WorldWideScience

Sample records for pipe element testing

  1. Finite-element analysis of flawed and unflawed pipe tests

    International Nuclear Information System (INIS)

    James, R.J.; Nickell, R.E.; Sullaway, M.F.

    1989-12-01

    Contemporary versions of the general purpose, nonlinear finite element program ABAQUS have been used in structural response verification exercises on flawed and unflawed austenitic stainless steel and ferritic steel piping. Among the topics examined, through comparison between ABAQUS calculations and test results, were: (1) the effect of using variations in the stress-strain relationship from the test article material on the calculated response; (2) the convergence properties of various finite element representations of the pipe geometry, using shell, beam and continuum models; (3) the effect of test system compliance; and (4) the validity of ABAQUS J-integral routines for flawed pipe evaluations. The study was culminated by the development and demonstration of a ''macroelement'' representation for the flawed pipe section. The macroelement can be inserted into an existing piping system model, in order to accurately treat the crack-opening and crack-closing static and dynamic response. 11 refs., 20 figs., 1 tab

  2. Finite Element Analysis of Pipe T-Joint

    OpenAIRE

    P.M.Gedkar; Dr. D.V. Bhope

    2012-01-01

    This paper reports stress analysis of two pressurized cylindrical intersection using finite element method. The different combinations of dimensions of run pipe and the branch pipe are used to investigate thestresses in pipe at the intersection. In this study the stress analysis is accomplished by finite element package ANSYS.

  3. Pipe rupture test results: 4-inch pipe whip tests under PWR LOCA conditions

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki; Ueda, Shuzo; Isozaki, Toshikuni; Kato, Rokuro; Kurihara, Ryoichi; Yano, Toshikazu; Miyazono, Shohachiro

    1982-09-01

    This report summarizes the results of 4-inch pipe whip tests (RUN No. 5506, 5507, 5508 and 5604) under the PWR LOCA conditions. The dynamic behaviors of the test pipe and restraints were studied in the tests. In the tests, the gap between the test pipe and the restraints was kept at the constant value of 8.85 mm and the overhang length was varied from 250 mm to 650 mm. The dynamic behaviors of the test pipe and the restraint were made clear by the outputs of strain gages and the measurements of residual deformations. The data of water hammer in subcooled water were also obtained by the pressure transducers mounted on the test pipe. The main conclusions obtained from the tests are as follows. (1) The whipping of pipe can be prevented more effectively as the overhang length becomes shorter. (2) The load acting on the restraint-support structure becomes larger as the overhang length becomes shorter. (3) The restraint farther from the break location does not limit the pipe movement except for the first impact when the overhang length is long. (4) The ultimate moment M sub(u) of the pipe at the restraint location can be used to predict the plastic collapse of the whipping pipe. (5) The restraints slide along the pipe axis and are subjected to bending moment, when the overhang length is long. (author)

  4. Fatigue evaluation of piping systems with limited vibration test data

    International Nuclear Information System (INIS)

    Huang, S.N.

    1990-11-01

    The safety-related piping in a nuclear power plant may be subjected to pump- or fluid-induced vibrations that, in general, affect only local areas of the piping systems. Pump- or fluid-induced vibrations typically are characterized by low levels of amplitudes and a high number of cycles over the lifetime of plant operation. Thus, the resulting fatigue damage to the piping systems could be an important safety concern. In general, tests and/or analyses are used to evaluate and qualify the piping systems. Test data, however, may be limited because of lack of instrumentation in critical piping locations and/or because of difficulty in obtaining data in inaccessible areas. This paper describes and summarizes a method to use limited pipe vibration test data, along with analytical harmonic response results from finite-element analyses, to assess the fatigue damage of nuclear power plant safety-related piping systems. 5 refs., 2 figs., 11 tabs

  5. High-level seismic tests of piping at the HDR [Heissdampfreaktor

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.; Costello, J.F.

    1989-01-01

    As part of the second-phase testing at the Heissdampfreaktor (HDR) Test Facility in Kahl/Main, Federal Republic of Germany (FRG), high-level seismic experiments, designated SHAM, were performed on an in-plant piping system during the period of 19 April to 27 May 1988. The objectives of the SHAM experiments were to (1) study the response of piping subjected to seismic excitation levels that exceed design levels manifold and which may result in failure/plastification of pipe supports and pipe elements; (2) provide data for the validation of linear and nonlinear pipe response analyses; (3) compare and evaluate, under identical loading conditions, the performance of various dynamic support system, ranging from very flexible to very stiff support configurations; (4) establish seismic margins for piping, dynamic pipe supports, and pipe anchorages; and (5) investigate the response, operability, and fragility of dynamic supports and of a typical US gate valve under extreme levels of seismic excitation. A brief description of the SHAM tests is provided, followed by highlights of the test results that are given primarily in the form of maximum response values. Also presented are very limited comparisons of experimental data and pretest analytical predictions. 6 refs., 8 figs

  6. SHAM: High-level seismic tests of piping at the HDR

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.; Malcher, L.; Schrammel, D.; Steinhilber, H.; Costello, J.F.

    1988-01-01

    As part of the second phase of vibrational/earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, high-level simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) in-plant piping system with two servohydraulic actuators, each capable of generating 40 tons of force. The purpose of these experiments was to study the behavior of piping subjected to seismic excitation levels that exceed design levels manifold and may result in failure/plastification of pipe supports and pipe elements, and to establish seismic margins for piping and pipe supports. The performance of six different dynamic pipe support systems was compared in these tests and the response, operability, and fragility of dynamic supports and of a typical US gate valve were investigated. Data obtained in the tests are used to validate analysis methods. Very preliminary evaluations lead to the observation that, in general, failures of dynamic supports (in particular snubbers) occur only at load levels that substantially exceed the design capacity. Pipe strains at load levels exceeding the design level threefold are quite small, and even when exceeding the design level eightfold are quite tolerable. Hence, under seismic loading, even at extreme levels and in spite of multiple support failures, pipe failure is unlikely. 5 refs., 16 figs

  7. Finite element analysis of stemming loads on pipes

    International Nuclear Information System (INIS)

    Maiden, D.E.

    1979-08-01

    A computational model has been developed for calculating the loads and displacements on a pipe placed in a hole which is subsequently filled with soil. A composite soil-pipe finite element model which employs fundamental material constants in its formalism is derived. The shear modulus of the soil, and the coefficient of friction at the pipe are the important constants to be specified. The calculated loads on the pipe are in agreement with experimental data for layered and unlayered stemming designs. As a result more economical designs of the pipe string can be realized

  8. Pipe rupture test results; 6 in. pipe whip test under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi; Yano, Toshikazu; Ueda, Shuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kato, Rokuro; Miyazono, Shohachiro

    1983-02-01

    A series of pipe rupture tests has been performed in JAERI to demonstrate the safety of the primary coolant circuits in the event of pipe rupture, in nuclear power plants. The present report summarizes the results of 6 in. pipe whip tests (RUN 5605, 5606), under BWR LOCA conditions (285 0 C, 6.8 MPa), which were performed in August, 1981. The test pipe is made of Type 304 stainless steel and its outer diameter is 6 in. and its thickness is 11.1 mm. The restraints are made of Type 304 stainless steel and its diameter is 16.0 mm. Two restraints were set on the restraint support with clearance of 100 mm. Overhang length was varied as the parameter in these tests and was 300 mm or 700 mm. The following results are obtained. (1) The deformations of a pipe and restraints are limited effectively by shorter overhang length of 300. However, they become larger when the overhang length is 700 mm, and the pipe deforms especially at the setting point of restraints. (2) Velocity at the free end of pipe becomes about 30 m/sec just after the break. However, velocity at the setting point of restraint becomes about only 4 m/sec just after the break. (3) It seems from the comparison between the 4 in. tests and 6 in. tests that the maximum restraint force of 6 in. tests is about two times as large as that of 4 in. tests. (author)

  9. Response of HDR-VKL piping system to seismic test excitations: Comparison of analytical predictions and test measurements

    International Nuclear Information System (INIS)

    Srinivasan, M.G.; Kot, C.A.; Hsieh, B.J.

    1989-01-01

    As part of the earthquake investigations at the HDR (Heissdampfreaktor) Test Facility in Kahl/Main, FRG, simulated seismic tests (SHAM) were performed during April--May 1988 on the VKL (Versuchskreislauf) piping system. The purpose of these experiments was to study the behavior of piping subjected to a range of seismic excitation levels including those that exceed design levels manifold and that might induce failure of pipe supports or plasticity in the pipe runs, and to establish seismic margins for piping and pipe supports. Data obtained in the tests are also used to validate analysis methods. Detailed reports on the SHAM experiments are given elsewhere. The objective of this document is to evaluate a subsystem analysis module of the SMACS code. This module is a linear finite-element based program capable of calculating the response of nuclear power plant subsystems subjected to independent multiple-acceleration input excitation. The evaluation is based on a comparison of computational results of simulation of SHAM tests with corresponding test measurements

  10. Analysis of pipe mitred bends using beam models - by finite element method

    International Nuclear Information System (INIS)

    Salles, A.C.S.L. de.

    1984-01-01

    The formulation of a recently proposed displacement based straight pipe element for the analysis of pipe mitred bends is summarized in this work. The element kinematics includes axial, bending, torsional and ovalisation displacements, all varying cubically along the axis of the element. Interaction effects between angle adjoined straight pipe section are modeled including the appropriate additional strain terms in the stiffness matrix formulation and by using a penalty procedure to enforce continuity of pipe skin flexural rotations at the common helical edge. The element model capabilities are ilustrated in some sample analysis and the results are compared with other available experimental, analytical or more complex numerical models. (Author) [pt

  11. Ten Year Operating Test Results and Post-Test Analysis of a 1/10 Segment Stirling Sodium Heat Pipe, Phase III

    Science.gov (United States)

    Rosenfeld, John, H; Minnerly, Kenneth, G; Dyson, Christopher, M.

    2012-01-01

    High-temperature heat pipes are being evaluated for use in energy conversion applications such as fuel cells, gas turbine re-combustors, Stirling cycle heat sources; and with the resurgence of space nuclear power both as reactor heat removal elements and as radiator elements. Long operating life and reliable performance are critical requirements for these applications. Accordingly, long-term materials compatibility is being evaluated through the use of high-temperature life test heat pipes. Thermacore, Inc., has carried out a sodium heat pipe 10-year life test to establish long-term operating reliability. Sodium heat pipes have demonstrated favorable materials compatibility and heat transport characteristics at high operating temperatures in air over long time periods. A representative one-tenth segment Stirling Space Power Converter heat pipe with an Inconel 718 envelope and a stainless steel screen wick has operated for over 87,000 hr (10 yr) at nearly 700 C. These life test results have demonstrated the potential for high-temperature heat pipes to serve as reliable energy conversion system components for power applications that require long operating lifetime with high reliability. Detailed design specifications, operating history, and post-test analysis of the heat pipe and sodium working fluid are described.

  12. Small-bore-piping seismic-test findings

    International Nuclear Information System (INIS)

    Severud, L.K.; Barta, D.A.; Anderson, M.J.

    1981-12-01

    Description is given of a test series in which a 1-inch diameter stainless steel pipe system was subjected to dynamic testing. The test system consisted of approximately 40-feet of schedule 40 pipe, with several bends and risers, supported from a rigid test frame. FFTF prototypic pipe clamps, dead weight supports, mechanical snubbers, and insulation were utilized. Several variations of the pipe support configuration were tested. Measured test results are compared with analytical predictions for each configuration. Plans for future testing are discussed

  13. Application of a nonlinear spring element to analysis of circumferentially cracked pipe under dynamic loading

    International Nuclear Information System (INIS)

    Olson, R.; Scott, P.; Wilkowski, G.M.

    1992-01-01

    As part of the US NRC's Degraded Piping Program, the concept of using a nonlinear spring element to simulate the response of cracked pipe in dynamic finite element pipe evaluations was initially proposed. The nonlinear spring element is used to represent the moment versus rotation response of the cracked pipe section. The moment-rotation relationship for the crack size and material of interest is determined from either J-estimation scheme analyses or experimental data. In this paper, a number of possible approaches for modeling the nonlinear stiffness of the cracked pipe section are introduced. One approach, modeling the cracked section moment rotation response with a series of spring-slider elements, is discussed in detail. As part of this discussion, results from a series of finite element predictions using the spring-slider nonlinear spring element are compared with the results from a series of dynamic cracked pipe system experiments from the International Piping Integrity Research Group (IPIRG) program

  14. Pipe rupture test results; 4 inch pipe whip tests under BWR operational condition-clearance parameter experiments

    International Nuclear Information System (INIS)

    Ueda, Syuzo; Isozaki, Toshikuni; Miyazaki, Noriyuki; Kurihara, Ryoichi; Kato, Rokuro; Saito, Kazuo; Miyazono, Shohachiro

    1981-05-01

    The purpose of pipe rupture studies in JAERI is to perform the model tests on pipe whip, restraint behavior, jet impingement and jet thrust force, and to establish the computational method for analyzing these phenomena. This report describes the experimental results of pipe whip on the pipe specimens of 4 inch in diameter under BWR condition on which the pressure is 6.77 MPa and the temperature is 285 0 C. The pipe specimens were 114.3 mm (4 inch) in diameter and 8.6 mm in thickness and 4500 mm in length. Four pipe whip restraints used in the tests were the U-bar type of 8 mm in diameter and fabricated from type 304 stainless steel. The experimental parameter was the clearance (30, 50 and 100 mm). The dynamic strain behavior of the pipe specimen and the restraints was investigated by strain gages and their residual deformation was obtained by measuring marking points provided on their surface. The Pressure-time history in the pipe specimens was also obtained by pressure gages. The maximum pipe strain is caused near the restraints and increases with increase of the clearance. The experimental results of pipe whip tests indicate the effectiveness of pipe whip restraints. The ratio of absorbed strain energy of the pipe specimen to that of the restraints is nearly constant for different clearances at the overhang length of 400 mm. (author)

  15. Pipe-to-pipe impact tests

    Energy Technology Data Exchange (ETDEWEB)

    Bampton, M C.C.; Alzheimer, J M; Friley, J R; Simonen, F A

    1985-11-01

    Existing licensing criteria express what damage shall be assumed for various pipe sizes as a consequence of a postulated break in a high energy system. The criteria are contained in Section 3.6.2 of the Standard Review Plan, and the purpose of the program described with this paper is to evaluate the impact criteria by means of a combined experimental and analytical approach. A series of tests has been completed. Evaluation of the test showed a deficiency in the range of test parameters. These deficiencies are being remedied by a second series of tests and a more powerful impact machine. A parallel analysis capability has been developed. This capability has been used to predict the damage for the first test series. The quality of predictions has been improved by tests that establish post-crush and bending relationships. Two outputs are expected from this project: data that may, or may not, necessitate changes to the criteria after appropriate value impact evaluations and an analytic capability for rapidly evaluating the potential for pipe whip damage after a postulated break. These outputs are to be contained in a value-impact document and a program final report. (orig.).

  16. Heat pipes as perspective base elements of heat recovery in heat supply and ventilating systems

    Directory of Open Access Journals (Sweden)

    Matveev Andrey

    2017-01-01

    Full Text Available Thermotechnical characteristics of heat pipes are considered as high-efficient heat-transfer devices, which can provide energy-saving technologies for heat supply and ventilating systems and for different branches of industry. Thermotechnical and working (”performance capability” characteristics of heat pipes are investigated. By ”performance capability” of heat pipes and heat-transfer devices on heat pipes we mean the system state, where it can perform set functions and keep parameter values (thermal power, conductivity, thermal resistance, heat-transfer coefficient, temperature level and differential, etc. within the regulations of standardized specifications. The article presents theoretical and experimental methods of «gaslock» length determination on noncondensable gases during long-lasting tests of ammonia heat pipes made of aluminum shape АS – КRА 7.5 – R1 (alloy АD – 31. The paper gives results of research of thermotechnical characteristics of heat pipes in horizontal and vertical states (separate and as a set part while using different systems of thermal insulation. The obtained results of thermotechnical and resource tests show the advantages of ammonia heat pipes as basic elements for heat exchanger design in heating and ventilation systems.

  17. Pipe whip: a summary of the damage observed in BNL pipe-on-pipe impact tests

    International Nuclear Information System (INIS)

    Baum, M.R.

    1987-01-01

    This paper describes examples of the damage resulting from the impact of a whipping pipe on a nearby pressurised pipe. The work is a by-product of a study of the motion of a whipping pipe. The tests were conducted with small-diameter pipes mounted in rigid supports and hence the results are not directly applicable to large-scale plant applications where flexible support mountings are employed. The results illustrate the influence of whipping pipe energy, impact position and support type on the damage sustained by the target pipe. (author)

  18. Seismic proving test of ultimate piping strength (current status of preliminary tests)

    International Nuclear Information System (INIS)

    Suzuki, K.; Namita, Y.; Abe, H.; Ichihashi, I.; Suzuki, K.; Ishiwata, M.; Fujiwaka, T.; Yokota, H.

    2001-01-01

    In 1998 Fiscal Year, the 6 year program of piping tests was initiated with the following objectives: i) to clarify the elasto-plastic response and ultimate strength of nuclear piping, ii) to ascertain the seismic safety margin of the current seismic design code for piping, and iii) to assess new allowable stress rules. In order to resolve extensive technical issues before proceeding on to the seismic proving test of a large-scale piping system, a series of preliminary tests of materials, piping components and simplified piping systems is intended. In this paper, the current status of the material tests and the piping component tests is reported. (author)

  19. Applicability of ANSYS ELBOW290 element for flexibility calculation of tight radius bends on feeder pipes in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, X., E-mail: Xuan.Zhang@candu.com [Candu Energy Inc, Mississauga, ON (Canada)

    2015-07-01

    A curved pipe element, ELBOW290, became available in ANSYS 12. This element was developed based on a simplified shell theory, and maintains the ability to capture cross-sectional deformations of elbows. Numerical testing on the applicability of this element for the flexibility calculation of the tight radius bends in CANDU reactors is carried out to determine the usability of this element in completing stress analyses for feeder pipes. Comparisons are made between the ELBOW290 and the shell element for various feeder bend types found in domestic and overseas CANDU reactors. The comparisons show that the ELBOW290 element is suitable for calculating the flexibility of the tight radius bends. (author)

  20. Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings

    International Nuclear Information System (INIS)

    Brekow, G.; Wuestenberg, H.; Hesselmann, H.; Rathgeb, W.

    1991-01-01

    Ultrasonic testing with the phased array method at the pipe connection inner edges in pipings. The pipe connection inner corner tests in feedwater lines to the main coolant pipe were carried out by Preussen-Elektra in cooperation with Siemens KWU and the BAM with the ultrasonic phased array method. The testing plan was developed by means of a computed model. For a trial of the testing plan, numerous ultrasonic measurements with the phased array method were carried out using a pipe test piece with TH-type inner edges, which was a 1:1 model of the reactor component to be tested. The data measured at several test notches in the pipe connection inner edge area covered by a plating of 6 mm were analyzed. (orig./MM) [de

  1. Pipe connector

    International Nuclear Information System (INIS)

    Sullivan, T.E.; Pardini, J.A.

    1978-01-01

    A safety test facility for testing sodium-cooled nuclear reactor components includes a reactor vessel and a heat exchanger submerged in sodium in the tank. The reactor vessel and heat exchanger are connected by an expansion/deflection pipe coupling comprising a pair of coaxially and slidably engaged tubular elements having radially enlarged opposed end portions of which at least a part is of spherical contour adapted to engage conical sockets in the ends of pipes leading out of the reactor vessel and in to the heat exchanger. A spring surrounding the pipe coupling urges the end portions apart and into engagement with the spherical sockets. Since the pipe coupling is submerged in liquid a limited amount of leakage of sodium from the pipe can be tolerated

  2. MODEL TESTS AND 3D ELASTIC FINITE ELEMENT ANALYSIS FOR STEEL PIPE PILES WITH WINGS IN STALLED IN SOIL CEMENT COLUMN

    Science.gov (United States)

    Tamai, Toshiyuki; Teramoto, Shuntarou; Kimura, Makoto

    Steel pipe piles with wings installed in soil cement column is a composite foundation of pile consisting of soil improvement with cement and steel pipe with wings. This type of pile shows higher vertical bearing capacity when compared to steel pipe piles that are installed without soil cement. It is thought the wings contribute to higher bearing capacity of this type of piles. The wings are also thought to play the role of structural unification of pile foundations and load transfer. In this study, model test and 3D elastic finite element analysis was carried out in order to elucidate the effect of wings on the structural unification of pile foundation and the load transfer mechanism. Firstly, the model test was carried out in order to grasp the influence of pile with and without wings, the shape of wings of the pile and the unconfined compression strength of the soil cement on the structural unification of the pile foundation. The numerical analysis of the model test was then carried out on the intermediate part of the pile foundation with wings and mathematical model developed. Finally load tran sfer mechanism was checked for the entire length of the pile through this mathematical model and the load sharing ratio of the wings and stress distribution occurring in the soil cement clarified. In addition, the effect of the wing interval on the structural unification of the pile foundation and load transfer was also checked and clarified.

  3. Pipe robots for internal inspection, non-destructive testing and machining of pipelines

    International Nuclear Information System (INIS)

    Reiss, Alexander

    2016-01-01

    Inspector Systems is a specialist in manufacturing of tethered self-propelled pipe robots for internal inspection, non-destructive testing and machining of pipeline systems. Our industrial sectors, which originates from 30 year experience in the nuclear industry, are Gas and Oil (On-/Offshore, Refineries), Chemical, Petrochemical, Water etc. The pipe robots are able to get inserted through poor access points (e.g. valves) and to pass in bi-directional travelling vertical sections and numerous bends with small arc radius. The paper describes the system concept and performance of the pipe robot technology. A modular construction allows to equip the robots with different operational elements for the respective application.

