Michiya Sakai ◽  
Shinichi Matsuura ◽  
Naoki Miura
Keyword(s):  
Experimental Investigation ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Estimation Method ◽  
Bending Moment ◽  
Limit Load ◽  
Piping System ◽  
Torsion Moment

When a crack is detected in a nuclear piping system during in-service inspections, the failure estimation method provided in codes such as the ASME Boiler and Pressure Vessel Code Section XI or JSME Rules on Fitness-for-Service for Nuclear Power Plants can be applied to evaluate the structural integrity of the cracked pipe. In the current codes, the failure estimation method for circumferentially cracked pipes includes bending moment and axial force due to pressure. Torsion moment is not considered. The Working Group on Pipe Flaw Evaluation for the ASME Boiler and Pressure Vessel Code Section XI is developing guidance for combining torsion load within the existing solutions provided in Appendix C for bending and pressure loadings on a pipe. A failure estimation method for circumferentially cracked pipes subjected to general loading conditions including bending moment, internal pressure and torsion moment with general magnitude has been proposed based on analytical investigations on the limit load for cracked pipes. In this study, experimental investigation was conducted to confirm the applicability of the proposed failure estimation method. Experiments were carried out on 8-inch diameter Schedule 80 stainless steel pipes containing a circumferential surface crack. Based on the experimental results, the proposed failure estimation method was confirmed to be applicable to cracked pipes subjected to combined bending and torsion moments.

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.

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.

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.

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.

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