  4. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  5. Comparison of a nonlinear dynamic model of a piping system to test data

    International Nuclear Information System (INIS)

    Blakely, K.D.; Howard, G.E.; Walton, W.B.; Johnson, B.A.; Chitty, D.E.

    1983-01-01

    Response of a nonlinear finite element model of the Heissdampfreaktor recirculation piping loop (URL) was compared to measured data, representing the physical benchmarking of a nonlinear model. Analysis-test comparisons of piping response are presented for snapback tests that induced extreme nonlinear behavior of the URL system. Nonlinearities in the system are due to twelve swaybraces (pipe supports) that possessed nonlinear force-deflection characteristics. These nonlinearities distorted system damping estimates made by using the half-power bandwidth method on Fourier transforms of measured accelerations, with the severity of distortion increasing with increasing degree of nonlinearity. Time domain methods, which are not so severely affected by the presence of nonlinearities, were used to compute system damping ratios. Nonlinear dynamic analyses were accurately and efficiently performed using the pseudo-force technique and the finite element program MSC/NASTRAN. Measured damping was incorporated into the model for snapback simulations. Acceleration time histories, acceleration Fourier transforms, and swaybrace force time histories of the nonlinear model, plus several linear models, were compared to test measurements. The nonlinear model predicted three-fourths of the measured peak accelerations to within 50%, half of the accelerations to within 25%, and one-fifth of the accelerations to within 10%. This nonlinear model predicted accelerations (in the time and frequency domains) and swaybrace forces much better than did any of the linear models, demonstrating the increased accuracy resulting from properly simulating nonlinear support behavior. In addition, earthquake response comparisons were made between the experimentally validated nonlinear model and a linear model. Significantly lower element stresses were predicted for the nonlinear model, indicating the potential usefulness of nonlinear simulations in piping design assessments. (orig.)

  6. Stress analysis of the O-element pipe during the process of flue gases purification

    Directory of Open Access Journals (Sweden)

    Nekvasil R.

    2008-11-01

    Full Text Available Equipment for flue gases purification from undesired substances is used throughout power and other types of industry. This paper deals with damaging of the O-element pipe designed to remove sulphur from the flue gases, i.e. damaging of the pipe during flue gases purification. This purification is conducted by spraying the water into the O-shaped pipe where the flue gases flow. Thus the sulphur binds itself onto the water and gets removed from the flue gas. Injection of cold water into hot flue gases, however, causes high stress on the inside of the pipe, which can gradually damage the O-element pipe. In this paper initial injection of water into hot pipe all the way to stabilization of temperature fields will be analyzed and the most dangerous places which shall be considered for fatigue will be determined.

  7. ADIMEW: Fracture assessment and testing of an aged dissimilar metal weld pipe assembly

    International Nuclear Information System (INIS)

    Wintle, J.B.; Hayes, B.; Goldthorpe, M.R.

    2004-01-01

    ADIMEW (Assessment of Aged Piping Dissimilar Metal Weld Integrity) was a three-year collaborative research programme carried out under the EC 5th Framework Programme. The objective of the study was to advance the understanding of the behaviour and safety assessment of defects in dissimilar metal welds between pipes representative of those found in nuclear power plant. ADIMEW studied and compared different methods for predicting the behaviour of defects located near the fusion boundaries of dissimilar metal welds typically used to join sections of austenitic and ferritic piping operating at high temperature. Assessment of such defects is complicated by issues that include: severe mis-match of yield strength of the constituent parent and weld metals, strong gradients of material properties, the presence of welding residual stresses and mixed mode loading of the defect. The study includes the measurement of material properties and residual stresses, predictive engineering analysis and validation by means of a large-scale test. The particular component studied was a 453mm diameter pipe that joins a section of type A508 Class 3 ferritic pipe to a section of type 316L austenitic pipe by means of a type 308 austenitic weld with type 308/309L buttering laid on the ferritic pipe. A circumferential, surface-breaking defect was cut using electro discharge machining into the 308L/309L weld buttering layer parallel to the fusion line. The test pipe was subjected to four-point bending to promote ductile tearing of the defect. This paper presents the results of TWI contributions to ADIMEW including: fracture toughness testing, residual stress measurements and assessments of the ADIMEW test using elastic-plastic, cracked body, finite element analysis. (orig.)

  8. Ring tests on high density polyethylene: Full investigation assisted by finite element modeling

    International Nuclear Information System (INIS)

    Laiarinandrasana, L.; Devilliers, C.; Oberti, S.; Gaudichet, E.; Fayolle, B.; Lucatelli, J.M.

    2011-01-01

    In order to characterize the mechanical behavior of HDPE pipes, the ASTM D 2290-04 standard recommends carrying out tensile tests on notched rings, cut out from the pipe. This very simple test is also utilized to investigate the aging effect of the pipe by determining the strain at failure. Comparison between full ring and notched ring mechanical responses are discussed. Constitutive modeling including strain rate effects was performed by finite element analysis. This allowed a better understanding of the stress state in the cross section perpendicular to the loading direction. Additionally, the influence of a thin layer of oxidized HDPE in the inner wall of the ring was studied in the light of the finite element results.

  9. Detecting failed elements on phased array ultrasound transducers using the Edinburgh Pipe Phantom

    Science.gov (United States)

    Inglis, Scott; Pye, Stephen D

    2016-01-01

    Aims Imaging faults with ultrasound transducers are common. Failed elements on linear and curvilinear array transducers can usually be detected with a simple image uniformity or ‘paperclip’ test. However, this method is less effective for phased array transducers, commonly used in cardiac imaging. The aim of this study was to assess whether the presence of failed elements could be detected through measurement of the resolution integral (R) using the Edinburgh Pipe Phantom. Methods A 128-element paediatric phased array transducer was studied. Failed elements were simulated using layered polyvinyl chloride (PVC) tape as an attenuator and measurements of resolution integral were carried out for several widths of attenuator. Results All widths of attenuator greater than 0.5 mm resulted in a significant reduction in resolution integral and low contrast penetration measurements compared to baseline (p tests to detect failed elements on phased array transducers. Particularly encouraging is the result for low contrast penetration as this is a quick and simple measurement to make and can be performed with many different test objects, thus enabling ‘in-the-field’ checks. PMID:27482276

  10. Niobium 1 percent zirconium/potassium and titanium/potassium life-test heat pipe design and testing

    Science.gov (United States)

    Sena, J. Tom; Merrigan, Michael A.

    Experimental lifetime performance studies currently in progress use Niobium 1 percent Zirconium (Nb-1Zr) and Titanium (Ti) heat pipes with potassium (K) as the working fluid. A heat pipe life test matrix was developed for testing the heat pipes. Because the corrosion rates in alkali metal heat pipes are affected by temperature and working fluid evaporation flux, the variable parameters of the experimental matrix are established as steady operating temperature and input heat flux density. Total impurity inventory is a factor in corrosion rate so impurity levels are being evaluated in the heat pipe materials before and after testing. Eight Nb-1Zr/K heat pipes were designed, fabricated, and tested. Two of the heat pipes have completed testing whereas the other six are currently in test. These are gravity assist heat pipes operating in a reflux mode. The heat pipes are tested by sets, one set of two and two sets of three heat pipes. Three Ti/K heat pipes are also in life test. These heat pipes are tested as a set in a horizontal position in a capillary pumped annular flow mode. Each of the heat pipes is encapsulated in a quartz vacuum container with a water calorimeter over the vacuum container for power throughput measurements. Thermocouples are attached to the heat pipes for measuring temperature. Heat input to the heat pipes is via an RF coil. The heat pipes are operating at between 800 and 900 K, with heat input fluxes of 13.8 to 30 W/sq cm. Of the Nb-1Zr/K heat pipes, two of the heat pipes have been in operation for 14,000 hours, three over 10,000 hours, and three over 7,000 hours. The Ti/K heat pipes have been in operation for 1,266 hours.

  11. Fatigue analysis of flexible pipes using alternative element types and bend stiffener data

    OpenAIRE

    Chen, Minghao

    2011-01-01

    The flexible pipe is a vital part of a floating production system. The lifetime of a flexible riser system is crucial for the Health Safety and Environment (HSE) management. As a result of this, it is very necessary to carry out research on the lifetime of flexible pipe. In this thesis we formalized analysis on flexible pipes, utilizing the finite element analysis software BFLEX 2010, developed by MARINTEK. Chapter 1 describes basic knowledge about flexible pipe and relevant facilities. C...

  12. Bayesian analysis of heat pipe life test data for reliability demonstration testing

    International Nuclear Information System (INIS)

    Bartholomew, R.J.; Martz, H.F.

    1985-01-01

    The demonstration testing duration requirements to establish a quantitative measure of assurance of expected lifetime for heat pipes was determined. The heat pipes are candidate devices for transporting heat generated in a nuclear reactor core to thermoelectric converters for use as a space-based electric power plant. A Bayesian analysis technique is employed, utilizing a limited Delphi survey, and a geometric mean accelerated test criterion involving heat pipe power (P) and temperature (T). Resulting calculations indicate considerable test savings can be achieved by employing the method, but development testing to determine heat pipe failure mechanisms should not be circumvented

  13. Seismic testing and analysis of a prototypic nonlinear piping system

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.

    1982-11-01

    A series of seismic tests and analyses of a nonlinear Fast Flux Test Facility (FFTF) prototypic piping system are described, and measured responses are compared with analytical predictions. The test loop was representative of a typical LMFBR insulated small bore piping system and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps. Various piping support configurations were tested and analyzed to evaluate the effects of free play and other nonlinear stiffness characteristics on the piping system response

  14. Investigation into the cause of leak in the pipe of the corrosion test apparatus of IS process

    International Nuclear Information System (INIS)

    Tanaka, Nobuyuki; Furukawa, Tomohiro; Inagaki, Yoshiyuki; Suwa, Hirokazu

    2008-12-01

    The thermochemical water-splitting hydrogen production IS process utilizes corrosive chemicals such as sulfuric acid and hydriodic acid. Corrosion tests in IS process environments have been carried out to get the corrosion data of materials. In the corrosion test in 90wt% sulfuric acid at 400degC, the leak of sulfuric acid was observed in a pipe connected with a reflux condenser. The cause of the leakage is a significant knowledge for the operation of the test apparatus. Therefore the cause was investigated. A 1mm wide through hole was detected in the pipe around the welding bead. By visual observation after cutting the pipe, the wall thickness of the pipe became thin at the inside welding bead around the through hole. In addition, EMPA showed that the inhomogeneous distribution of the constituent elements of the pipe was observed around the through hole. For these reasons, it is estimated that the lowering of the corrosion resistance by the sensitization at the welding caused the leakage. (author)

  15. BNL NONLINEAR PRE TEST SEISMIC ANALYSIS FOR THE NUPEC ULTIMATE STRENGTH PIPING TEST PROGRAM

    International Nuclear Information System (INIS)

    DEGRASSI, G.; HOFMAYER, C.; MURPHY, C.; SUZUKI, K.; NAMITA, Y.

    2003-01-01

    The Nuclear Power Engineering Corporation (NUPEC) of Japan has been conducting a multi-year research program to investigate the behavior of nuclear power plant piping systems under large seismic loads. The objectives of the program are: to develop a better understanding of the elasto-plastic response and ultimate strength of nuclear piping; to ascertain the seismic safety margin of current piping design codes; and to assess new piping code allowable stress rules. Under this program, NUPEC has performed a large-scale seismic proving test of a representative nuclear power plant piping system. In support of the proving test, a series of materials tests, static and dynamic piping component tests, and seismic tests of simplified piping systems have also been performed. As part of collaborative efforts between the United States and Japan on seismic issues, the US Nuclear Regulatory Commission (USNRC) and its contractor, the Brookhaven National Laboratory (BNL), are participating in this research program by performing pre-test and post-test analyses, and by evaluating the significance of the program results with regard to safety margins. This paper describes BNL's pre-test analysis to predict the elasto-plastic response for one of NUPEC's simplified piping system seismic tests. The capability to simulate the anticipated ratcheting response of the system was of particular interest. Analyses were performed using classical bilinear and multilinear kinematic hardening models as well as a nonlinear kinematic hardening model. Comparisons of analysis results for each plasticity model against test results for a static cycling elbow component test and for a simplified piping system seismic test are presented in the paper

  16. Analysis of piping systems by finite element method using code SAP-IV

    International Nuclear Information System (INIS)

    Cizelj, L.; Ogrizek, D.

    1987-01-01

    Due to extensive and multiple use of the computer code SAP-IV we have decided to install it on VAX 11/750 machine. Installation required a large quantity of programming due to great discrepancies between the CDC (the original program version) and the VAX. Testing was performed basically in the field of pipe elements, based on a comparison between results obtained with the codes PSAFE2, DOCIJEV, PIPESD and SAP -V. Besides, the model of reactor pressure vessel with 3-D thick shell elements was done. The capabilities show good agreement with the results of other programs mentioned above. Along with the package installation, the graphical postprocessors being developed for mesh plotting. (author)

  17. Seismic test of high temperature piping for HTGR

    International Nuclear Information System (INIS)

    Kobatake, Kiyokazu; Midoriyama, Shigeru; Ooka, Yuzi; Suzuki, Michiaki; Katsuki, Taketsugu

    1983-01-01

    Since the high temperature pipings for the high temperature gas-cooled reactor contain helium gas at 1000 deg C and 40 kgf/cm 2 , the double-walled pipe type consisting of the external pipe serving as the pressure boundary and the internal pipe with heat insulating structure was adopted. Accordingly, their aseismatic design is one of the important subjects. Recently, for the purpose of grasping the vibration characteristics of these high temperature pipings and obtaining the data required for the aseismatic design, two specimens, that is, a double-walled pipe model and a heat-insulating structure, were made, and the vibration test was carried out on them, using a 30 ton vibration table of Kawasaki Heavy Industries Ltd. In the high temperature pipings of the primary cooling system for the multi-purpose, high temperature gas-cooled experimental reactor, the external pipes of 32 B bore as the pressure boundary and the internal pipes of 26 B bore with internal heat insulation consisting of double layers of fiber and laminated metal insulators as the temperature boundary were adopted. The testing method and the results are reported. As the spring constant of spacers is larger and clearance is smaller, the earthquake wave response of double-walled pipes is smaller, and it is more advantageous. The aseismatic property of the heat insulation structure is sufficient. (Kako, I.)

  18. Application of the cracked pipe element to creep crack growth prediction

    Energy Technology Data Exchange (ETDEWEB)

    Brochard, J.; Charras, T.

    1997-04-01

    The modification of a computer code for leak before break analysis is very briefly described. The CASTEM2000 code was developed for ductile fracture assessment of piping systems with postulated circumferential through-wall cracks under static or dynamic loading. The modification extends the capabilities of the cracked pipe element to the determination of fracture parameters under creep conditions (C*, {phi}c and {Delta}c). The model has the advantage of evaluating significant secondary effects, such as those from thermal loading.

  19. Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

    International Nuclear Information System (INIS)

    Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

    2000-01-01

    The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

  20. Internal testing of pipe systems with IRIS inspection system

    International Nuclear Information System (INIS)

    1986-01-01

    The internal piping inspection system IRIS allows inside testing of pipes with an internal diameter of NW 70 as a minimum, and of any horizontal or vertical layout of the piping system. Visual testing is done by means of an integrated CCD video system with high resolution power. Technical data are given and examples of applications, in the German and English language. (DG) [de

  1. Critical element development of standard pipe connector for remote handling

    International Nuclear Information System (INIS)

    Taguchi, Kou; Kakudate, Satoshi; Kanamori, Naokazu; Oka, Kiyoshi; Nakahira, Masataka; Obara, Kenjiro; Tada, Eisuke; Shibanuma, Kiyoshi; Seki, Masahiro

    1994-08-01

    In fusion experimental reactors such as ITER, the in-vessel components such as blanket and divertor are actively cooled and a large number of cooling pipes are located around the core of reactor, where personnel access is prohibited. Mechanical pipe connectors are highly required as standard components for easy and reliable connection/disconnection of cooling pipe by remote handling. For this purpose, a clamping chain type connector has been developed with special mechanisms such as plate springs and guide structures so as to enable concentric and axial movement of clamping chain for easy mounting and dismounting. The basic performance test of a prototypical connector for a 80-A pipe shows sufficient leak tightness and proof pressure capability as well as simple connection/disconnection operation. In addition to the clamp chain type connector, design efforts have been made to develop a quick coupling type connector and a preliminary model of air-actuated quick connector has been fabricated for further investigations. This paper gives the design concept of mechanical pipe connectors such as clamping chain type and quick coupler type, and the basic performance tests results of clamping chain type connector. (author)

  2. Structural dynamics and fracture mechanics calculations of the behaviour of a DN 425 test piping system subjected to transient loading by water hammer

    International Nuclear Information System (INIS)

    Kussmaul, K.; Kobes, E.; Diem, H.; Schrammel, D.; Brosi, S.

    1994-01-01

    Within the scope of the German HDR safety programme, several tests were carried out to investigate transient pipe loading initiated by a simulated double-ended guillotine break event, and subsequent closure of a feedwater check valve (water hammer, blow-down). Numerical analyses by means of finite element programmes were performed in parallel to the experiments. Using water hammer tests of a DN 425 piping system with predamaged components, the procedure of such analyses will be demonstrated. The results are presented, beginning with structural dynamic calculations of the undamaged piping; followed by coupling of structural dynamics and fracture mechanics computations with simple flaw elements (line spring); and finishing with costly three-dimensional fracture mechanics analyses. A good description of the real piping behaviour can be made by the numerical methods, even in the case of high plastification processes. ((orig.))

  3. High temperature heat pipe experiments in low earth orbit

    International Nuclear Information System (INIS)

    Woloshun, K.; Merrigan, M.A.; Sena, J.T.; Critchley, E.

    1993-01-01

    Although high temperature, liquid metal heat pipe radiators have become a standard component on most high power space power system designs, there is no experimental data on the operation of these heat pipes in a zero gravity or micro-gravity environment. Experiments to benchmark the transient and steady state performance of prototypical heat pipe space radiator elements are in preparation for testing in low earth orbit. It is anticipated that these heat pipes will be tested aborad the Space Shuttle in 1995. Three heat pipes will be tested in a cargo bay Get Away Special (GAS) canister. The heat pipes are SST/potassium, each with a different wick structure; homogeneous, arterial, and annular gap, the heat pipes have been designed, fabricated, and ground tested. In this paper, the heat pipe designs are specified, and transient and steady-state ground test data are presented

  4. Analysis of the jet pipe electro-hydraulic servo valve with finite element methods

    Directory of Open Access Journals (Sweden)

    Kaiyu Zhao

    2018-01-01

    Full Text Available The dynamic characteristics analysis about the jet pipe electro-hydraulic servo valve based on experience and mathematical derivation was difficult and not so precise. So we have analysed the armature feedback components, torque motor and jet pipe receiver in electrohydraulic servo valve by sophisticated finite element analysis tools respectively and have got physical meaning data on these parts. Then the data were fitted by Matlab and the mathematical relationships among them were calculated. We have done the dynamic multi-physical fields’ Simulink co-simulation using above mathematical relationship, and have got the input-output relationship of the overall valve, the frequency response and step response. This work can show the actual working condition accurately. At the same time, we have considered the materials and the impact of the critical design dimensions in the finite element analysis process. It provides some new ideas to the overall design of jet pipe electro-hydraulic servo valve.

  5. Fatigue test results of straight pipe with flaws in inner surface

    International Nuclear Information System (INIS)

    Shibata, Katsuyuki; Oba, Toshihiro; Kawamura, Takaichi; Yokoyama, Norio; Miyazono, Shohachiro

    1981-01-01

    Fatigue and fracture tests of piping models with flaws in the inner surface were carried out to investigate the fatigue crack growth, coalescence of multiple cracks and fracture behavior. Two straight test pipes with and without weldment in the test section of SUS304L stainless steel were tested under almost the same test conditions. Three artificial defects were machined in the inner surface of the test section of the test pipes. The fatigue test were performed untill the cracks coalesced and grew through the thickness. Subsequently, a static load was imposed on test pipe which contained a large crack in the test section. The test results show that the fatigue crack growth is slower than that predicted by the method specified in the Section XI of ASME Boiler and Pressure Vessel Code, and that the test pipes can endure more than the static load of 3Sm without an unstable fracture. (author)

  6. Correlation of analysis with high level vibration test results for primary coolant piping

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.; Costello, J.F.

    1992-01-01

    Dynamic tests on a modified 1/2.5-scale model of pressurized water reactor (PWR) primary coolant piping were performed using a large shaking table at Tadotsu, Japan. The High Level Vibration Test (HLVT) program was part of a cooperative study between the United States (Nuclear Regulatory Commission/Brookhaven National Laboratory, NRC/BNL) and Japan (Ministry of International Trade and Industry/Nuclear Power Engineering Center). During the test program, the excitation level of each test run was gradually increased up to the limit of the shaking table and significant plastic strains, as well as cracking, were induced in the piping. To fully utilize the test results, NRC/BNL sponsored a project to develop corresponding analytical predictions for the nonlinear dynamic response of the piping for selected test runs. The analyses were performed using both simplified and detailed approaches. The simplified approaches utilize a linear solution and an approximate formulation for nonlinear dynamic effects such as the use of a deamplification factor. The detailed analyses were performed using available nonlinear finite element computer codes, including the MARC, ABAQUS, ADINA and WECAN codes. A comparison of various analysis techniques with the test results shows a higher prediction error in the detailed strain values in the overall response values. A summary of the correlation analyses was presented before the BNL. This paper presents a detailed description of the various analysis results and additional comparisons with test results

  7. Fatigue analysis of aluminum drill pipes

    Directory of Open Access Journals (Sweden)

    João Carlos Ribeiro Plácido

    2005-12-01

    Full Text Available An experimental program was performed to investigate the fundamental fatigue mechanisms of aluminum drill pipes. Initially, the fatigue properties were determined through small-scale tests performed in an optic-mechanical fatigue apparatus. Additionally, full-scale fatigue tests were carried out with three aluminum drill pipe specimens under combined loading of cyclic bending and constant axial tension. Finally, a finite element model was developed to simulate the stress field along the aluminum drill pipe during the fatigue tests and to estimate the stress concentration factors inside the tool joints. By this way, it was possible to estimate the stress values in regions not monitored during the fatigue tests.

  8. Significance of high level test data in piping design

    International Nuclear Information System (INIS)

    McLean, J.L.; Bitner, J.L.

    1991-01-01

    During the 1980's the piping technical community in the U.S. initiated a series of research activities aimed at reducing the conservatism inherent in nuclear piping design. One of these activities was directed at the application of the ASME Code rules to the design of piping subjected to dynamic loads. This paper surveys the test data obtained from three groups in the U.S. and none in the U.K., and correlates the findings as they relate to the failure modes of piping subjected to seismic loads. The failure modes experienced as the result of testing at dynamic loads significantly in excess of anticipated loads specified for any of the ASME Code service levels are discussed. A recommendation is presented for modifying the Code piping rules to reduce the conservatism inherent in seismic design

  9. Guided waves dispersion equations for orthotropic multilayered pipes solved using standard finite elements code.

    Science.gov (United States)

    Predoi, Mihai Valentin

    2014-09-01

    The dispersion curves for hollow multilayered cylinders are prerequisites in any practical guided waves application on such structures. The equations for homogeneous isotropic materials have been established more than 120 years ago. The difficulties in finding numerical solutions to analytic expressions remain considerable, especially if the materials are orthotropic visco-elastic as in the composites used for pipes in the last decades. Among other numerical techniques, the semi-analytical finite elements method has proven its capability of solving this problem. Two possibilities exist to model a finite elements eigenvalue problem: a two-dimensional cross-section model of the pipe or a radial segment model, intersecting the layers between the inner and the outer radius of the pipe. The last possibility is here adopted and distinct differential problems are deduced for longitudinal L(0,n), torsional T(0,n) and flexural F(m,n) modes. Eigenvalue problems are deduced for the three modes classes, offering explicit forms of each coefficient for the matrices used in an available general purpose finite elements code. Comparisons with existing solutions for pipes filled with non-linear viscoelastic fluid or visco-elastic coatings as well as for a fully orthotropic hollow cylinder are all proving the reliability and ease of use of this method. Copyright © 2014 Elsevier B.V. All rights reserved.

  10. Pipe supports and anchors - LMFBR applications

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1983-06-01

    Pipe design and support design can not be treated as separate disciplines. A coordinated design approach is required if LMFBR pipe system adequacy is to be achieved at a reasonable cost. It is particularly important that system designers understand and consider those factors which influence support train flexibility and thus the pipe system dynamic stress levels. The system approach must not stop with the design phase but should continue thru the erection and acceptance test procedures. The factors that should be considered in the design of LMFBR pipe supports and anchors are described. The various pipe support train elements are described together with guidance on analysis, design and application aspects. Post erection acceptance and verification test procedures are then discussed

  11. Leak test of the pipe line for radioactive liquid waste

    International Nuclear Information System (INIS)

    Machida, Chuji; Mori, Shoji.

    1976-01-01

    In the Tokai Research Establishment, most of the radioactive liquid waste is transferred to a wastes treatment facility through pipe lines. As part of the pipe lines a cast iron pipe for town gas is used. Leak test has been performed on all joints of the lines. For the joints buried underground, the test was made by radioactivity measurement of the soil; and for the joints in drainage ditch by the pressure and bubble methods. There were no leakage at all, indicating integrity of all the joints. On the other hand, it is also known by the other test that the corrosion of inner surface of the piping due to liquid waste is only slight. The pipe lines for transferring radioactive liquid waste are thus still usable. (auth.)

  12. The nature thickness pipe element testing method to validate the application of LBB conception

    Energy Technology Data Exchange (ETDEWEB)

    Vasilchenko, G.S.; Artemyev, V.I.; Merinov, G.N. [and others

    1997-04-01

    To validate the application of leak before break analysis to the VVER-1000 reactor, a procedure for testing a large-scale specimen on electrohydraulic machinery was developed. Steel pipe with a circular weld and stainless cladding inside was manufactured and large-scale longitudinal cross-sections were cut. The remaining parts of the weld after cut out were used to determination standard tensile mechanical properties, critical temperature of brittlness and for manufacture of compact specimens. Experimental mechanical properties of the weld are summarized.

  13. The nature thickness pipe element testing method to validate the application of LBB conception

    International Nuclear Information System (INIS)

    Vasilchenko, G.S.; Artemyev, V.I.; Merinov, G.N.

    1997-01-01

    To validate the application of leak before break analysis to the VVER-1000 reactor, a procedure for testing a large-scale specimen on electrohydraulic machinery was developed. Steel pipe with a circular weld and stainless cladding inside was manufactured and large-scale longitudinal cross-sections were cut. The remaining parts of the weld after cut out were used to determination standard tensile mechanical properties, critical temperature of brittlness and for manufacture of compact specimens. Experimental mechanical properties of the weld are summarized

  14. Inelastic finite element analysis of a pipe-elbow assembly (benchmark problem 2)

    Energy Technology Data Exchange (ETDEWEB)

    Knapp, H P [Internationale Atomreaktorbau GmbH (INTERATOM) Bergisch Gladbach (Germany); Prij, J [Netherlands Energy Research Foundation (ECN) Petten (Netherlands)

    1979-06-01

    In the scope of the international benchmark problem effort on piping systems, benchmark problem 2 consisting of a pipe elbow assembly, subjected to a time dependent in-plane bending moment, was analysed using the finite element program MARC. Numerical results are presented and a comparison with experimental results is made. It is concluded that the main reason for the deviation between the calculated and measured values is due to the fact that creep-plasticity interaction is not taken into account in the analysis. (author)

  15. Methodology for Life Testing of Refractory Metal / Sodium Heat Pipes

    International Nuclear Information System (INIS)

    Martin, James J.; Reid, Robert S.

    2006-01-01

    This work establishes an approach to generate carefully controlled data to find heat pipe operating life with material-fluid combinations capable of extended operation. To accomplish this goal acceleration is required to compress 10 years of operational life into 3 years of laboratory testing through a combination of increased temperature and mass fluence. Specific test series have been identified, based on American Society for Testing and Materials (ASTM) specifications, to investigate long-term corrosion rates. The refractory metal selected for demonstration purposes is a molybdenum-44.5% rhenium alloy formed by powder metallurgy. The heat pipes each have an annular crescent wick formed by hot isostatic pressing of molybdenum-rhenium wire mesh. The heat pipes are filled by vacuum distillation with purity sampling of the completed assembly. Round-the-clock heat pipe tests with 6-month destructive and non-destructive inspection intervals are conducted to identify the onset and level of corrosion. Non-contact techniques are employed to provide power to the evaporator (radio frequency induction heating at 1 to 5 kW per heat pipe) and calorimetry at the condenser (static gas gap coupled water cooled calorimeter). The planned operating temperature range extends from 1123 to 1323 K. Accomplishments before project cancellation included successful development of the heat pipe wick fabrication technique, establishment of all engineering designs, baseline operational test requirements, and procurement/assembly of supporting test hardware systems. (authors)

  16. The IPIRG-1 pipe system fracture tests: Experimental results

    International Nuclear Information System (INIS)

    Scott, P.; Olson, R.J.; Wilkowski, G.M.

    1994-01-01

    As part of the First International Piping Integrity Research Group (IPIRG-1) program, six dynamic pipe system experiments were conducted. The objective of these experiments was to generate experimental data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system subjected to combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The pipe system evaluated was an expansion loop with over 30 m (100 feet) of 16-inch nominal diameter Schedule 100 pipe. The experimental facility was equipped with special hardware to ensure that system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe system experiments. The uncracked-pipe experiment was conducted to evaluate the piping system damping and natural frequency characteristics. The cracked-pipe experiments were conducted to evaluate the fracture behavior, piping system response, and fracture stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided the tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Key results from the six pipe system experiments and material characterization efforts are presented. Detailed analyses will be published in a companion paper

  17. Cryogenic and Gas System Piping Pressure Tests (A Collection of PT Permits)

    International Nuclear Information System (INIS)

    Rucinski, Russell A.

    2002-01-01

    This engineering note is a collection of pipe pressure testing documents for various sections of piping for the D-Zero cryogenic and gas systems. High pressure piping must conform with FESHM chapter 5031.1. Piping lines with ratings greater than 150 psig have a pressure test done before the line is put into service. These tests require the use of pressure testing permits. It is my intent that all pressure piping over which my group has responsibility conforms to the chapter. This includes the liquid argon and liquid helium and liquid nitrogen cryogenic systems. It also includes the high pressure air system, and the high pressure gas piping of the WAMUS and MDT gas systems. This is not an all inclusive compilation of test documentation. Some piping tests have their own engineering note. Other piping section test permits are included in separate safety review documents. So if it isn't here, that doesn't mean that it wasn't tested. D-Zero has a back up air supply system to add reliability to air compressor systems. The system includes high pressure piping which requires a review per FESHM 5031.1. The core system consists of a pressurized tube trailer, supply piping into the building and a pressure reducing regulator tied into the air compressor system discharge piping. Air flows from the trailer if the air compressor discharge pressure drops below the regulator setting. The tube trailer is periodically pumped back up to approximately 2000 psig. A high pressure compressor housed in one of the exterior buildings is used for that purpose. The system was previously documented, tested and reviewed for Run I, except for the recent addition of piping to and from the high pressure compressor. The following documents are provided for review of the system: (1) Instrument air flow schematic, drg. 3740.000-ME-273995 rev. H; (2) Component list for air system; (3) Pressure testing permit for high pressure piping; (4) Documentation from Run I contained in D-Zero Engineering note

  18. Finite element analysis of a fluid-structure interaction in flexible pipe ...

    African Journals Online (AJOL)

    The obtained mathematical system is constituted of four non-linear hyperbolic partial differential equations describing the wave propagation in both pipe wall and liquid flow. The fluid-structure interaction is found to be governed by Poisson's ratio. In this steady finite element method based on Galerkin formulation is applied.

  19. Test of Seal System for Flexible Pipe End Fitting

    DEFF Research Database (Denmark)

    Banke, Lars; Jensen, Thomas Gregers

    1999-01-01

    The purpose of the end fitting seal system is to ensure leak proof termination of flexible pipes. The seal system of an NKT end fitting normally consists of a number of ring joint gaskets mounted in a steel sleeve on the outside of the polymeric inner liner of the pipe. The seal system is activated...... by compression of the gaskets, thus using the geometry to establish a seal towards the inner liner of the pipe and the steel sleeve of the end fitting. This paper describes how the seal system of an end fitting can be tested using an autoclave. By regulating temperature and pressure, the seal system can...... be tested up to 130oC and 51.7 MPa. Pressure, temperature and the mechanical behaviours of the pipe are measured for use in further research. The set-up is used to test the efficiency of the seal system as function of parameters such as cross sectional shapes of the gaskets, tolerances between gaskets...

  20. PWR pressurizer discharge piping system on-site testing

    International Nuclear Information System (INIS)

    Anglaret, G.; Lasne, M.

    1983-08-01

    Framatome PWR systems includes the installation of safety valves and relief valves wich permit the discharge of steam from the pressurizer to the pressurizer relief tank through discharge piping system. Water seal expulsion pluration then depends on valve stem lift dynamics which can vary according to water-stem interaction. In order to approaches the different phenomenons, it was decided to perform a test on a 900 MWe French plant, test wich objectives are: characterize the mechanical response of the discharge piping to validate a mechanical model; open one, two or several valves among the following: one safety valve and three pilot operated relief valves, at a time or sequentially and measure the discharge piping transient response, the support loads, the

  1. Rupture disc opening property for using pipe rupture test in JAERI

    International Nuclear Information System (INIS)

    Kato, Rokuro

    1983-03-01

    In the Mechanical Strength and Structure Lab of JAERI there are being performed pipe break tests which are a postulated instantaneous guillotine break of the primary coolant piping in nuclear power plants. The test being performed are pipe whip tests and jet discharging tests. The bursting of the rupture disc is initiated by an electrical arc and is concluded by the internal pressure. Because the time characteristics during the opening of the rupture disc affects the dynamic thrust force of the pipe, it is necessary to measure these time characteristics. However, it is difficult to measure the conditions during this continuous opening because at the same time of the opening the high temperature and high pressure water is flashing. Therefore, the rupture disc opening was postulated on the measuring of the effective opening characteristics with electric contraction terminals which were attached to the inner surface of the test pipe downstream of the rupture disc and were extended toward the pipe centerline in a ring whose area is about 60 % of the area of the pipe flow sectional area. The measurement voltage was recorded when the data recorder was started in sequence with the electrical arc release from a trigger signal. As a result, it is evident that under high temperature and high pressure water the effective opening time is delayed by a few milliseconds. (author)

  2. An approximative solution for limit load of piping branch junctions with circumferential crack and finite element validation

    International Nuclear Information System (INIS)

    Xuan Fuzhen; Liu Changjun; Li Peining

    2005-01-01

    This paper is concerned with the prediction of limit load of the piping branch junctions with circumferential crack under internal pressure. Recently, we have developed a new approach for predicting the limit load of two-cylinder intersection structures with diameter ratio larger than 0.5, which has been successfully applied to defect free cases under various loading conditions. In the present work, we consider the extension of the approach to cover cracked piping branch junctions. On the basis of stress analysis in the vicinity of intersection line, a closed form of limit load solution for piping branch junctions with circumferential crack was developed. Then, 36 finite element (FE) models of piping branch junction with various dimensions of structure and crack were analyzed by using nonlinear finite element software. The limit loads from FE analysis and the proposed solution are compared with each other. Overall good agreement between the estimated solutions and the FE results provides confidence in the use of the proposed formulae for defect assessment of piping branch junctions in practice

  3. Performance testing of a hydrogen heat pipe

    International Nuclear Information System (INIS)

    Alario, J.; Kosson, R.

    1980-01-01

    Test results are presented for a reentrant groove heat pipe with hydrogen working fluid. The heat pipe became operational between 20 and 30 K after a cooldown from 77 K without any difficulty. Steady-state performance data taken over a 19 to 23 K temperature range indicated the following: (1) maximum heat transport capacity 5.4 W-m (2) static wicking height 1.42 cm and (3) overall heat pipe conductance 1.7 W/C. These data agreed remarkably well with extrapolations made from comparable ammonia test results. The maximum heat transport capacity is 9.5% larger than the extrapolated value, but the static wicking height is the same. The overall conductance is 29% of the ammonia value, which is close to the ratio of liquid thermal conductivities (24%). Also, recovery from a completely frozen condition was accomplished within 5 min by simply applying an evaporater heat load of 1.8 W

  4. Numerical ductile tearing simulation of circumferential cracked pipe tests under dynamic loading conditions

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Hyun Suk; Kim, Ji Soo; Ryu, Ho Wan; Kim, Yun Jae [Dept. of Mechanical Engineering, Korea University, Seoul (Korea, Republic of); Kim, Jin Weon [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2016-10-15

    This paper presents a numerical method to simulate ductile tearing in cracked components under high strain rates using finite element damage analysis. The strain rate dependence on tensile properties and multiaxial fracture strain is characterized by the model developed by Johnson and Cook. The damage model is then defined based on the ductility exhaustion concept using the strain rate dependent multiaxial fracture strain concept. The proposed model is applied to simulate previously published three cracked pipe bending test results under two different test speed conditions. Simulated results show overall good agreement with experimental results.

  5. Development, manufacturing and testing of a gas-loaded variable conductance methanol heat pipe

    Science.gov (United States)

    Vanbuggenum, R. I. J.; Daniels, D. H. W.

    1987-02-01

    The experimental technology required to measure the performance of moderate temperature heat pipes is presented. The heat pipe manufacturing process is described. The hydrodynamic characteristics of the porous structure inside the heat pipe envelope were examined using a specially developed test rig, based upon a steady-state evaporation test. A fully automated test facility was developed and validated by testing constant conductance and variable conductance heat pipes (VCHP). Theoretical performance predictions are illustrated in terms of pressure, depicted in 3D-plots, and compared with the test results of the heat pipe performance tests. The design of the VCHP was directed towards the verification of the VCHP mathematical model. The VCHP design is validated and ready for the final testing and model verification.

  6. Comparison Between Stress Obtained by Numerical Analysis and In-Situ Measurements on a Flexible Pipe Subjected to In-Plane Bending Test

    DEFF Research Database (Denmark)

    Vestergaard Lukassen, Troels; Glejbøl, Kristian; Lyckegaard, Anders

    2016-01-01

    to stress patterns obtained during in-situ OMS measurements carried out during an actual experimental inplane bending test. The study showed a good correlation between the stress variation predicted with the finite element model and the measured stress variation.......To predict the lifetime and long-term properties of tensile armour wires in a dynamically loaded pipe, it is essential to have a tool which allows detailed prediction of the stress variations in the tensile armour wires during global pipe loading. Furthermore, detailed understanding of the stress...... variations will allow for performance optimization of the armour layers. To study the detailed stress variations in flexible pipes during dynamic loading, a comprehensive three-dimensional implicit nonlinear finite element model has been developed. The predicted numerical stress variations will be compared...

  7. Inelastic analysis of Battelle-Columbus piping elbow creep test

    International Nuclear Information System (INIS)

    Dhalla, A.K.; Newman, S.Z.

    1979-01-01

    Analytical results are presented for room temperature and 593 deg. C creep bending deformation of a piping elbow structure tested at the Battelle-Columbus Laboratory. This analysis was performed in support of the International Piping Benchmark Problem Program being coordinated by ORNL. Results are presented for both simplified and refined structural models, and compared with test measurements reported by the Battelle-Columbus Laboratory. (author)

  8. Stands for testing the strength of welded pipe materials under the action of a corrosive medium

    Directory of Open Access Journals (Sweden)

    M.A. Kolodyi

    2017-12-01

    Full Text Available In order to study the features of the destruction of materials of pipelines for the transportation of oil, gas, products of processing of oil, water and other substances in the laboratory of the department of development of minerals named by prof. Bakka N.T. the complex of installations is invented, for which Ukrainian patents were obtained as utility models No. 30794, No. 52493, for the study of the working capacity of the elements of the listed pipeline systems in conditions that are as close as possible to the operational under the influence of the corrosive medium. Rotary vacuum devices were used as the basic elements of the proposed installations for testing the materials of the welded tubes for durability at single tensile and under flat stress conditions. The article presents the design of research stands for testing the durability of pipe materials and welds of pipelines using samples of materials and natural pipes (shortened under the influence of static, low cyclic and dynamic loads, and analyzes the influence of aggressive media.

  9. Steel Fibers Reinforced Concrete Pipes - Experimental Tests and Numerical Simulation

    Science.gov (United States)

    Doru, Zdrenghea

    2017-10-01

    The paper presents in the first part a state of the art review of reinforced concrete pipes used in micro tunnelling realised through pipes jacking method and design methods for steel fibres reinforced concrete. In part two experimental tests are presented on inner pipes with diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with metal fibres (35 kg / m3). In part two experimental tests are presented on pipes with inner diameters of 1410mm and 2200mm, and specimens (100x100x500mm) of reinforced concrete with steel fibres (35 kg / m3). The results obtained are analysed and are calculated residual flexural tensile strengths which characterise the post-cracking behaviour of steel fibres reinforced concrete. In the third part are presented numerical simulations of the tests of pipes and specimens. The model adopted for the pipes test was a three-dimensional model and loads considered were those obtained in experimental tests at reaching breaking forces. Tensile stresses determined were compared with mean flexural tensile strength. To validate tensile parameters of steel fibres reinforced concrete, experimental tests of the specimens were modelled with MIDAS program to reproduce the flexural breaking behaviour. To simulate post - cracking behaviour was used the method σ — ε based on the relationship stress - strain, according to RILEM TC 162-TDF. For the specimens tested were plotted F — δ diagrams, which have been superimposed for comparison with the similar diagrams of experimental tests. The comparison of experimental results with those obtained from numerical simulation leads to the following conclusions: - the maximum forces obtained by numerical calculation have higher values than the experimental values for the same tensile stresses; - forces corresponding of residual strengths have very similar values between the experimental and numerical calculations; - generally the numerical model estimates a breaking force greater

  10. Development and testing of restraints for nuclear piping systems

    International Nuclear Information System (INIS)

    Kelly, J.M.; Skinner, M.S.

    1980-06-01

    As an alternative to current practice of pipe restraint within nuclear power plants it has been proposed to adopt restraints capable of dissipating energy in the piping system. The specific mode of energy dissipation focused upon in these studies is the plastic yielding of steels utilizing relative movement between the pipe and the base of the restraint, a general mechanism which has been proven as reliable in several allied studies. This report discusses the testing of examples of two energy-absorbing devices, the results of this testing and the conclusions drawn. This study concentrated on the specific relevant performance characteristics of hysteretic behavior and degradation with use. The testing consisted of repetitive continuous loadings well into the plastic ranges of the devices in a sinusoidal or random displacement controlled mode

  11. Comparative study of computational model for pipe whip analysis

    International Nuclear Information System (INIS)

    Koh, Sugoong; Lee, Young-Shin

    1993-01-01

    Many types of pipe whip restraints are installed to protect the structural components from the anticipated pipe whip phenomena of high energy lines in nuclear power plants. It is necessary to investigate these phenomena accurately in order to evaluate the acceptability of the pipe whip restraint design. Various research programs have been conducted in many countries to develop analytical methods and to verify the validity of the methods. In this study, various calculational models in ANSYS code and in ADLPIPE code, the general purpose finite element computer programs, were used to simulate the postulated pipe whips to obtain impact loads and the calculated results were compared with the specific experimental results from the sample pipe whip test for the U-shaped pipe whip restraints. Some calculational models, having the spring element between the pipe whip restraint and the pipe line, give reasonably good transient responses of the restraint forces compared with the experimental results, and could be useful in evaluating the acceptability of the pipe whip restraint design. (author)

  12. Seismic fragility test of a 6-inch diameter pipe system

    International Nuclear Information System (INIS)

    Chen, W.P.; Onesto, A.T.; DeVita, V.

    1987-02-01

    This report contains the test results and assessments of seismic fragility tests performed on a 6-inch diameter piping system. The test was funded by the US Nuclear Regulatory Commission (NRC) and conducted by ETEC. The objective of the test was to investigate the ability of a representative nuclear piping system to withstand high level dynamic seismic and other loadings. Levels of loadings achieved during seismic testing were 20 to 30 times larger than normal elastic design evaluations to ASME Level D limits would permit. Based on failure data obtained during seismic and other dynamic testing, it was concluded that nuclear piping systems are inherently able to withstand much larger dynamic seismic loadings than permitted by current design practice criteria or predicted by the probabilistic risk assessment (PRA) methods and several proposed nonlinear methods of failure analysis

  13. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  14. High temperature superconducting current lead test facility with heat pipe intercepts

    International Nuclear Information System (INIS)

    Blumenfeld, P.E.; Prenger, C.; Roth, E.W.; Stewart, J.A.

    1998-01-01

    A high temperature superconducting (HTS) current lead test facility using heat pipe thermal intercepts is under development at the Superconducting Technology Center at Los Alamos National Laboratory. The facility can be configured for tests at currents up to 1,000 A. Mechanical cryocoolers provide refrigeration to the leads. Electrical isolation is maintained by intercepting thermal energy from the leads through cryogenic heat pipes. HST lead warm end temperature is variable from 65 K to over 90 K by controlling heat pipe evaporator temperature. Cold end temperature is variable up to 30 K. Performance predictions in terms of heat pipe evaporator temperature as a function of lead current are presented for the initial facility configuration, which supports testing up to 200 A. Measurements are to include temperature and voltage gradient in the conventional and HTS lead sections, temperature and heat transfer rate in the heat pipes. as well as optimum and off-optimum performance of the conventional lead sections

  15. Structural integrity assessment of piping components

    International Nuclear Information System (INIS)

    Kushwaha, H.S.; Chattopadhyay, J.

    2008-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

  16. Seismic response and damping tests of small bore LMFBR piping and supports

    International Nuclear Information System (INIS)

    Barta, D.A.; Anderson, M.J.; Severud, L.K.; Lindquist, M.R.

    1984-01-01

    Seismic testing and analysis of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described. Measured responses to simulated seismic excitations are compared with analytical predictions based on NRC Regulatory Guide 1.61 and measured system damping values. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers and pipe clamps

  17. Leak-thight seals got high pressure testing of pipes, tanks, valves

    International Nuclear Information System (INIS)

    Estrade, J.

    1985-01-01

    Leak-tight seals ensure quick, safe and efficient testing of pipes with plain-ended or flanged openings, valves with flanged or welded edges, manifields, recipients, etc. They are inserted into the pipe end manually then simply a slight turn of the seal treated wheel commences the pressure test. Hydraulic pressure is supplied by a pump through the inlet seal and air is purged through the outlet seal which then closes. The higher the pressure, the greater the sealing strength of the seal which prevents accidental unplugging. There are different types of seals: for interior plain-ended openings, for pipes with plain-ended opening, for flanged pipes. (author)

  18. hree-Dimensional Finite Element Simulation of the Buried Pipe Problem in Geogrid Reinforced Soil

    Directory of Open Access Journals (Sweden)

    Mohammed Yousif Fattah

    2016-05-01

    Full Text Available Buried pipeline systems are commonly used to transport water, sewage, natural oil/gas and other materials. The beneficial of using geogrid reinforcement is to increase the bearing capacity of the soil and decrease the load transfer to the underground structures. This paper deals with simulation of the buried pipe problem numerically by finite elements method using the newest version of PLAXIS-3D software. Rajkumar and Ilamaruthi's study, 2008 has been selected to be reanalyzed as 3D problem because it is containing all the properties needed by the program such as the modulus of elasticity, Poisson's ratio, angle of internal friction. It was found that the results of vertical crown deflection for the model without geogrid obtained from PLAXIS-3D are higher than those obtained by two-dimensional plane strain by about 21.4% while this percent becomes 12.1 for the model with geogrid, but in general, both have the same trend. The two dimensional finite elements analysis predictions of pipe-soil system behavior indicate an almost linear displacement of pipe deflection with applied pressure while 3-D analysis exhibited non-linear behavior especially at higher loads.

  19. Pipe-to-pipe impact program

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Bampton, M.C.C.; Friley, J.R.; Simonen, F.A.

    1984-06-01

    This report documents the tests and analyses performed as part of the Pipe-to-Pipe Impact (PTPI) Program at the Pacific Northwest Laboratory. This work was performed to assist the NRC in making licensing decisions regarding pipe-to-pipe impact events following postulated breaks in high energy fluid system piping. The report scope encompasses work conducted from the program's start through the completion of the initial hot oil tests. The test equipment, procedures, and results are described, as are analytic studies of failure potential and data correlation. Because the PTPI Program is only partially completed, the total significance of the current test results cannot yet be accurately assessed. Therefore, although trends in the data are discussed, final conclusions and recommendations will be possible only after the completion of the program, which is scheduled to end in FY 1984

  20. Pipe clamp effects on thin-walled pipe design

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1980-01-01

    Clamp induced stresses in FFTF piping are sufficiently large to require structural assessment. The basic principles and procedures used in analyzing FFTF piping at clamp support locations for compliance with ASME Code rules are given. Typical results from a three-dimensional shell finite element pipe model with clamp loads applied over the clamp/pipe contact area are shown. Analyses performed to categorize clamp induced piping loads as primary or secondary in nature are described. The ELCLAMP Computer Code, which performs analyses at clamp locations combining clamp induced stresses with stresses from overall piping system loads, is discussed. Grouping and enveloping methods to reduce the number of individual clamp locations requiring analysis are described

  1. Finite element limit loads for non-idealized through-wall cracks in thick-walled pipe

    International Nuclear Information System (INIS)

    Shim, Do-Jun; Han, Tae-Song; Huh, Nam-Su

    2013-01-01

    Highlights: • The lower bound bulging factor of thin-walled pipe can be used for thick-walled pipe. • The limit loads are proposed for thick-walled, transition through-wall cracked pipe. • The correction factors are proposed for estimating limit loads of transition cracks. • The limit loads of short transition cracks are similar to those of idealized cracks. - Abstract: The present paper provides plastic limit loads for non-idealized through-wall cracks in thick-walled pipe. These solutions are based on detailed 3-dimensional finite element (FE) analyses which can be used for structural integrity assessment of nuclear piping. To cover a practical range of interest, the geometric variables and loading conditions affecting the plastic limit loads of thick-walled pipe with non-idealized through-wall cracks were systematically varied. In terms of crack orientation, both circumferential and axial through-wall cracks were considered. As for loading conditions, axial tension, global bending, and internal pressure were considered for circumferential cracks, whereas only internal pressure was considered for axial cracks. Furthermore, the values of geometric factor representing shape characteristics of non-idealized through-wall cracks were also systematically varied. In order to provide confidence in the present FE analyses results, plastic limit loads of un-cracked, thick-walled pipe resulting from the present FE analyses were compared with the theoretical solutions. Finally, correction factors to the idealized through-wall crack solutions were developed to determine the plastic limit loads of non-idealized through-wall cracks in thick-walled pipe

  2. Seismic damping factors of small-bore piping as influenced by insulation and support elements

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Barta, D.A.

    1983-01-01

    Seismic damping tests of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small-bore piping system is described, and measured transient responses to pulse excitations are reported. The test specimen was representative of a typical LMFBR insulated small-bore piping system, and it was supported from a rigid test fame by prototypic dead-weight supports, mechanical snubbers, and pipe clamps. Various support configurations were tested to assess the reponse sensitivity to insulation and use of mechanical snubbers. Factors much higher than the magnitudes currently allowed in design, by the USNRC Regulatory Guide 1.61, were found. This verified the design values but also pointed out the possibility of undue conservatism and costly overdesign resulting from use of the present design values

  3. Pipe damping: experimental results from laboratory tests in the seismic frequency range

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1986-06-01

    The Idaho National Engineering Laboratory (INEL) has been conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for the seismic analysis of nuclear piping systems. As part of this program, a 5-in. piping system was tested by the INEL, and data from USNRC/EPRI piping vibration tests at the ANCO Engineers facility were evaluated. These systems were subjected to various types of excitation methods and magnitudes, the support configurations were varied, and the effects of pipe insulation and internal pressure were investigated on the INEL system. The INEL has used several different methods to reduce the data to determine the damping in both these piping systems under the various test conditions. It was concluded that at representative seismic excitation levels, pressure was not a contributing factor, but the supports, insulation, and magnitude of response all were major influences contributing to damping. These tests are part of the ongoing program to determine how various parameters and data reduction methods affect piping system damping. The evaluation of all relevant test results, including these two series, will potentially lead to revised damping guidelines for the seismic analysis of nuclear plants, making them safer, less costly, and easier to inspect and maintain. The test results as well as accompanying evaluations and recommendations are presented in this report. 27 refs., 72 figs., 13 tabs

  4. A finite element model for the stress and flexibility analysis of curved pipes

    International Nuclear Information System (INIS)

    Guerreiro, J.N.C.

    1987-03-01

    We present a finite element model for the analysis of pipe bends with flanged ends or flanged tangents. Comments are made on the consideration of the internal pressure load. Flexibility and stress instensification factores obtained with the present model are compared with others available. (Author) [pt

  5. Critical element development of standard components for pipe welding/cutting by CO{sub 2} laser

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1994-11-01

    In D-T burning reactors such as International Thermonuclear Experimental Reactor(ITER), an internal access is inevitable for welding/cutting of cooling pipes of in-vessel components, because of spatial constraint due to a narrow port opening space. An internal-access pipe welding/cutting equipment is being developed in JAERI. Internal access is to approach through inside a pipe to a welding/cutting position, to use 10kW CO{sub 2} laser beam, and to be applicable to both welding and cutting with using a same processing head. A welding/cutting processing head with 10kW CO{sub 2} laser beam has been fabricated and the basic feasibility has been successfully demonstrated for studies of the internal-access pipe welding/cutting concept using 100-A stainless steel pipe with a thickness of 6.3mm. In this study, the optimum focal point of laser beam, laser power and traveling speed of the head have been investigated together with an adjusting mechanism of a relative distance between the head and the pipe wall. In addition, the radiation resistance of critical elements such as optical lens has been investigated. (author).

  6. Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

    International Nuclear Information System (INIS)

    French, R.T.

    1975-08-01

    Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

  7. Proportioning equipment for vibration filling and compacting of grain materials in pipe containers, especially of fuel elements

    International Nuclear Information System (INIS)

    Pinkas, V.; Filip, Z.; Beranek, J.

    1981-01-01

    The equipment consists of a base plate to which are attached the fastening collar fo the pipe container and the guide column with the height-adjustable support. The filling pipe is fixed to the support. The proportioning equipment prevents particles of grain material from segregation, thus allowing to achieve homogeneity of the material in the whole volume to be compacted. It also allows determining the height of the column of material in the pipe container without destructive effects on the stacked material. The equipment is designed for the manufacture of shortened fuel elements. (J.B.)

  8. Seismic damping factors of small bore piping as influenced by insulation and support elements

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Barta, D.A.

    1985-01-01

    Seismic damping tests of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described, and measured transient responses to pulse excitations are reported. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers, and pipe clamps. Various support configurations were tested to assess the response sensitivity to insulation and other nonlinear support characteristics. Damping factors increased significantly due to the insulation and use of mechanical snubbers. Factors much higher than the magnitudes currently allowed in design by the USNRC Regulatory Guide 1.61, were found. This verified the design values, and it also pointed out the possibility of undue conservatism and costly overdesign resulting from use of the present design values

  9. Seismic damping factors of small bore piping as influenced by insulation and support elements

    International Nuclear Information System (INIS)

    Severud, L.K.; Barta, D.A.

    1983-01-01

    Seismic damping tests of a prototypical Liquid Metal Fast Breeder Reactor (LMFBR) small bore piping system is described, and measured transient responses to pulse excitations are reported. The test specimen was representative of a typical LMFBR insulated small bore piping system, and it was supported from a rigid test frame by prototypic dead weight supports, mechanical snubbers, and pipe clamps. Various support configurations were tested to assess the response sensitivity to insulation and other nonlinear support characteristics. Damping factors increased significantly due to the insulation and use of mechanical snubbers. Factors much higher than the magnitudes currently allowed in design, by the USNRC Regulatory Guide 1.61, were found. This verified the design values but also pointed out the possibility of undue conservatism and costly overdesign resulting from use of the present design values

  10. Welding of sule elements for nuclear reactors with solid state YAG laser using instrumentated testing equipments

    International Nuclear Information System (INIS)

    Bourgault, F.; Lacoste, J.; Schley, R.; Kluzinski, C.; Piednoir, P.

    1985-09-01

    The instrumentation of the equipment for carrying out safety tests on fuel elements for nuclear reactors requires special thermocouples adapted to the prevailing agressive medium. The investigations described deal essentially with the operational and metallurgical weldability tests out on the safety test zircaloy piping in the pressurized water circuit (PHEBUS-programme) [fr

  11. Sodium heat pipe module test for the SAFE-30 reactor prototype

    International Nuclear Information System (INIS)

    Reid, Robert S.; Sena, J. Tom; Martinez, Adam L.

    2001-01-01

    Reliable, long-life, low-cost heat pipes can enable safe, affordable space fission power and propulsion systems. Advanced versions of these systems can in turn allow rapid access to any point in the solar system. Twelve stainless steel-sodium heat pipe modules were built and tested at Los Alamos for use in a non-nuclear thermohydraulic simulation of the SAFE-30 reactor (Poston et al., 2000). SAFE-30 is a near-term, low-cost space fission system demonstration. The heat pipes were designed to remove thermal power from the SAFE-30 core, and transfer this power to an electrical power conversion system. These heat pipe modules were delivered to NASA Marshall Space Flight Center in August 2000 and were assembled and tested in a prototypical configuration during September and October 2000. The construction and test of one of the SAFE-30 modules is described

  12. Technical report on the Piping Reliability Proving Tests at the Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1993-05-01

    Japan Atomic Energy Research Institute (JAERI) conducts Piping Reliability Proving Tests from 1975 to 1992 based upon the contracts between JAERI and Science and Technology Agency of Japan (STA) under the auspices of the special account law for electric power development promotion. The purpose of these tests are to prove the structural reliability of the primary cooling piping constituting a part of the pressure boundary in the light water reactor power plants. The tests with large experimental facilities had ended already in 1990. Presently piping reliability analysis by the probabilistic fracture mechanics method is being done. Until now annual reports concerning the proving tests were produced and submitted to STA, whereas this report summarizes the test results done during these 16 years. Objectives of the piping reliability proving tests are to prove that the primary piping of the light water reactor (1) be reliable throughout the service period, (2) have no possibility of rupture, (3) bring no detrimental influence on the surrounding instrumentations or equipments near the break location even if it ruptured suddenly. To attain these objectives (i) pipe fatigue tests, (ii) unstable pipe fracture tests, (iii) pipe rupture tests and also the analyses by computer codes were done. After carrying out these tests, it is verified that the piping is reliable throughout the service period. The authors of this report are T. Isozaki, K. Shibata, S. Ueda, R. Kurihara, K. Onizawa and A. Kohsaka. The parts they wrote are shown in contents. (author)

  13. Structural damping results from vibration tests of straight piping sections

    International Nuclear Information System (INIS)

    Ware, A.G.; Thinnes, G.L.

    1984-01-01

    EG and G Idaho is assisting the USNRC and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on 76-mm and 203-mm (3-in. amd 8-in.) Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 10-m (33-ft) straight sections of piping were rigidly supported at the ends. Spring, rod, and constant force hangers, as well as a sway brace and snubbers were included as intermediate supports. Excitation was provided by low-force level hammer inpacts, a hydraulic shaker, and a 445-kN (50-ton) overhead crane. Data was recorded using acceleration, strain, and displacement time histories. This paper presents results from the testing showing the effect of stress level and type of supports on structural damping in piping

  14. J-integral estimation analysis for circumferential throughwall cracked pipes

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.

    J-integral estimation solution is derived for pipes containing a circumferential throughwall crack. Bending moment and axial tension loadings are considered. These solutions are useful for calculating J from single load-displacement record obtained as part of pipe fracture testing, and are applicable for a wide range of flaw length to pipe circumference ratios. Results for J at initiation of crack growth generated using the solution developed in this paper agree well with J results from finite elements analyses.

  15. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead.

    Science.gov (United States)

    Wu, Jianbo; Fang, Hui; Li, Long; Wang, Jie; Huang, Xiaoming; Kang, Yihua; Sun, Yanhua; Tang, Chaoqing

    2017-01-21

    To meet the great needs for MFL (magnetic flux leakage) inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatability MFL probing system is designed and manufactured, which was embedded with the developed sensors. It can track the swing movement of drill pipes and allow the pipe ends to pass smoothly. Finally, the developed system is employed in a drilling field for drill pipe inspection. Test results show that the proposed method can fulfill the requirements for drill pipe inspection at wellheads, which is of great importance in drill pipe safety.

  16. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead

    Directory of Open Access Journals (Sweden)

    Jianbo Wu

    2017-01-01

    Full Text Available To meet the great needs for MFL (magnetic flux leakage inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatability MFL probing system is designed and manufactured, which was embedded with the developed sensors. It can track the swing movement of drill pipes and allow the pipe ends to pass smoothly. Finally, the developed system is employed in a drilling field for drill pipe inspection. Test results show that the proposed method can fulfill the requirements for drill pipe inspection at wellheads, which is of great importance in drill pipe safety.

  17. Elasto-plastic finite element analysis of axial surface crack in PHT piping of 500 MWe PHWR

    International Nuclear Information System (INIS)

    Chawla, D.S.; Bhate, S.R.; Kushwaha, H.S.; Mahajan, S.C.

    1994-01-01

    The leak before break (LBB) approach in nuclear piping design envisages demonstrating that the pressurized pipe with a postulated flaw will leak at a detectable rate leading to corrective action well before catastrophic rupture would occur. This requires analysis of cracked pipe to study the crack growth and its stability. This report presents the behaviour of a surface crack in the wall of a thick primary heat transport (PHT) pipe of 500 MWe Indian PHWR. The line spring model (LSM) finite element is used to model the flawed pipe geometry. The variation of crack driving force (J-integral) across the crack front has been presented. The influence of crack geometry factors such as depth, shape, aspect ratio, and loading on peak values of J-integral as well as crack mouth opening displacement has been studied. Several crack shapes have been used to study the shape influence. The results are presented in dimensionless form so as to widen their applicability. The accuracy of the results is validated by comparison with results available in open literature. (author). 47 refs., 8 figs

  18. Plastic Limit Loads for Slanted Circumferential Through-Wall Cracked Pipes Using 3D Finite-Element Limit Analyses

    International Nuclear Information System (INIS)

    Jang, Hyun Min; Cho, Doo Ho; Kim, Young Jin; Huh, Nam Su; Shim, Do Jun; Choi, Young Hwan; Park, Jung Soon

    2011-01-01

    On the basis of detailed 3D finite-element (FE) limit analyses, the plastic limit load solutions for pipes with slanted circumferential through-wall cracks (TWCs) subjected to axial tension, global bending, and internal pressure are reported. The FE model and analysis procedure employed in the present numerical study were validated by comparing the present FE results with existing solutions for plastic limit loads of pipes with idealized TWCs. For the quantification of the effect of slanted crack on plastic limit load, slant correction factors for calculating the plastic limit loads of pipes with slanted TWCs from pipes with idealized TWCs are newly proposed from extensive 3D FE calculations. These slant-correction factors are presented in tabulated form for practical ranges of geometry and for each set of loading conditions

  19. J-integral estimation analysis for circumferential throughwall cracked pipes

    International Nuclear Information System (INIS)

    Zahoor, A.

    1988-01-01

    J-integral estimation solution is derived for pipes containing a circumferential throughwall crack. Bending moment and axial tension loadings are considered. These solutions are useful for calculating J from single load-displacement record obtained as part of pipe fracture testing, and are applicable for a wide range of flaw length to pipe circumference ratios. Results for J at initiation of crack growth generated using the solution developed in this paper agree well with J results from finite elements analyses. (orig.)

  20. Test results of a jet impingement from a 4 inch pipe under BWR LOCA conditions

    International Nuclear Information System (INIS)

    Isozaki, Toshikuni; Yano, Toshikazu; Miyazaki, Noriyuki; Kato, Rokuro; Kurihara, Ryoichi; Ueda, Shuzo; Miyazono, Shohachiro

    1982-09-01

    Hypothetical instantaneous pipe rupture is now considered to be one of the design basis accidents during the operation of the light water reactor. If a pipe rupture accidnet occurs, the pipe will start moving with the sudden discharge of internal fluid. So, the various apparatus such as pipe whip restraints and jet deflectors are being installed near the postulated break location to protect the nuclear power plants against the effect of postulated pipe rupture. Pipe whipping test and jet discharge test are now being conducted at the Division of Reactor Safety of the Japan Atomic Energy Research Institute. This report describes the test results of the jet discharge from a 4 inch pipe under BWR LOCA condition. In front of the pipe exit the target disk of 1000 mm in diameter was installed. The distance between the pipe exit and the target was 500 mm. 13 pressure transducers and 13 thermocouples were mounted on the target disk to measure the pressure and temperature increase due to jet impingement on the target. (author)

  1. Radiant heating tests of several liquid metal heat-pipe sandwich panels

    International Nuclear Information System (INIS)

    Camarda, C.J.; Basiulis, A.

    1983-08-01

    Integral heat pipe sandwich panels, which synergistically combine the thermal efficiency of heat pipes and the structural efficiency of honeycomb sandwich construction, were conceived as a means of alleviating thermal stress problems in the Langley Scramjet Engine. Test panels which utilized two different wickable honeycomb cores, facesheets with screen mesh sintered to the internal surfaces, and a liquid metal working fluid (either sodium or potassium) were tested by radiant heating at various heat load levels. The heat pipe panels reduced maximum temperature differences by 31 percent with sodium working fluid and 45 percent with potassium working fluid. Results indicate that a heat pipe sandwich panel is a potential, simple solution to the engine thermal stress problem. Other interesting applications of the concept include: cold plates for electronic component and circuit card cooling, radiators for large space platforms, low distortion large area structures (e.g., space antennas) and laser mirrors

  2. Selenide isotope generator for the Galileo Mission. Axially-grooved heat pipe: accelerated life test results

    International Nuclear Information System (INIS)

    1979-08-01

    The results through SIG/Galileo contract close-out of accelerated life testing performed from June 1978 to June 1979 on axially-grooved, copper/water heat pipes are presented. The primary objective of the test was to determine the expected lifetime of axially-grooved copper/water heat pipes. The heat pipe failure rate, due to either a leak or a build-up of non-condensible gas, was determined. The secondary objective of the test was to determine the effects of time and temperature on the thermal performance parameters relevant to long-term (> 50,000 h) operation on a space power generator. The results showed that the gas generation rate appears to be constant with time after an initial sharp rise although there are indications that it drops to approximately zero beyond approx. 2000 h. During the life test, the following pipe-hours were accumulated: 159,000 at 125 0 C, 54,000 at 165 0 C, 48,000 at 185 0 C, and 8500 at 225 0 C. Heated hours per pipe ranged from 1000 to 7500 with an average of 4720. Applying calculated acceleration factors yields the equivalent of 930,000 pipe-h at 125 0 C. Including the accelerated hours on vendor tested pipes raises this number to 1,430,000 pipe-hours at 125 0 C. It was concluded that, for a heat pipe temperature of 125 0 C and a mission time of 50,000 h, the demonstrated heat pipe reliability is between 80% (based on 159,000 actual pipe-h at 125 0 C) and 98% (based on 1,430,000 accelerated pipe-h at 125 0 C). Measurements indicate some degradation of heat transfer with time, but no detectable degradation of heat transport

  3. Inspection of piping wall loss with flow accelerated corrosion accelerated simulation test

    International Nuclear Information System (INIS)

    Ryu, Kyung Ha; Kim, Ji Hak; Hwang, Il Soon; Lee, Na Young; Kim, Ji Hyun

    2009-01-01

    Flow Accelerated Corrosion (FAC) has become a hot issue for aging of passive components. Ultrasonic Technique (UT) has been adopted to inspect the secondary piping of Nuclear Power Plants (NPPs). UT, however, uses point detection method, which results in numerous detecting points and thus takes time. We developed an Equipotential Switching Direct Current Potential Drop (ES-DCPD) method to monitor the thickness of piping that covers wide range of piping at once time. Since the ES-DCPD method covers area, not a point, it needs less monitoring time. This can be a good approach to broad carbon steel piping system such as secondary piping of NPPs. In this paper, FAC accelerated simulation test results is described. We realized accelerated FAC phenomenon by 2 times test: 23.7% thinning in 216.7 hours and 51% thinning in 795 hours. These were monitored by ES-DCPD and traditional UT. Some parameters of water chemistry are monitored and controlled to accelerate FAC process. As sensitive factors on FAC, temperature and pH was changed during the test. The wall loss monitored results reflected these changes of water chemistry successfully. Developed electrodes are also applied to simulation loop to monitor water chemistry. (author)

  4. Flow induced vibrations of piping

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    In order to design the supports of piping systems, estimations of the vibrations induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary to calculate the model parameters of liquid containing pipes. In most computer codes, fluid effects are accounted for just by adding the fluid mass to the structure. This may lead to serious errors. This paper presents a method to take into account these effects, by solving a coupled mechanical-acoustical problem: the computer code TEDEL of the C.E.A /D.E.M.T. System, based on the finite-elements method, has been extended to calculate simultaneously the pressure fluctuations in the fluid and the vibrations of the pipe. By this way the mechanical-acoustical coupled eigenmodes of any piping system can be obtained. These eigenmodes are used to determine the response of the system to various sources. Equations have been written in the hypohesis that acoustical wave lengths remain large compared to the diameter of the pipe. The method has been checked by an experiment performed on the GASCOGNE loop at D.E.M.T. The piping system under test consists of a tube with four elbows. The circuit is ended at each extremity by a large vessel which performs acoustical isolation by generating modes for the pressure. Excitation of the circuit is caused by a valve located near the downstream vessel. This provides an efficient localised broad band acoustical source. The comparison between the test results and the calculations has shown that the low frequency resonant characteristics of the pipe and the vibrational amplitude at various flow-rates can be correctly predicted

  5. Corrosion evaluation in insulated pipes by non destructive testing method

    International Nuclear Information System (INIS)

    Abd Razak Hamzah; Azali Muhammad; Mohammad Pauzi Ismail; Abd Nassir Ibrahim; Abd Aziz Mohamed; Sufian Saad; Saharuddin Sayuti; Shukri Ahmad

    2002-01-01

    In engineering plants, detection of corrosion and evaluation of deposit in insulated pipes using radiography method are considered as a very challenging tasks. In General this degradation problem is attributed to water condensation. It causes the formation of deposit and scale inside the pipe, as well as between the insulation and pipe in cold temperature pipes. On the other hand, for hot temperature pipes the main problem is mainly due to corrosion/erosion attack inside the pipe. In the study of corrosion in pipelines, one of the most important parameters to be monitored and measured is the wall thickness. Currently, most pipeline corrosion monitoring and evaluation for both insulated and non-insulated pipes is performed using an ultrasonic method. The most common technique is that based on the A-Scan, using either a normal flaw detector or some form of dedicated equipment. However, with recent development of ultrasonic technology, more advance method, namely B-Scan and C-scan techniques are also available. The most notable disadvantage of using this method is that the insulation covering the pipe has to be removed before the inspection can be carried out and this is considered as not so cost effective. Due to this reason, the possibility of employing other alternative NDT method, namely radiographic testing method were studied. The technique used in this studied are known as tangential technique. In this study it was found that the result found using tangential technique is consistent with the actual thickness of the pipe. Result of this study is presented and discussed in this paper. (Author)

  6. Device for the automatic X-ray testing of welded joints of pipes

    International Nuclear Information System (INIS)

    Ries, K.; Hannoschieck, K.; Rozic, K.M.; Basler, G.

    1979-01-01

    The notification flows of the tested pipes determined by the ultrasonic inspection are transmitted to the X-ray film automatic charger in the X-ray test room. The roll table for the pipes from the ultrasonic inspection to the X-ray test room is provided with an arrangement for weld detection and tube lathe, so that the X-ray films can be set on the corresponding spot by means of a cantilever. (RW) [de

  7. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  8. Impact analyses after pipe rupture

    International Nuclear Information System (INIS)

    Chun, R.C.; Chuang, T.Y.

    1983-01-01

    Two of the French pipe whip experiments are reproduced with the computer code WIPS. The WIPS results are in good agreement with the experimental data and the French computer code TEDEL. This justifies the use of its pipe element in conjunction with its U-bar element in a simplified method of impact analyses

  9. Development of seismic design method for piping system supported by elastoplastic damper. 3. Vibration test of three-dimensional piping model and its response analysis

    International Nuclear Information System (INIS)

    Namita, Yoshio; Kawahata, Jun-ichi; Ichihashi, Ichiro; Fukuda, Toshihiko.

    1995-01-01

    Component and piping systems in current nuclear power plants and chemical plants are designed to employ many supports to maintain safety and reliability against earthquakes. However, these supports are rigid and have a slight energy-dissipating effect. It is well known that applying high-damping supports to the piping system is very effective for reducing the seismic response. In this study, we investigated the design method of the elastoplastic damper [energy absorber (EAB)] and the seismic design method for a piping system supported by the EAB. Our final goal is to develop technology for applying the EAB to the piping system of an actual plant. In this paper, the vibration test results of the three-dimensional piping model are presented. From the test results, it is confirmed that EAB has a large energy-dissipating effect and is effective in reducing the seismic response of the piping system, and that the seismic design method for the piping system, which is the response spectrum mode superposition method using each modal damping and requires iterative calculation of EAB displacement, is applicable for the three-dimensional piping model. (author)

  10. ASME code and ratcheting in piping components. Final technical report

    International Nuclear Information System (INIS)

    Hassan, T.; Matzen, V.C.

    1999-01-01

    The main objective of this research is to develop an analysis program which can accurately simulate ratcheting in piping components subjected to seismic or other cyclic loads. Ratcheting is defined as the accumulation of deformation in structures and materials with cycles. This phenomenon has been demonstrated to cause failure to piping components (known as ratcheting-fatigue failure) and is yet to be understood clearly. The design and analysis methods in the ASME Boiler and Pressure Vessel Code for ratcheting of piping components are not well accepted by the practicing engineering community. This research project attempts to understand the ratcheting-fatigue failure mechanisms and improve analysis methods for ratcheting predictions. In the first step a state-of-the-art testing facility is developed for quasi-static cyclic and seismic testing of straight and elbow piping components. A systematic testing program to study ratcheting is developed. Some tests have already been performed and the rest will be completed by summer'99. Significant progress has been made in the area of constitutive modeling. A number of sophisticated constitutive models have been evaluated in terms of their simulations for a broad class of ratcheting responses. From the knowledge gained from this evaluation study two improved models are developed. These models are demonstrated to have promise in simulating ratcheting responses in piping components. Hence, implementation of these improved models in widely used finite element programs, ANSYS and/or ABAQUS, is in progress. Upon achieving improved finite element programs for simulation of ratcheting, the ASME Code provisions for ratcheting of piping components will be reviewed and more rational methods will be suggested. Also, simplified analysis methods will be developed for operability studies of piping components and systems. Some of the future works will be performed under the auspices of the Center for Nuclear Power Plant Structures

  11. Study on structural integrity of thinned wall piping against seismic loading-overview and future program

    International Nuclear Information System (INIS)

    Nakamura, Izumi; Otani, Akihito; Shiratori, Masaki

    2005-01-01

    In order to clarify the behavior of thinned wall pipes under seismic events, cyclic in-plane and/or out-of-plane bending tests on thinned straight pipe and elbow and also shaking table tests using degraded piping system models were conducted. Relation between the failure mode and thinned condition and the influence of the final failure mode of degraded piping systems were investigated. In addition to these experiments, elastic-plastic FEM analysis using ABAQUS were conducted on thinned piping elements. It has been found that the strain concentrated point could be predicted and the cause of its generation could be explained by the simulated deformation behavior of the pipe. In order to predict the piping system's maximum response under elastic-plastic response, a simple response prediction method was proposed. Further tests and safety margin analyses of thinned pipes against seismic loading will be performed. (T. Tanaka)

  12. Piping equipment; Materiel petrole

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This 'blue bible' of the perfect piping-man appeals to end-users of industrial facilities of the petroleum and chemical industries (purchase services, standardization, new works, maintenance) but also to pipe-makers and hollow-ware makers. It describes the characteristics of materials (carbon steels, stainless steels, alloyed steels, special alloys) and the dimensions of pipe elements: pipes, welding fittings, flanges, sealing products, forged steel fittings, forged steel valves, cast steel valves, ASTM standards, industrial valves. (J.S.)

  13. In-Pipe Wireless Communication for Underground Sampling and Testing

    NARCIS (Netherlands)

    Nguyen, Nhan D.T.; Le, Duc V.; Meratnia, Nirvana; Havinga, Paul J.M.

    2017-01-01

    In this paper, we present an effective and low- cost wireless communication system for extremely long and narrow pipes that can replay the extant wire system in underground sensor network applications such as soil sampling and testing with the Cone Penetration Test (CPT), the most widely used

  14. Evaluation of clamp effects on LMFBR piping systems

    International Nuclear Information System (INIS)

    Jones, G.L.

    1980-01-01

    Loop-type liquid metal breeder reactor plants utilize thin-wall piping to mitigate through-wall thermal gradients due to rapid thermal transients. These piping loops require a support system to carry the combined weight of the pipe, coolant and insulation and to provide attachments for seismic restraints. The support system examined here utilizes an insulated pipe clamp designed to minimize the stresses induced in the piping. To determine the effect of these clamps on the pipe wall a non-linear, two-dimensional, finite element model of the clamp, insulation and pipe wall was used to determine the clamp/pipe interface load distributions which were then applied to a three-dimensional, finite element model of the pipe. The two-dimensional interaction model was also utilized to estimate the combined clamp/pipe stiffness

  15. Development and test of a space-reactor-core heat pipe

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Runyan, J.E.; Martinez, H.E.; Keddy, E.S.

    1983-01-01

    A heat pipe designed to meet the heat transfer requirements of a 100-kW/sub e/ space nuclear power system has been developed and tested. General design requirements for the device included an operating temperature of 1500 0 K with an evaporator radial flux density of 100 w/cm 2 . The total heat-pipe length of 2 m comprised an evaporator length of 0.3 m, a 1.2-m adiabatic section, and a condenser length of 0.5 m. A four-artery design employing screen arteries and distribution wicks was used with lithium serving as the working fluid. Molybdenum alloys were used for the screen materials and tube shell. Hafnium and zirconium gettering materials were used in connection with a pre-purified distilled lithium charge to ensure internal chemical compatibility. After initial performance verification, the 14.1-mm i.d. heat pipe was operated at 15 kW throughput at 1500 0 K for 100 hours. No performance degradation was observed during the test

  16. Alkali Metal Heat Pipe Life Issues

    International Nuclear Information System (INIS)

    Reid, Robert S.

    2004-01-01

    One approach to fission power system design uses alkali metal heat pipes for the core primary heat-transfer system. Heat pipes may also be used as radiator elements or auxiliary thermal control elements. This synopsis characterizes long-life core heat pipes. References are included where information that is more detailed can be found. Specifics shown here are for demonstration purposes and do not necessarily reflect current Nasa Project Prometheus point designs. (author)

  17. Users manual on database of the Piping Reliability Proving Tests at the Japan Atomic Energy Research Institute

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Japan Atomic Energy Research Institute(JAERI) conducted Piping Reliability Proving Tests from 1975 to 1992 based upon the contracts between JAERI and Science and Technology Agency of Japan under the auspices of the special account law for electric power development promotion. The purposes of those tests are to prove the structural reliability of the primary cooling piping constituting a part of the pressure boundary in the water reactor power plants. The tests with large experimental facilities had ended already in 1990. After that piping reliability analysis by the probabilistic method followed until 1992. This report describes the users manual on databases about the test results using the large experimental facilities. Objectives of the piping reliability proving tests are to prove that the primary piping of the water reactor (1) be reliable throughout the service period, (2) have no possibility of rupture, (3) bring no detrimental influence on the surrounding instrumentations or equipments near the break location. The research activities using large scale piping test facilities are described. The present report does the database about the test results pairing the former report. With these two reports, all the feature of Piping Reliability Proving Tests is made clear. Briefings of the tests are described also written in Japanese or English. (author)

  18. Nondestructive testing during the fabrication of pressure vessels with half pipe jackets

    International Nuclear Information System (INIS)

    Scherner, D.

    1985-01-01

    The most important precondition to guarantee the optimum quality of half pipe jackets is the precise fixing and observance of the manufacturing conditions. For this reason the manufacturing conditions are explained in detail. The second important point is the test for gas tightness of the half pipe jacket system. The sources of mistakes in connection with the test for gas tightness are of fundamental importance. (orig./PW) [de

  19. Study on seismic design margin based upon inelastic shaking test of the piping and support system

    International Nuclear Information System (INIS)

    Ishiguro, Takami; Eto, Kazutoshi; Ikeda, Kazutoyo; Yoshii, Toshiaki; Kondo, Masami; Tai, Koichi

    2009-01-01

    In Japan, according to the revised Regulatory Guide for Aseismic Design of Nuclear Power Reactor Facilities, September 2006, criteria of design basis earthquakes of Nuclear Power Reactor Facilities become more severe. Then, evaluating seismic design margin took on a great importance and it has been profoundly discussed. Since seismic safety is one of the major key issues of nuclear power plant safety, it has been demonstrated that nuclear piping system possesses large safety margins by various durability test reports for piping in ultimate conditions. Though the knowledge of safety margin has been accumulated from these reports, there still remain some technical uncertainties about the phenomenon when both piping and support structures show inelastic behavior in extremely high seismic excitation level. In order to obtain the influences of inelastic behavior of the support structures to the whole piping system response when both piping and support structures show inelastic behavior, we examined seismic proving tests and we conducted simulation analyses for the piping system which focused on the inelastic behavior of the support to the whole piping system response. This paper introduces major results of the seismic shaking tests of the piping and support system and the simulation analyses of these tests. (author)

  20. Experimental electro-thermal method for nondestructively testing welds in stainless steel pipes

    International Nuclear Information System (INIS)

    Green, D.R.

    1979-01-01

    Welds in austenitic stainless steel pipes are notoriously difficult to nondestructively examine using conventional ultrasonic and eddy current methods. Survace irregularities and microscopic variations in magnetic permeability cause false eddy current signal variations. Ultrasonic methods have been developed which use computer processing of the data to overcome some of the problems. Electro-thermal nondestructive testing shows promise for detecting flaws that are difficult to detect using other NDT methods. Results of a project completed to develop and demonstrate the potential of an electro-thermal method for nondestructively testing stainless steel pipe welds are presented. Electro-thermal NDT uses a brief pulse of electrical current injected into the pipe. Defects at any depth within the weld cause small differences in surface electrical current distribution. These cause short-lived transient temperature differences on the pipe's surface that are mapped using an infrared scanning camera. Localized microstructural differences and normal surface roughness in the welds have little effect on the surface temperatures

  1. Determination of the concentration profile of chemical elements in superheater pipes

    International Nuclear Information System (INIS)

    Aldape U, F.; Aspiazu F, J.

    1986-05-01

    This work has for object to determine the profile of concentration of chemical elements at trace level in a superheater pipe of Thermoelectric Plants using the X-ray emission spectroscopy technique induced by protons coming from the Accelerator of the Nuclear Center. In the X-ray detection, a Si Li detector was used. The technique was chosen because it allows a multielemental analysis, of high sensitivity and precision. The results can help to understand the problems that are had in the change of flexibility or of corrosion. This will be from utility to the Federal Electricity Commission (CFE). (Author)

  2. Specifying and manufacturing piping for the fast flux test facility

    International Nuclear Information System (INIS)

    Moen, R.A.; O'Keefe, D.P.; Irvin, J.E.; Tobin, J.C.

    1974-01-01

    Specification of materials for liquid metal reactor coolant piping, at service temperatures up to 1200 0 F, involves a number of considerations unique to these systems. The mechanical property/design allowable stress considerations which led to the selection and specification of specific materials for the Fast Flux Test Facility piping are discussed. Additional considerations are described indicating allowances made for material changes anticipated in service. These measures primarily involved raising the minimum carbon content to a value that would insure the strength of the material always remains above that assumed in the initial design, although other considerations are discussed. The processes by which this piping was manufactured, its resulting characteristics and methods of subsequent handling/assembly are briefly discussed. (U.S.)

  3. Development of new assessment methodology for locally corroded pipe

    International Nuclear Information System (INIS)

    Lim, Hwan; Shim, Do Jun; Kim, Yun Jae; Kim, Young Jin

    2002-01-01

    In this paper, a unified methodology based on the local stress concept to estimate residual strength of locally thinned pipes is proposed. An underlying idea of the proposed methodology is that the local stress in the minimum section for locally thinned pipe is related to the reference stress, popularly used in creep problems. Then the problem remains how to define the reference stress, that is the reference load. Extensive three-dimensional Finite Element (FE) analyses were performed to simulate full-scale pipe tests conducted for various shapes of wall thinned area under internal pressure and bending moment. Based on these FE results, the reference load is proposed, which is independent of materials. A natural outcome of this method is the maximum load capacity. By comparing with existing test results, it is shown that the reference stress is related to the fracture stress, which in turn can be posed as the fracture criterion of locally thinned pipes. The proposed method is powerful as it can be easily generalised to more complex problems, such as pipe bends and tee-joints

  4. Computation of the temperatures of a fluid flowing through a pipe from temperature measurements on the pipe's outer surface

    International Nuclear Information System (INIS)

    Sauer, G.

    1999-01-01

    A method for computing the temperatures of a fluid flowing through a pipe on the basis of temperatures recorded at the pipe's outer surface is presented. The heat conduction in the pipe wall is described by one-dimensional heat conduction elements. Heat transfer between fluid, pipe and surrounding is allowed for. The equation system resulting from the standard finite element discretization is reformulated to enable the computation of temperature events preceding the recorded temperature in time. It is shown that the method can be used to identify the actual fluid temperature from temperature data obtained only at the outer surface of the pipe. The temperatures in the pipe wall are computed with good accuracy even in the case of a severe thermal shock. (orig.) [de

  5. Determination of hoop direction effective elastic moduli of non-circular profile, fiber reinforced polymer composite sewer liner pipes from lateral ring compression tests

    International Nuclear Information System (INIS)

    Czél, Gergely; Takács, Dénes

    2015-01-01

    A new material property determination method is presented for the calculation of effective elastic moduli of non-circular ring specimens cut from filament wound oval profile polymer composite sewer liner pipes. The hoop direction elastic moduli was determined using the test results obtained from ring compression tests, which is a very basic setup, and requires no special equipment. Calculations were executed for many different oval profiles, and diagrams were constructed, from which the cross section dependent C_e_f_f constants can be taken. The new method was validated by the comparison of tests and finite element analysis results. The calculation method and the diagrams are essential design tools for engineers, and a big step forward in sizing non-circular profile liner pipes. - Highlights: • A simple modulus measurement method is presented for non-circular ring specimens. • The evaluation method is validated against a finite element model. • Profile shape dependent constants are presented for a wide range of cross-sections. • A set of charts with the constants are provided to aid design engineers.

  6. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes

    International Nuclear Information System (INIS)

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping

  7. Finite element-based limit load of piping branch junctions under combined loadings

    International Nuclear Information System (INIS)

    Xuan Fuzhen; Li Peining

    2004-01-01

    The limit load is an important input parameter in engineering defect-assessment procedures and strength design. In the present work, a total of 100 different piping branch junction models for the limit load calculation were performed under combined internal pressure and moments in use of non-linear finite element (FE) method. Three different existing accumulation rules for limit load, i.e., linear equation, parabolic equation and quadratic equation were discussed on the basis of FE results. A novel limit load solution was developed based on detailed three-dimensional FE limit analyses which accommodated the geometrical parameter influence, together with analytical solutions based on equilibrium stress fields. Finally, six experimental results were provided to justify the presented equation. According to the FE limit analysis, limit load interaction of the piping tees under combined pressure and moments has a relationship with the geometrical parameters, especially with the diameter ratio d/D. The predicted limit loads from the presented formula are very close to the experimental data. The resulting limit load solution is given in a closed form, and thus can be easily used in practice

  8. Evaluation of aluminum drill-pipe material and design

    Energy Technology Data Exchange (ETDEWEB)

    Placido, Joao C. [PETROBRAS, Rio de Janeiro, RJ (Brazil); Lourenco, Marcelo I.; Netto, Theodoro Antoun [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia (COPPE)

    2008-07-01

    Experimental program and numerical analyses were carried out to investigate the fatigue mechanisms of aluminum drill pipes designed and manufactured in compliance with ISO 15546. The main objective is to improve the fatigue performance of these components by selecting the appropriate aluminum alloy and by enhancing the mechanical design of the threaded steel connector. This paper presents the experimental test program and numerical analyses conducted on a drill-pipe of different materials (Al-Cu-Mg and Al-Zn-Mg system aluminum alloys) and geometry. Material mechanical properties, including S-N curve, were determined through small-scale tests on specimens cut from actual drill pipes. Full-scale experiments were also performed in laboratory. A finite element model of the drill pipe, including the tool-joint region, was developed. The model simulates, through different load steps, the tool-joint hot assembly, and then reproduces the physical experiments numerically in order to obtain the actual stress distribution. Good correlation between full-scale and small-scale fatigue tests was obtained by adjusting the strain/stress levels monitored in the full-scale tests in light of the numerical simulations and performing fatigue life calculations via multiaxial fatigue models. The weak points of the current practice design are highlighted for further development. (author)

  9. Reduced Multivariate Polynomial Model for Manufacturing Costs Estimation of Piping Elements

    Directory of Open Access Journals (Sweden)

    Nibaldo Rodriguez

    2013-01-01

    Full Text Available This paper discusses the development and evaluation of an estimation model of manufacturing costs of piping elements through the application of a Reduced Multivariate Polynomial (RMP. The model allows obtaining accurate estimations, even when enough and adequate information is not available. This situation typically occurs in the early stages of the design process of industrial products. The experimental evaluations show that the approach is capable, with a low complexity, of reducing uncertainties and to predict costs with significant precision. Comparisons with a neural network showed also that the RMP performs better considering a set of classical performance measures with the corresponding lower complexity and higher accuracy.

  10. Pipe damping

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1984-01-01

    A program has been developed to assess the available piping damping data, to generate additional data and conduct seperate effects tests, and to establish a plan for reporting and storing future test results into a data bank. This effort is providing some of the basis for developing higher allowable damping values for piping seismic analyses, which will potentially permit removal of a considerable number of piping supports, particularly snubbers. This in turn will lead to more flexible piping systems which will be less susceptible to thermal cracking, will be easier to maintain and inspect, as well as less costly

  11. Failure Analysis Of Industrial Boiler Pipe

    International Nuclear Information System (INIS)

    Natsir, Muhammad; Soedardjo, B.; Arhatari, Dewi; Andryansyah; Haryanto, Mudi; Triyadi, Ari

    2000-01-01

    Failure analysis of industrial boiler pipe has been done. The tested pipe material is carbon steel SA 178 Grade A refer to specification data which taken from Fertilizer Company. Steps in analysis were ; collection of background operation and material specification, visual inspection, dye penetrant test, radiography test, chemical composition test, hardness test, metallography test. From the test and analysis result, it is shown that the pipe failure caused by erosion and welding was shown porosity and incomplete penetration. The main cause of failure pipe is erosion due to cavitation, which decreases the pipe thickness. Break in pipe thickness can be done due to decreasing in pipe thickness. To anticipate this problem, the ppe will be replaced with new pipe

  12. Theoretical and experimental study on dynamic responses of piping systems with combined dampers

    International Nuclear Information System (INIS)

    Gershtein, M.; Fridman, Ya.; Perelmiter, A.

    1996-01-01

    Vibrations of pipelines transporting fluids, gases, and granular materials are excited by the air flow, internal pressure pulsation, or seismic ground motion. The susceptibility of oil and gas pipelines to seismic damage has been demonstrated in earthquakes everywhere around the world. Devices for above-ground pipelines and piping systems vibration suppression with combination of dry friction and viscous energy dissipation are developed by AVIBRA, Shear deformation of viscous-elastic material in these devices occurs prior to interfacial slip. The way to account this phenomenon is to model the damper as an elastic-viscous element in series with an ideal Coulomb dry friction element. The harmonic balance method was applied to obtain an equivalent viscous damping constant for a combined damper. Iteration process was used to predict a dynamic response of a piping system with combined dampers subjected to sinusoidal excitation. Every iteration step was based on ANSYS procedures. Time integration of systems with hysteretic friction models presents computational difficulties. Some examples of dynamic responses of piping systems were analyzed by a time integration procedure for finite-element models. Combined dry friction-viscous dissipation dampers were tested on a piping model under harmonic excitation. It was clarified that combined dampers are very effective to reduce dynamic response. The seismic response of the piping system with combined dampers was calculated using time history finite-element analysis. The excellent effectiveness of AVIBRA combined dampers for aseismic design and retrofitting of pipelines and piping systems was confirmed by the analysis

  13. Pipe damping

    International Nuclear Information System (INIS)

    Ware, A.G.

    1985-01-01

    Studies are being conducted at the Idaho National Engineering Laboratory to determine whether an increase in the damping values used in seismic structural analyses of nuclear piping systems is justified. Increasing the allowable damping would allow fewer piping supports which could lead to safer, more reliable, and less costly piping systems. Test data from availble literature were examined to determine the important parameters contributing to piping system damping, and each was investigated in separate-effects tests. From the combined results a world pipe damping data bank was established and multiple regression analyses performed to assess the relative contributions of the various parameters. The program is being extended to determine damping applicable to higher frequency (33 to 100 Hz) fluid-induced loadings. The goals of the program are to establish a methodology for predicting piping system damping and to recommend revised guidelines for the damping values to be included in analyses

  14. Probabilistic procedure to evaluate integrity of degraded pipes under internal pressure and bending moment

    International Nuclear Information System (INIS)

    Roos, E.; Herter, K.-H.; Julisch, P.; Otremba, F.; Schuler, X.

    2003-01-01

    The determination of critical crack sizes or permissible/allowable loading levels in pipes with degraded pipe sections (circumferential cracks) for the assurance of component integrity is usually based on deterministic approaches. Therefore along with numerical calculational methods (finite element (FE) analyses) limit load calculations, such as e.g. the 'Plastic limit load concept' and the 'Flow stress concept' as well as fracture mechanics approximation methods as e.g. the R-curve method or the 'Ductile fracture handbook' and the R6-Method are currently used for practical application. Numerous experimental tests on both ferritic and austenitic pipes with different pipe dimensions were investigated at MPA Stuttgart. The geometries of the pipes were comparable to actual piping systems in Nuclear Power Plants, both BWR as well as PWR. Through wall cracks and part wall through cracks on the inside surface of the pipes were considered. The results of these tests were used to determine the flow stresses used within the limit load calculations. Therefore the deterministic concepts assessing the integrity of degraded pipes are available A new post-calculation of the above mentioned tests was performed using probabilistic approaches to assure the component integrity of degraded piping systems. As a result the calculated probability of failure was compared to experimental behaviour during the pipe test. Different reliability techniques were used for the verification of the probabilistic approaches. (author)

  15. Furnace testing of electrical and pipe-penetration seals based on foamed silicone elastomer: 60, 90, and 120-minute fire ratings

    International Nuclear Information System (INIS)

    Brown, A.

    1979-03-01

    Fire tests of foamed silicone seals for electrical and pipe penetrations have been performed using a furnace with temperature control as heat source. The tests were performed in principle in accordance with the requirements of NORDTEST 5A (ISO 834). The purpose of the tests was to obtain appropriate fire ratings for different seal thicknesses. The report covers. - Description of material used to prepare the seals and method of application - Description of furnace test assembly and method of performing test - Listing of penetrating elements and of the thermocouple array used to measure temperature - Curves of thermocouple readouts and photographs of seals during and after completion of the test. (author)

  16. The behavior of welded joint in steel pipe members under monotonic and cyclic loading

    International Nuclear Information System (INIS)

    Chang, Kyong-Ho; Jang, Gab-Chul; Shin, Young-Eui; Han, Jung-Guen; Kim, Jong-Min

    2006-01-01

    Most steel pipe members are joined by welding. The residual stress and weld metal in a welded joint have the influence on the behavior of steel pipes. Therefore, to accurately predict the behavior of steel pipes with a welded joint, the influence of welding residual stress and weld metal on the behavior of steel pipe must be investigated. In this paper, the residual stress of steel pipes with a welded joint was investigated by using a three-dimensional non-steady heat conduction analysis and a three-dimensional thermal elastic-plastic analysis. Based on the results of monotonic and cyclic loading tests, a hysteresis model for weld metal was formulated. The hysteresis model was proposed by the authors and applied to a three-dimensional finite elements analysis. To investigate the influence of a welded joint in steel pipes under monotonic and cyclic loading, three-dimensional finite elements analysis considering the proposed model and residual stress was carried out. The influence of a welded joint on the behavior of steel pipe members was investigated by comparing the analytical result both steel pipe with a welded joint and that without a welded joint

  17. Development of Pipe Holding Mechanism for Pipe Inspection Robot Using Flexible Pneumatic Cylinder

    Directory of Open Access Journals (Sweden)

    Choi Kyujun

    2016-01-01

    Full Text Available A pipe inspection robot is useful to reduce the inspection cost. In the previous study, a novel pipe inspection robot using a flexible pneumatic cylinder that can move forward along to the pipe by changing the robot’s body naturally was proposed and tested. In this paper, to improve its mobility for a corner of a pipe, the thin pipe holding mechanism using pneumatic bellows was proposed and tested. As a result of its driving test, the holding performance of the mechanism was confirmed.

  18. Pipe grabber

    Energy Technology Data Exchange (ETDEWEB)

    Sharafutdinov, I.G.; Mubashirov, S.G.; Prokopov, O.I.

    1981-05-15

    A pipe grabber is suggested which contains a housing, clamping elements and centering mechanism with drive installed on the lower end of the housing. In order to improve the reliable operation of the pipe grabber, the centering mechanism is made in the form of a reinforced ringed flexible shaft, while the drive is made in the form of elastic rotating discs. In this case the direction of rotation of the discs and the flexible shaft is the opposite.

  19. Automated numerical simulation of cracked plates, pipes and elbows

    International Nuclear Information System (INIS)

    Reddy, Babu; Sreehari Kumar, B.; Bhate, S.R.; Kushwaha, H.S.

    2008-01-01

    In the nuclear industry, piping components are one of the key elements participating in its operation. Integrity of structural tubes and pipes plays a major role in nuclear power plants. The ideal procedure to ensure this aspect would be to conduct experimental studies on pilot/test specimens. However, it may not always be feasible to carry out the experimental investigation, as it requires pre-requisite infrastructure which may not be economically viable. This makes it imperative to conduct numerical simulations of the same particularly in the study of presence of cracks in the critical components. While performing the effect of cracks, the quality of the finite element mesh nearer to the crack tip plays a critical role while estimating J-integral value. The designer is often familiar with design methodology only and he obviously requires a convenient and reliable numerical tool to model and perform the analysis. In this context, an effort has been made in NISA, the general purpose finite element software, to automate the generation of FE meshes for a set of pre-defined components with different crack configurations. To simplify the procedure of FE mesh generation, analysis, and post processing, a graphical user interface (GUI) has been developed accordingly. This paper discusses the automated numerical simulation of plates and pipes with different crack configurations. This simulation software is also designed to help parametric study of cracked pipes. (author)

  20. Effect of pipe rupture loads inside containment in the break exclusionary piping outside containment

    International Nuclear Information System (INIS)

    Weiss, G.

    1987-01-01

    The plant design for protection against piping failures outside containment should make sure that fluid system piping in containment penetration areas are designed to meet the break exclusionary provisions contained in the BTP MEB 3-1. According to these provisions, following a piping failure (main steam line) inside containment, the part of the flued head connected to the piping outside containment, should not exceed the ASME Code stress limits for the appropriate load combinations. A finite element analysis has been performed to evaluate the stress level in this area. (orig./HP)

  1. Inelastic analysis of SNR-300 piping

    International Nuclear Information System (INIS)

    Huebel, H.; Di Luna, L.J.; Moy, G.

    1983-01-01

    This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

  2. Inelastic analysis of SNR-300 piping

    Energy Technology Data Exchange (ETDEWEB)

    Huebel, H [INTERATOM, Bergisch Gladbach (Germany); Di Luna, L J; Moy, G [Teledyne Engineering Services, Waltham, MA (United States)

    1983-05-01

    This paper investigates plasticity, creep, and elastic follow-up effects on a full size hot primary piping system of the German fast breeder reactor prototype, the SNR-300. A large model (327 elements, 419 nodes) of straight pipe, special elbow and hanger elements of the general purpose finite element program, MARC-CDC, is used to predict piping behavior for a heat-up, sodium loading-unloading-reloading cycle and other significant operating conditions. Included in this work are many time-dependent solution increments for a 5,000 hour creep period. Creep strains and relaxed stress results, after 5,000 hours, for the complete model are used with uniaxial and biaxial models and results to extrapolate conclusions for a 100,000 hour operating life. (author)

  3. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  4. Vibration test on KMRR reactor structure and primary cooling system piping

    International Nuclear Information System (INIS)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

  5. Failure pressure of straight pipe with wall thinning under internal pressure

    International Nuclear Information System (INIS)

    Kamaya, Masayuki; Suzuki, Tomohisa; Meshii, Toshiyuki

    2008-01-01

    The failure pressure of pipe with wall thinning was investigated by using three-dimensional elastic-plastic finite element analyses (FEA). With careful modeling of the pipe and flaw geometry in addition to a proper stress-strain relation of the material, FEA could estimate the precise burst pressure obtained by the tests. FEA was conducted by assuming three kinds of materials: line pipe steel, carbon steel, and stainless steel. The failure pressure obtained using line pipe steel was the lowest under the same flaw size condition, when the failure pressure was normalized by the value of unflawed pipe defined using the flow stress. On the other hand, when the failure pressure was normalized by the results of FEA obtained for unflawed pipe under various flaw and pipe configurations, the failure pressures of carbon steel and line pipe steel were almost the same and lower than that of stainless steel. This suggests that the existing assessment criteria developed for line pipe steel can be applied to make a conservative assessment of carbon steel and stainless steel

  6. A novel numerical model for estimating the collapse pressure of flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Nogueira, Victor P.P.; Antoun Netto, Theodoro [Universidade Federal do Rio de Janeiro (COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-graduacao em Engenharia], e-mail: victor@lts.coppe.ufrj.br

    2009-07-01

    As the worldwide oil and gas industry operational environments move to ultra-deep waters, failure mechanisms in flexible pipes such as instability of the armor layers under compression and hydrostatic collapse are more likely to occur. Therefore, it is important to develop reliable numerical tools to reproduce the failure mechanisms that may occur in flexible pipes. This work presents a representative finite element model of flexible pipe capable to reproduce its pre and post-collapse behavior under hydrostatic pressure. The model, developed in the scope of this work, uses beam elements and includes nonlinear kinematics and material behavior influences. The dependability of the numerical results is assessed in light of experimental tests on flexible pipes with 4 inches and 8 inches nominal diameter available in the literature (Souza, 2002). The applied methodology provided coherent values regarding the estimation of the collapse pressures and results have shown that the proposed model is capable to reproduce experimental results. (author)

  7. Design Evaluation of a Piping System in the SELFA Sodium Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seok-Kwon; Jo, Young-Chul; Lee, Hyeong-Yeon; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, design evaluations on the SELFA piping system has been conducted according to the ASME B31.1 and RCC-MRx RD-3600. The conservatism of the two codes was quantified based on the evaluation results. It was shown that B31.1 was more conservative for the sustained loads while less conservative for thermal expansion loads when compare with those of RD-3600. However, all the evaluation results according to the two codes were within the code allowables. There are two main piping systems in the SELFA test loop. In this study, the integrity of the SELFA piping system has been evaluated according to the two design-by-rule (DBR) codes of ASME B31.1 and RCC-MRx RD-3600. B31.1 is an industry design code for power piping while RD-3600 is a class 3 nuclear DBR code. The conservatism of the two codes was quantified based on the evaluation results as per the two DBR codes. The sodium test facility of the SELFA is under construction at KAERI for the investigation of thermo-hydraulic behavior of finned-tube sodium-to-air heat exchanger.

  8. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. The design considerations and methods along with the development tests are presented. Special considerations to guard against adverse cracking of the insulation material, to maintain the clamp-pipe stiffness desired during a seismic event, to minimize clamp restraint on the pipe during normal pipe heatup, and to resist clamp rotation or spinning on the pipe are emphasized

  9. Comparison of fracture toughness values from large-scale pipe system tests and C(T) specimens

    International Nuclear Information System (INIS)

    Olson, R.; Scott, P.; Marschall, C.; Wilkowski, G.

    1993-01-01

    Within the International Piping Integrity Research Group (IPIRG) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe and the experiments were conducted at 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment of standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria. 4 refs., 14 figs., 1 tab

  10. Assessment of Pipe Wall Loss Using Guided Wave Testing

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Jin, Seuk Hong; Moon, Yong Sig

    2010-01-01

    Flow accelerated corrosion(FAC) of carbon steel pipes in nuclear power plants has been known as one of the major degradation mechanisms. It could have bad influence on the plant reliability and safety. Also detection of FAC is a significant cost to the nuclear power plant because of the need to remove and replace insulation. Recently, the interest of the guided wave testing(GWT) has grown because it allows long range inspection without removing insulation of the pipe except at the probe position. If GWT can be applied to detection of FAC damages, it will can significantly reduce the cost for the inspection of the pipes. The objective of this study was to determine the capability of GWT to identify location of FAC damages. In this paper, three kinds of techniques were used to measure the amplitude ratio between the first and the second welds at the elbow area of mock-ups that contain real FAC damages. As a result, optimal inspection technique and minimum detectability to detect FAC damages drew a conclusion

  11. Fabrication of a multi-walled metal pipe

    International Nuclear Information System (INIS)

    Shimamune, Koji; Toda, Saburo; Ishida, Ryuichi; Hatanaka, Tatsuo.

    1969-01-01

    In concentrically arranged metal pipes for simulated fuel elements in the form of a multi-walled pipe, their one end lengthens gradually in the axial direction from inner and outer pipes toward a central pipe for easy adjustment of deformation which occurs when the pipes are drawn. A plastic electrical insulator is disposed between adjacent pipes. Each end of the pipes is equipped with an annular flexible stopper which is allowed to travel in the axial direction so as to prevent the insulator from falling during drawing work. At the other end, all pipes are constricted and joined to each other to thereby form the desired multi-walled pipe. (Mikami, T.)

  12. Incremental-hinge piping analysis methods for inelastic seismic response prediction

    International Nuclear Information System (INIS)

    Jaquay, K.R.; Castle, W.R.; Larson, J.E.

    1989-01-01

    This paper proposes nonlinear seismic response prediction methods for nuclear piping systems based on simplified plastic hinge analyses. The simplified plastic hinge analyses utilize an incremental series of flat response spectrum loadings and replace yielded components with hinge elements when a predefined hinge moment is reached. These hinge moment values, developed by Rodabaugh, result in inelastic energy dissipation of the same magnitude as observed in seismic tests of piping components. Two definitions of design level equivalent loads are employed: one conservatively based on the peaks of the design acceleration response spectra, the other based on inelastic frequencies determined by the method of Krylov and Bogolyuboff recently extended by Lazzeri to piping. Both definitions account for piping system inelastic energy dissipation using Newmark-Hall inelastic response spectrum reduction factors and the displacement ductility results of the incremental-hinge analysis. Two ratchet-fatigue damage models are used: one developed by Rodabaugh that conservatively correlates Markl static fatigue expressions to seismic tests to failure of piping components; the other developed by Severud that uses the ratchet expression of Bree for elbows and Edmunds and Beer for straights, and defines ratchet-fatigue interaction using Coffin's ductility based fatigue equation. Comparisons of predicted behavior versus experimental results are provided for a high-level seismic test of a segment of a representative nuclear plant piping system. (orig.)

  13. Design criteria for piping and nozzles program

    International Nuclear Information System (INIS)

    Moore, S.E.; Bryson, J.W.

    1977-01-01

    This report reviews the activities and accomplishments of the Design Criteria for Piping and Nozzles program being conducted by the Oak Ridge National Laboratory for the period July 1, 1975, to September 30, 1976. The objectives of the program are to conduct integrated experimental and analytical stress analysis studies of piping system components and isolated and closely-spaced pressure vessel nozzles in order to confirm and/or improve the adequacy of structural design criteria and analytical methods used to assure the safe design of nuclear power plants. Activities this year included the development of a finite-element program for analyzing two closely spaced nozzles in a cylindrical pressure vessel; a limited-parameter study of vessels with isolated nozzles, finite-element studies of piping elbows, a fatigue test of an out-of-round elbow, summary and evaluation of experimental studies on the elastic-response and fatigue failure of tees, parameter studies on the behavior of flanged joints, publication of fifteen topical reports and papers on various experimental and analytical studies; and the development and acceptance of a number of design rules changes to the ASME Code. 2 figures, 2 tables

  14. Vibration monitoring of the primary piping systems during the hot functional tests of the Mulheim-Karlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1989-01-01

    During the hot functional tests of the Muelheim--Kaerlich first-of-a-kind plant, vibration measurements were made on the reactor pressure vessel and its' internals and on the primary piping system and main coolant pumps. This paper contains results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement program is to confirm that the components, which are of new design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. Good agreement was found. In the course of these comparisons, information on the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained

  15. Parametric Study on Ultimate Failure Criteria of Elbow Piping Components in Seismically Isolated NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Ki, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    It is well known that the interface pipes between isolated and non-isolated structures will become the most critical in the seismically isolated NPPs. Therefore, seismic performance of such interface pipes should be evaluated comprehensively especially in terms of the seismic fragility capacity. To evaluate the seismic capacity of interface pipes in the isolated NPP, firstly, we should define the failure mode and failure criteria of critical pipe components. Hence, in this study, we performed the dynamic tests of elbow components which were installed in a seismically isolated NPP, and evaluated the ultimate failure mode and failure criteria by using the test results. To do this, we manufactured 25 critical elbow component specimens and performed cyclic loading tests under the internal pressure condition. The failure mode and failure criteria of a pipe component will be varied by the design parameters such as the internal pressure, pipe diameter, loading type, and loading amplitude. From the tests, we assessed the effects of the variation parameters onto the failure criteria. For the tests, we generated the seismic input protocol of relative displacement between the ends of elbow component. In this paper, elbow in piping system was defined as a fragile element and numerical model was updated by component test. Failure mode of piping component under seismic load was defined by the dynamic tests of ultimate pipe capacity. For the interface piping system, the seismic capacity should be carefully estimated since that the required displacement absorption capacity will be increased significantly by the adoption of the seismic isolation system. In this study, the dynamic tests were performed for the elbow components which were installed in an actual NPPs, and the ultimate failure mode and failure criteria were also evaluated by using the test results.

  16. Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

    International Nuclear Information System (INIS)

    Cunha, Sérgio B.; Netto, Theodoro A.

    2012-01-01

    The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3″ diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: ► An analytical model for the burst of a pipe with a volumetric flaw is developed. ► Deformation, strain and stress are modeled in the elastic and plastic domains. ► The model is comprehensively validated by experiments and numerical simulations. ► The burst pressure model’s accuracy is equivalent to finite element simulations.

  17. Design and testing of a heat pipe gas combustion system for the STM4-120 Stirling engine

    Science.gov (United States)

    Khalili, K.; Godett, T. M.; Meijer, R. J.; Verhey, R. P.

    Evaporators of a novel geometry, designed to have a more compact size yet the same output as larger, conventional heat pipes, have been fabricated and tested. A technique was developed to calculate capillary pressure required inside the heat pipe. Several quarter- and full-scale evaporators were designed and successfully tested. The burner, film-cooled combustion chamber, and preheater were designed and tested separately. A complete heat pipe gas combustion system (HPGC) was tested, showing an efficiency of 89 percent was measured at 20 kWth. A film-cooled combustion chamber was tested with flame temperatures of 2200 C and wall temperatures below 1000 C using preheated air for film cooling. Also, a full-scale HPGC was tested at an excess of 95 kWth, showing efficiency in the range of 85 to 90 percent under steady-state conditions. Results of transient and startup tests, carried out to evaluate the performance of the heat pipe, all also reported.

  18. Heat pipe nuclear reactor for space power

    Science.gov (United States)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  19. Mechanical Interaction in Pressurized Pipe Systems: Experiments and Numerical Models

    Directory of Open Access Journals (Sweden)

    Mariana Simão

    2015-11-01

    Full Text Available The dynamic interaction between the unsteady flow occurrence and the resulting vibration of the pipe are analyzed based on experiments and numerical models. Waterhammer, structural dynamic and fluid–structure interaction (FSI are the main subjects dealt with in this study. Firstly, a 1D model is developed based on the method of characteristics (MOC using specific damping coefficients for initial components associated with rheological pipe material behavior, structural and fluid deformation, and type of anchored structural supports. Secondly a 3D coupled complex model based on Computational Fluid Dynamics (CFD, using a Finite Element Method (FEM, is also applied to predict and distinguish the FSI events. Herein, a specific hydrodynamic model of viscosity to replicate the operation of a valve was also developed to minimize the number of mesh elements and the complexity of the system. The importance of integrated analysis of fluid–structure interaction, especially in non-rigidity anchored pipe systems, is equally emphasized. The developed models are validated through experimental tests.

  20. Evaluation of fracture mode for local wall-thinned pipes

    International Nuclear Information System (INIS)

    Herman, Irwan; Suzuki, Tomohisa; Sato, Yasumoto; Meshii, Toshiyuki

    2007-01-01

    In this study, by referring to our burst pressure tests results, firstly, the effects of flaw length δ z and pipe size (mean radius R) on burst pressure p f were investigated by using Finite Element Method (FEM). Then, fracture mode evaluation was made by using history data of strain ratio ε z /ε θ along with load increment. Furthermore, the effect of flaw depth t 1 on fracture mode was studied and finally, the evaluation method of fracture mode for local wall-thinned pipes was introduced. (author)

  1. Heat pipes with variable thermal conductance property for space applications

    Energy Technology Data Exchange (ETDEWEB)

    Kravets, V.; Alekseik, Ye.; Alekseik, O.; Khairnasov, S. [National Technical University of Ukraine, Kyiv (Ukraine); Baturkin, V.; Ho, T. [Explorationssysteme RY-ES, Bremen (Germany); Celotti, L. [Active Space Technologies GmbH, Berlin (Germany)

    2017-06-15

    The activities presented in this paper demonstrate a new approach to provide passive thermal control using heat pipes, as demonstrated on the electronic unit of DLR’s MASCOT lander, which embarked on the NEA sample return mission Hayabusa 2 (JAXA). The focus is on the development and testing of heat pipes with variable thermal conductance in a predetermined temperature range. These heat pipes act as thermal switches. Unlike standard gasloaded heat pipes and thermal-diode heat pipes construction of presented heat pipes does not include any additional elements. Copper heat pipes with metal fibrous wicks were chosen as baseline design. We obtained positive results by choosing the heat carrier and structural parameters of the wick (i.e., pore diameter, porosity, and permeability). The increase in the thermal conductivity of the heat pipes from 0.04 W/K to 2.1 W/K was observed in the temperature range between −20 °C and +55 °C. Moreover, the heat pipes transferred the predetermined power of not less than 10 W within the same temperature range. The heat pipes have been in flight since December 2014, and the supporting telemetry data were obtained in September 2015. The data showed the nominal operation of the thermal control system.

  2. Valve for the mechanical isolation of a pipe to take up a test probe

    International Nuclear Information System (INIS)

    Uecker, D.F.

    1976-01-01

    A valve is introduced for application in a pipe in which a test probe is arranged. The valve serves to isolate the pipe in a gas-tight way, thus preventing the escape of radioactive gas or dust during operation in a nuclear reactor. (TK) [de

  3. Ratcheting of pressurized piping subjected to seismic loading

    International Nuclear Information System (INIS)

    Scavuzzo, R.J.; Lam, P.C.; Gau, J.S.

    1992-01-01

    The ABAQUS finite element code was used to model a pressurized pipe and subjected to cyclic bending loads to investigate ratcheting. A 1-in. schedule 40 pipe was loaded with a slow (static) cyclic load. The pipe internal pressure was varied from 0 to 6000 psi. In this paper, two types of materials were considered: an elastic perfectly plastic and a bilinear elastic-plastic material. Two types of finite elements of the ABAQUS program were compared to analytical solutions to evaluate the element accuracy in the plastic regime. Depending upon loading conditions and specified material properties, three different responses were observed from the finite element analyses: cyclic plasticity, ratcheting of the hoop strain, or shakedown. These analytical results are compared to some experimental measurements

  4. Numerical simulation of X90 UOE pipe forming process

    Science.gov (United States)

    Zou, Tianxia; Ren, Qiang; Peng, Yinghong; Li, Dayong; Tang, Ding; Han, Jianzeng; Li, Xinwen; Wang, Xiaoxiu

    2013-12-01

    The UOE process is an important technique to manufacture large-diameter welding pipes which are increasingly applied in oil pipelines and offshore platforms. The forming process of UOE mainly consists of five successive operations: crimping, U-forming, O-forming, welding and mechanical expansion, through which a blank is formed into a pipe in a UOE pipe mill. The blank with an appropriate edge bevel is bent into a cylindrical shape by crimping (C-forming), U-forming and O-forming successively. After the O-forming, there is an open-seam between two ends of the plate. Then, the blank is welded by automatic four-electrode submerged arc welding technique. Subsequently, the welded pipe is expanded with a mechanical expander to get a high precision circular shape. The multiple operations in the UOE mill make it difficult to control the quality of the formed pipe. Therefore, process design mainly relies on experience in practical production. In this study, the UOE forming of an API X90 pipe is studied by using finite element simulation. The mechanical properties tests are performed on the API X90 pipeline steel blank. A two-dimensional finite element model under the hypothesis of plane strain condition is developed to simulate the UOE process according to data coming from the workshop. A kinematic hardening model is used in the simulation to take the Bauschinger effect into account. The deformation characteristics of the blank during the forming processes are analyzed. The simulation results show a significant coherence in the geometric configurations comparing with the practical manufacturing.

  5. Test method for measuring insulation values of cryogenic pipes

    NARCIS (Netherlands)

    Velthuis, J.F.M.; Blokland, H.; Klaver, B.W.; Beld, C. van de

    2010-01-01

    In this paper a large-area heat flux and temperature sensor (HFT) is used for the evaluation of the insulation value of cryogenic pipes. The HFT is flexible and clamp-on. The test method is relatively simple and can be used in-situ. The HFT makes it possible to monitor insulation performance over

  6. Damping in LMFBR pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Barta, D.A.; Lindquist, M.R.; Renkey, E.J.; Ryan, J.A.

    1983-06-01

    LMFBR pipe systems typically utilize a thicker insulation package than that used on water plant pipe systems. They are supported with special insulated pipe clamps. Mechanical snubbers are employed to resist seismic loads. Recent laboratory testing has indicated that these features provide significantly more damping than presently allowed by Regulatory Guide 1.61 for water plant pipe systems. This paper presents results of additional in-situ vibration tests conducted on FFTF pipe systems. Pipe damping values obtained at various excitation levels are presented. Effects of filtering data to provide damping values at discrete frequencies and the alternate use of a single equivalent modal damping value are discussed. These tests further confirm that damping in typical LMFBR pipe systems is larger than presently used in pipe design. Although some increase in damping occurred with increased excitation amplitude, the effect was not significant. Recommendations are made to use an increased damping value for both the OBE and DBE seismic events in design of LMFBR pipe systems

  7. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    To safely assess the adequacy of the LMR piping, a three-dimensional piping code, SHAPS, has been developed at Argonne National Laboratory. This code was initially intended for calculating hydrodynamic-wave propagation in a complex piping network. It has salient features for treating fluid transients of fluid-structure interactions for piping with in-line components. The code also provides excellent structural capabilities of computing stresses arising from internal pressurization and 3-D flexural motion of the piping system. As part of the development effort, the SHAPS code has been further augmented recently by introducing the capabilities of calculating piping response subjected to seismic excitations. This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis

  8. Effects of Cross-Linking on the Hydrostatic Pressure Testing for HDPE Pipe Material using Electron Beam Machine

    International Nuclear Information System (INIS)

    Mohd Jamil Bin Hashim

    2011-01-01

    One of the most inventive, sustainable strategies used in engineering field is to improve the quality of material and minimize production cost of material for example in this paper is HDPE material. This is because HDPE is an oil base material. This paper proposes to improve its hydrostatic pressure performance for HDPE pipe. The burst test is the most direct measurement of a pipe materials resistance to hydrostatic pressure. Test will be conducted in accordance with ASTM standard for HDPE pipe that undergo electron beam irradiation cross-linking. Studies show the effect of electron beam irradiation will improve the mechanical properties of HDPE pipe. When cross-linking is induced, the mechanical properties such as tensile strength and young modulus is increase correspond to the radiation dose. This happen because the structure of HDPE, which is thermoplastic change to thermosetting. This will indicate the variability of irradiation dose which regard to the pipe pressure rating. Hence, the thickness ratio of pipe will be re-examining in order to make the production of HDPE pipe become more economical. This research review the effects of electron beam on HDPE pipe, as well as to reduce the cost of its production to improve key properties of selected plastic pipe products. (author)

  9. Pipe Overpack Container Fire Testing: Phase I & II

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, Victor G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lopez, Carlos [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2017-05-01

    The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. However, POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a new series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016, and described herein, were done in two phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. This report will describe the various tests conducted in phase I and II, present preliminary results from these tests, and discuss implications for the POCs.

  10. Analytical studies of blowdown thrust force and dynamic response of pipe at pipe rupture accident

    International Nuclear Information System (INIS)

    Miyazaki, Noriyuki

    1985-01-01

    The motion of a pipe due to blowdown thrust when the pipe broke is called pipe whip. In LWR power plants, by installing restraints, the motion of a pipe when it broke is suppressed, so that the damage does not spread to neighboring equipment by pipe whip. When the pipe whip of a piping system in a LWR power plant is analyzed, blowdown thrust and the dynamic response of a pipe-restraint system are calculated with a computer. The blowdown thrust can be calculated by using such physical quantities as the pressure, flow velocity, density and so on in the system at the time of blowdown, obtained by the thermal-fluid analysis code at LOCA. The dynamic response of a piping-restraint system can be determined by the stress analysis code using finite element method taking the blowdown thrust as an external force acting on the piping. In this study, the validity of the analysis techniques was verified by comparing with the experimental results of the measurement of blowdown thrust and the pipe whip of a piping-restraint system, carried out in the Japan Atomic Energy Research Institute. Also the simplified analysis method to give the maximum strain on a pipe surface is presented. (Kako, I.)

  11. Development of the monitoring system to detect the piping thickness reduction

    International Nuclear Information System (INIS)

    Lee, N. Y.; Ryu, K. H.; Oh, Y. J.; Hwang, I. S.

    2006-01-01

    As nuclear piping becomes aging, secondary piping which was considered safe, undergo thickness reduction problem these days. After some accidents caused by Flow Accelerated Corrosion (FAC), guidelines and recommendations for the thinned pipe management were issued, and thus need for monitoring increases. Through thinned pipe management program, monitoring activities based on the various analyses and the case study of other plants also increases. As the monitoring points increase, time needs to cover the recommended inspection area becomes increasing, while the time given to inspect the piping during overhaul becomes shortened. Existing Ultrasonic Technique (UT) can cover small area in a given time. Moreover, it cannot be applied to a complex geometry piping or a certain location like welded part. In this paper, we suggested Switching Direct Current Potential Drop (S-DCPD) method by which we can narrow down the FAC-susceptible area. To apply DCPD, we developed both resistance model and Finite Element Method (FEM) model to predict the DCPD feasibility. We tested elbow specimen to compare DCPD monitoring results with UT results to identify consistency. For the validation test, we designed simulation loop. To determine the text condition, we analyzed environmental parameters and introduced applicable wearing rate model. To obtain the model parameters, we developed electrodes and analyzed velocity profile in the test loop using CFX code. Based on the prediction model and prototype testing results, we are planning to perform validation test to identify applicability of S-DCPD in the NPP environment. Validation text plan will be described as a future work. (authors)

  12. Automated ultrasonic pipe weld inspection. Part 1

    International Nuclear Information System (INIS)

    Karl Deutsch, W.A.; Schulte, P.; Joswig, M.; Kattwinkel, R.

    2006-01-01

    This article contains a brief overview on automated ultrasonic welded inspection for various pipe types. Some inspection steps might by carried out with portable test equipment (e.g. pipe and test), but the weld inspection in all internationally relevant specification must be automated. The pipe geometry, the production process, and the pipe usage determine the number of required probes. Recent updates for some test specifications enforce a large number of ultrasonic probes, e.g. the Shell standard. Since seamless pipes are sometimes replaced by ERW pipes and LSAW pipes (in both cases to save production cost), the inspection methods change gradually between the various pipe types. Each testing system is unique and shows its specialties which have to be discussed by supplier, testing system user and final customer of the pipe. (author)

  13. Elastic-plastic dynamic behavior of guard pipes due to sudden opening of longitudinal cracks in the inner pipe and crash to the guard pipe wall

    International Nuclear Information System (INIS)

    Theuer, E.; Heller, M.

    1979-01-01

    Integrity of guard pipes is an important parameter in the design of nuclear steam supply systems. A guard pipe shall withstand all kinds of postulated inner pipe breaks without failure. Sudden opening of a crack in the inner pipe and crash of crack borders to the guard pipe wall represent a shock problem where complex phenomena of dynamic plastification as well as dynamic behavior of the entire system have to be taken in consideration. The problem was analyzed by means of Finite Element computation using the general purpose program MARC. Equation of motion was resolved by direct integration using the Newmark β-operator. Analysis shows that after 1,2 m sec crack borders touch the guard pipe wall for the first time. At this moment a considerable amount of local plastification appears in the inner pipe wall, while the guard pipe is nearly unstressed. After initial touching, the crack borders begin to slip along the guard pipe wall. Subsequently, a short withdrawal of the crack borders and a new crash occur, while the inner pipe rolls along the guard pipe wall. The analysis procedure described is suitable for designing numerous guard pipe geometries as well as U-Bolt restraint systems which have to withstand high-energy pipe rupture impact. (orig.)

  14. Damping test results for straight sections of 3-inch and 8-inch unpressurized pipes. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Ware, A.G.; Thinnes, G.L.

    1984-04-01

    EG and G Idaho is assisting the Nuclear Regulatory Commission and the Pressure Vessel Research Committee in supporting a final position on revised damping values for structural analyses of nuclear piping systems. As part of this program, a series of vibrational tests on unpressurized 3-in. and 8-in. Schedule 40 carbon steel piping was conducted to determine the changes in structural damping due to various parametric effects. The 33-ft straight sections of piping were supported at the ends. Additionally, intermediate supports comprising spring, rod, and constant-force hangers, as well as a sway brace and snubbers, were used. Excitation was provided by low-force-level hammer impacts, a hydraulic shaker, and a 50-ton overhead crane for snapback testing. Data was recorded using acceleration, strain, and displacement time histories. This report presents test results showing the effect of stress level and type of supports on structural damping in piping.

  15. Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

    Energy Technology Data Exchange (ETDEWEB)

    Cunha, Sergio B., E-mail: sbcunha@petrobras.com.br [PETROBRAS/TRANSPETRO, Av. Pres. Vargas 328 - 7th floor, Rio de Janeiro, RJ 20091-060 (Brazil); Netto, Theodoro A., E-mail: tanetto@lts.coppe.ufrj.br [COPPE, Federal University ot Rio de Janeiro, Ocean Engineering Department, PO BOX 68508, Rio de Janeiro - RJ (Brazil)

    2012-01-15

    The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3 Double-Prime diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: Black-Right-Pointing-Pointer An analytical model for the burst of a pipe with a volumetric flaw is developed. Black-Right-Pointing-Pointer Deformation, strain and stress are modeled in the elastic and plastic domains. Black-Right-Pointing-Pointer The model is comprehensively validated by experiments and numerical simulations. Black-Right-Pointing-Pointer The burst pressure model's accuracy is equivalent to finite element simulations.

  16. Ratcheting study in pressurized piping components under cyclic loading at room temperature

    International Nuclear Information System (INIS)

    Ravi Kiran, A.; Agrawal, M.K.; Reddy, G.R.; Vaze, K.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-07-01

    The nuclear power plant piping components and systems are often subjected to reversing cyclic loading conditions due to various process transients, seismic and other events. Earlier the design of piping subjected to seismic excitation was based on the principle of plastic collapse. It is believed that during such events, fatigue-ratcheting is likely mode of failure of piping components. The 1995 ASME Boiler and Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. Experimental and analytical studies are carried out to understand this failure mechanism. The biaxial ratcheting characteristics of SA 333, Gr. 6 steel and SS 304 stainless steel at room temperature are investigated in the present work. Experiments are carried out on straight pipes subjected to internal pressure and cyclic bending load applied in a three point and four point bend test configurations. A shake table test is also carried out on a pressurized elbow by applying sinusoidal base excitation. Analytical simulation of ratcheting in the piping elements is carried out. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. (author)

  17. Finite element limit analysis based plastic limit pressure solutions for cracked pipes

    International Nuclear Information System (INIS)

    Shim, Do Jun; Huh, Nam Su; Kim, Yun Jae; Kim, Young Jin

    2002-01-01

    Based on detailed FE limit analyses, the present paper provides tractable approximations for plastic limit pressure solutions for axial through-wall cracked pipe; axial (inner) surface cracked pipe; circumferential through-wall cracked pipe; and circumferential (inner) surface cracked pipe. Comparisons with existing analytical and empirical solutions show a large discrepancy in circumferential short through-wall cracks and in surface cracks (both axial and circumferential). Being based on detailed 3-D FE limit analysis, the present solutions are believed to be the most accurate, and thus to be valuable information not only for plastic collapse analysis of pressurised piping but also for estimating non-linear fracture mechanics parameters based on the reference stress approach

  18. Seismic analysis of liquid metal reactor piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1987-01-01

    This paper describes the finite-element numerical algorithm and its applications to LMR piping under seismic excitations. A time-history analysis technique using the implicit temporal integration scheme is addressed. A 3-D pipe element is formulated which has eight degrees of freedom per node (three displacements, three rotations, one membrane displacement, and one bending rotation) to account for the hoop, flexural, rotational, and torsional modes of the piping system. Both geometric and material nonlinearities are considered. This algorithm is unconditionally stable and is particularly suited for the seismic analysis. (orig./GL)

  19. Tests of a 3D Self Magnetic Field Solver in the Finite Element Gun Code MICHELLE

    CERN Document Server

    Nelson, Eric M

    2005-01-01

    We have recently implemented a prototype 3d self magnetic field solver in the finite-element gun code MICHELLE. The new solver computes the magnetic vector potential on unstructured grids. The solver employs edge basis functions in the curl-curl formulation of the finite-element method. A novel current accumulation algorithm takes advantage of the unstructured grid particle tracker to produce a compatible source vector, for which the singular matrix equation is easily solved by the conjugate gradient method. We will present some test cases demonstrating the capabilities of the prototype 3d self magnetic field solver. One test case is self magnetic field in a square drift tube. Another is a relativistic axisymmetric beam freely expanding in a round pipe.

  20. Design, fabrication and performance tests for a polymer-based flexible flat heat pipe

    International Nuclear Information System (INIS)

    Hsieh, Shou-Shing; Yang, Ya-Ru

    2013-01-01

    Highlights: ► Fabrication of a polymer-based flexible flat heat pipe. ► Bending angle of 15° will lead to a better thermal performance of the system. ► Powers higher than 12.67 W can be transferred/delivered. - Abstract: In this paper, we report on the novel design, fabrication and performance tests for a polymer-based flexible flat heat pipe (FHP) with a bending angle in the range of 15–90°. Each heat pipe is 4 mm thick, 20 mm wide and 80 mm long, with two layers of No. 250 copper mesh as the wicking material. A copper/silicone rubber hybrid structure is designed and fabricated to achieve the flexibility of the heat pipe. Thermal characteristics are measured and studied for de-ionized water under different working conditions. Experimental results reveal that a bending angle of 15° on the vertical plane has a better thermal performance than those of heat pipes with/without bending. In addition, a higher power of 12.67 W can be transferred/delivered

  1. Design and Integrity Evaluation of High-temperature Piping Systems in the STELLA-2 Sodium Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    Son, Seok-Kwon; Lee, Hyeong-Yeon; Eoh, JaeHyuk; Kim, Jong-Bum; Jeong, Ji-Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ju, Yong-Sun [KOASIS Inc., Daejeon (Korea, Republic of)

    2016-09-15

    In this study, elevated temperature design and integrity evaluation have been conducted using two different piping design codes for the high-temperature piping systems of sodium integral effect test loop for safety simulation and assessment(STELLA-2) being developed by KAERI(Korea Atomic Energy Research Institute). The design code of ASME B31.1 for power piping and French nuclear grade piping design guideline, RCC-MRx RD-3600 were applied, and conservatism of those codes was quantified based on the piping integrity evaluation results. The piping system of Model DHRS, Model IHTS and PSLS are to be installed in STELLA-2. The integrity evaluation results for the three piping systems according to the two design codes showed that integrity of the piping system was confirmed. As a code comparison result, ASME B31.1 was shown to be more conservative for sustained loads while RD-3600 was more conservative for thermal loads compared to B31.1.

  2. On the impact bending test technique for high-strength pipe steels

    Science.gov (United States)

    Arsenkin, A. M.; Odesskii, P. D.; Shabalov, I. P.; Likhachev, M. V.

    2015-10-01

    It is shown that the impact toughness (KCV-40 = 250 J/cm2) accepted for pipe steels of strength class K65 (σy ≥ 550 MPa) intended for large-diameter gas line pipes is ineffective to classify steels in fracture strength. The results obtained upon testing of specimens with a fatigue crack and additional sharp lateral grooves seem to be more effective. In energy consumption, a macrorelief with splits is found to be intermediate between ductile fracture and crystalline brittle fracture. A split formation mechanism is considered and a scheme is proposed for split formation.

  3. Accelerated life tests of specimen heat pipe from Communication Technology Satellite (CTS) project

    Science.gov (United States)

    Tower, L. K.; Kaufman, W. B.

    1977-01-01

    A gas-loaded variable conductance heat pipe of stainless steel with methanol working fluid identical to one now on the CTS satellite was life tested in the laboratory at accelerated conditions for 14 200 hours, equivalent to about 70 000 hours at flight conditions. The noncondensible gas inventory increased about 20 percent over the original charge. The observed gas increase is estimated to increase operating temperature by about 2.2 C, insufficient to harm the electronic gear cooled by the heat pipes in the satellite. Tests of maximum heat input against evaporator elevation agree well with the manufacturer's predictions.

  4. Assessment of grass root effects on soil piping in sandy soils using the pinhole test

    Science.gov (United States)

    Bernatek-Jakiel, Anita; Vannoppen, Wouter; Poesen, Jean

    2017-10-01

    Soil piping is an important land degradation process that occurs in a wide range of environments. Despite an increasing number of studies on this type of subsurface erosion, the impact of vegetation on piping erosion is still unclear. It can be hypothesized that vegetation, and in particular plant roots, may reduce piping susceptibility of soils because roots of vegetation also control concentrated flow erosion rates or shallow mass movements. Therefore, this paper aims to assess the impact of grass roots on piping erosion susceptibility of a sandy soil. The pinhole test was used as it provides quantitative data on pipeflow discharge, sediment concentration and sediment discharge. Tests were conducted at different hydraulic heads (i.e., 50 mm, 180 mm, 380 mm and 1020 mm). Results showed that the hydraulic head was positively correlated with pipeflow discharge, sediment concentration and sediment discharge, while the presence of grass roots (expressed as root density) was negatively correlated with these pipeflow characteristics. Smaller sediment concentrations and sediment discharges were observed in root-permeated samples compared to root-free samples. When root density exceeds 0.5 kg m- 3, piping erosion rates decreased by 50% compared to root-free soil samples. Moreover, if grass roots are present, the positive correlation between hydraulic head and both sediment discharge and sediment concentration is less pronounced, demonstrating that grass roots become more effective in reducing piping erosion rates at larger hydraulic heads. Overall, this study demonstrates that grass roots are quite efficient in reducing piping erosion rates in sandy soils, even at high hydraulic head (> 1 m). As such, grass roots may therefore be used to efficiently control piping erosion rates in topsoils.

  5. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M. Kothari, Gerard T. Pittard

    2004-01-01

    optimize the design of the robot elements and surface control electronics and software. Task 6 (Design & Build Surface Control and Monitoring System) has been completed with the control and computer display functions being operated through LabView. Task 7 (Design & Fabricate Large Diameter Live Access System) progressed to the design, fabrication and testing of a entry fitting in a 4-inch prototype and is now being used to complete drawings for use in 12-inch diameter pipe. Task 8--System Integration and Laboratory Validation continued developing the robot module inter-connects and development of a master LabView-based system display and control software.

  6. APPLICATION OF STEEL PIPE PILE LOADING TESTS TO DESIGN VERIFICATION OF FOUNDATION OF THE TOKYO GATE BRIDGE

    Science.gov (United States)

    Saitou, Yutaka; Kikuchi, Yoshiaki; Kusakabe, Osamu; Kiyomiya, Osamu; Yoneyama, Haruo; Kawakami, Taiji

    Steel sheet pipe pile foundations with large diameter steel pipe sheet pile were used for the foundation of the main pier of the Tokyo Gateway bridge. However, as for the large diameter steel pipe pile, the bearing mechanism including a pile tip plugging effect is still unclear due to lack of the practical examinations even though loading tests are performed on Trans-Tokyo Bay Highway. In the light of the foregoing problems, static pile loading tests both vertical and horizontal directions, a dynamic loading test, and cone penetration tests we re conducted for determining proper design parameters of the ground for the foundations. Design parameters were determined rationally based on the tests results. Rational design verification was obtained from this research.

  7. Analysis of noncoplanar pressurized laminations in X2 steel pipes by non-linear finite element

    Energy Technology Data Exchange (ETDEWEB)

    Morales, Alfredo [Instituto Tecnologico de Puebla (Mexico). Dept. de Posgrado; Gonzalez, Jorge L.; Hallen, Jose M. [Instituto Politecnico Nacional (Mexico). Escuela Superior de Ingenieria Quimica e Industrias Extractivas (ESIQIE). Dept. de Ingenieria Metalurgica

    2005-07-01

    Hydrogen induced cracking is of great interest in the mechanical integrity assessment of sour gas pipelines. Multiple stepwise cracks with internal pressure called laminations are often observed in pipelines and their interaction and coalescence may significantly affect the residual strength of the pipes. In this work, the interacting fields of non coplanar pressurized laminations in the wall of a pipe under pressure are analyzed by non-lineal finite element, considering an isotropic hardening law and the real tensile properties of the X52 steel. The results are presented as the evolution of the stress fields in the interlaminar region as a function of the pressure inside the laminations. It is found that for two approaching stepwise laminations the critical pressure follows a hyperbolic type law, thus the effect of the lamination length is principal for greater lengths and for shorter lengths the effect is minimum. The critical pressure is defined as pressure inside the lamination that causes plastification of the interlaminar region. (author)

  8. Testing in support of on-site storage of residues in the Pipe Overpack Container

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Bobbe, J.G.; Arviso, M.

    1997-02-01

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plans call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. It is desirable to use this same waste packaging for interim on-site storage in non-hardened buildings. To meet the safety concerns for this storage the Pipe Overpack Container has been subjected to a series of tests at Sandia National Laboratories in Albuquerque, New Mexico. In addition to the tests required to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II several tests were performed solely for the purpose of qualifying the container for interim storage. This report will describe these tests and the packages response to the tests. 12 figs., 3 tabs

  9. Thermostructural applications of heat pipes for cooling leading edges of high-speed aerospace vehicles

    Science.gov (United States)

    Camarda, Charles J.; Glass, David E.

    1992-01-01

    Heat pipes have been considered for use on wing leading edge for over 20 years. Early concepts envisioned metal heat pipes cooling a metallic leading edge. Several superalloy/sodium heat pipes were fabricated and successfully tested for wing leading edge cooling. Results of radiant heat and aerothermal testing indicate the feasibility of using heat pipes to cool the stagnation region of shuttle-type space transportation systems. The test model withstood a total seven radiant heating tests, eight aerothermal tests, and twenty-seven supplemental radiant heating tests. Cold-wall heating rates ranged from 21 to 57 Btu/sq ft-s and maximum operating temperatures ranged from 1090 to 1520 F. Follow-on studies investigated the application of heat pipes to cool the stagnation regions of single-stage-to-orbit and advanced shuttle vehicles. Results of those studies indicate that a 'D-shaped' structural design can reduce the mass of the heat-pipe concept by over 44 percent compared to a circular heat-pipe geometry. Simple analytical models for heat-pipe startup from the frozen state (working fluid initially frozen) were adequate to approximate transient, startup, and steady-state heat-pipe performance. Improvement in analysis methods has resulted in the development of a finite-element analysis technique to predict heat-pipe startup from the frozen state. However, current requirements of light-weight design and reliability suggest that metallic heat pipes embedded in a refractory composite material should be used. This concept is the concept presently being evaluated for NASP. A refractory-composite/heat-pipe-cooled wing leading edge is currently being considered for the National Aero-Space Plane (NASP). This concept uses high-temperature refractory-metal/lithium heat pipes embedded within a refractory-composite structure and is significantly lighter than an actively cooled wing leading edge because it eliminates the need for active cooling during ascent and descent. Since the

  10. Piping damping tests evaluating influence of types of support and excitation

    International Nuclear Information System (INIS)

    Arendts, J.G.; Ware, A.G.; Gorman, V.W.

    1985-01-01

    The United States Nuclear Regulatory Commission and the Electric Power Research Institute have jointly sponsored construction of two laboratory piping systems at the ANCO Engineers facility in California. EG and G Idaho used the second of these systems to obtain piping system damping data using different supports and methods of excitation. The 6-in. carbon steel piping system was approximately 50 ft in length with two 3-in. branch lines. It was supported at five locations and excited using a single electrohydraulic shaker. Both random and swept sine methods of excitations were used. A variable support attached near the shaker location allowed four different configurations to be tested: a rigid strut, a mechanical snubber, a hydraulic snubber, and a rigid strut with a gap. Data were recorded for the lowest nine significant modes. Damping for the first three modes ranged for 1 to 3% of critical damping and decreased as frequency increased. The random excitation produced a slightly higher average overall damping of the system

  11. A failure estimation method of steel pipe elbows under in-plane cyclic loading

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Park, Dong Uk [Seismic Simulation Tester Center, Pusan National University, Yangsan (Korea, Republic of); Kim, Nam Sik [Dept. of Civil and Environmental Engineering, Pusan National University, Busan (Korea, Republic of)

    2017-02-15

    The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation.

  12. A Failure Estimation Method of Steel Pipe Elbows under In-plane Cyclic Loading

    Directory of Open Access Journals (Sweden)

    Bub-Gyu Jeon

    2017-02-01

    Full Text Available The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation.

  13. A failure estimation method of steel pipe elbows under in-plane cyclic loading

    International Nuclear Information System (INIS)

    Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Park, Dong Uk; Kim, Nam Sik

    2017-01-01

    The relative displacement of a piping system installed between isolated and nonisolated structures in a severe earthquake might be larger when without a seismic isolation system. As a result of the relative displacement, the seismic risks of some components in the building could increase. The possibility of an increase in seismic risks is especially high in the crossover piping system in the buildings. Previous studies found that an elbow which could be ruptured by low-cycle ratcheting fatigue is one of the weakest elements. Fatigue curves for elbows were suggested based on component tests. However, it is hard to find a quantitative evaluation of the ultimate state of piping elbows. Generally, the energy dissipation of a solid structure can be calculated from the relation between displacement and force. Therefore, in this study, the ultimate state of the pipe elbow, normally considered as failure of the pipe elbow, is defined as leakage under in-plane cyclic loading tests, and a failure estimation method is proposed using a damage index based on energy dissipation

  14. Dynamic experiments on cracked pipes

    International Nuclear Information System (INIS)

    Petit, M.; Brunet, G.; Buland, P.

    1991-01-01

    In order to apply the leak before break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic loading must be studied. In a first phase, an experimental program on cracked stainless steel pipes under quasi-static monotonic loading has been conducted. In this paper, the dynamic tests on the same pipe geometry are described. These tests have been performed on a shaking table with a mono frequency input signal. The main parameter of the tests is the frequency of excitation versus the frequency of the system

  15. Development and testing of passive autocatalytic recombiners cooled by heat pipes

    International Nuclear Information System (INIS)

    Granzow, Christoph

    2012-01-01

    A severe accident in a nuclear power plant (NPP) can lead to core damage in conjunction with the release of large amounts of hydrogen. As hydrogen mitigation measure, passive autocatalytic recombiners (PARs) are used in today's pressurized water reactors. PARs recombine hydrogen and oxygen contained in the air to steam. The heat from this exothermic reaction causes the catalyst and its surroundings to heat up. If parts of the PAR heat up above the ignition temperature of the gas mixture, a spontaneous deflagration or detonation can occur. The aim of this work is the prevention of such high temperatures by means of passive cooling of the catalyst with heat pipes. Heat pipes are completely passive heat exchanger with a very high effective thermal conductivity. For a deeper understanding of the reaction kinetics at lower temperatures, single catalytic coated heat pipes are studied in a flow reactor. The development of a modular small-scale PAR model is then based on a test series with cooled catalyst sheets. Finally, the PAR model is tested inside a pressure vessel under boundary conditions similar to a real NPP. The experiments show, that the temperatures of the cooled catalytic sheets stay significantly below the temperature of the uncooled sheets and below the ignition temperature of the gas mixture under any set boundary conditions, although no significant reduction of the conversion efficiency can be observed. As a last point, a mathematical model of the reaction kinetics of the recombination process as well as a model of the fluid dynamic and thermohydraulic processes in a heat pipe are developed with the data obtained from the experiments.

  16. Geotechnical characterization and finite element pipe/soil interaction modeling of a pipeline installed in an actively moving, permafrost slope

    Energy Technology Data Exchange (ETDEWEB)

    Bidwell, A. [AMEC Earth and Environmental, Calgary, AB (Canada); Sen, M.; Pederson, I. [Enbridge Pipelines Inc., Edmonton, AB (Canada); Yoosef-Ghodsi, N. [C-FER Technologies, Edmonton, AB (Canada)

    2010-07-01

    This paper discussed a pipeline integrity analysis for a buried crude-oil pipeline at a site characterized by unstable permafrost slopes. Data collected from piezometers, inclinometers, and thermistor cables installed as part of a comprehensive geotechnical monitoring program were used to determine the geotechnical character of the site and model pipe/soil interactions. A finite element pipe/soil interaction model was developed to estimate the potential strain to the pipeline capacity in a worst-case scenario involving mass soil movement. The purpose was to determine the necessity of costly mitigation measures. The model showed that the pipeline strain capacity is unlikely to be exceeded in the event of a sudden ground movement at the slope. The soil, permafrost, and slope movement conditions at the site were described along with the methodology and results of the pipe/soil interaction model. The model, in which the pipeline is considered as a continuous structural beam, was used to analyze both the estimated current slope movement and the worst case large magnitude slope movement. To assess the pipeline integrity in the event of mass slope movement, the expected strain demand was compared to the strain capacity, taking into account whether the pipe is heavy wall, line pipe, or containing girth welds. The analysis indicated that the risk of pipeline failure is low in the event of a large magnitude slope movement. The pipe strain measurements were found to be within the design limits for the pipeline. The analysis is relevant to other northern pipeline and linear infrastructure developments. 8 refs., 6 figs.

  17. Insulated pipe clamp design

    International Nuclear Information System (INIS)

    Anderson, M.J.; Hyde, L.L.; Wagner, S.E.; Severud, L.K.

    1980-01-01

    Thin wall large diameter piping for breeder reactor plants can be subjected to significant thermal shocks during reactor scrams and other upset events. On the Fast Flux Test Facility, the addition of thick clamps directly on the piping was undesired because the differential metal temperatures between the pipe wall and the clamp could have significantly reduced the pipe thermal fatigue life cycle capabilities. Accordingly, an insulated pipe clamp design concept was developed. 5 refs

  18. Experimental research of heat recuperators in ventilation systems on the basis of heat pipes

    Directory of Open Access Journals (Sweden)

    Matveev Andrey

    2017-01-01

    Full Text Available The paper presents the results of experimental studies of heat pipes and their thermo-technical characteristics (heat power, conductivity, heat transfer resistance, heat-transfer coefficient, temperature level and differential, etc.. The theoretical foundations and the experimental methods of the research of ammonia heat pipes made of aluminum section АS – КRА 7.5 – R1 (made of the alloy AD - 31 are explained. The paper includes the analysis of the thermo-technical characteristics of heat pipes as promising highly efficient heat transfer devices, which may be used as the basic elements of heat exchangers - heat recuperators for exhaust ventilation air, capable of providing energy-saving technologies in ventilation systems for housing and public utilities and for various branches of industry. The thermo-technical characteristics of heat pipes (HP as the basic elements of a decentralized supply-extract ventilation system (DSEVS and energy-saving technologies are analyzed. As shown in the test report of the ammonia horizontal HP made of the section АS-КRА 7,5-R1-120, this pipe ensures safe operation under various loads.

  19. Free vibration analysis of multi-span pipe conveying fluid with dynamic stiffness method

    International Nuclear Information System (INIS)

    Li Baohui; Gao Hangshan; Zhai Hongbo; Liu Yongshou; Yue Zhufeng

    2011-01-01

    Research highlights: → The dynamic stiffness method was proposed to analysis the free vibration of multi-span pipe conveying fluid. → The main advantage of the proposed method is that it can hold a high precision even though the element size is large. → The flowing fluid can weaken the pipe stiffness, when the fluid velocity increases, the natural frequencies of pipe are decreasing. - Abstract: By taking a pipe as Timoshenko beam, in this paper the original 4-equation model of pipe conveying fluid was modified by taking the dynamic effects of fluid into account. The shape function that always used in the finite element method was replaced by the exact wave solution of the modified four equations. And then the dynamic stiffness was deduced for the free vibration of pipe conveying fluid. The proposed method was validated by comparing the results of critical velocity with analytical solution for a simply supported pipe at both ends. In the example, the proposed method was applied to calculate the first three natural frequencies of a three span pipe with twelve meters long in three different cases. The results of natural frequency for the pipe conveying stationary fluid fitted well with that calculated by finite element software Abaqus. It was shown that the dynamic stiffness method can still hold high precision even though the element's size was quite large. And this is the predominant advantage of the proposed method comparing with conventional finite element method.

  20. Heat Pipe Powered Stirling Conversion for the Demonstration Using Flattop Fission (DUFF) Test

    Science.gov (United States)

    Gibson, Marc A.; Briggs, Maxwell H.; Sanzi, James L.; Brace, Michael H.

    2013-01-01

    Design concepts for small Fission Power Systems (FPS) have shown that heat pipe cooled reactors provide a passive, redundant, and lower mass option to transfer heat from the fuel to the power conversion system, as opposed to pumped loop designs typically associated with larger FPS. Although many systems have been conceptually designed and a few making it to electrically heated testing, none have been coupled to a real nuclear reactor. A demonstration test named DUFF Demonstration Using Flattop Fission, was planned by the Los Alamos National Lab (LANL) to use an existing criticality experiment named Flattop to provide the nuclear heat source. A team from the NASA Glenn Research Center designed, built, and tested a heat pipe and power conversion system to couple to Flattop with the end goal of making electrical power. This paper will focus on the design and testing performed in preparation for the DUFF test.

  1. Investigation on method of elasto-plastic analysis for piping system (benchmark analysis)

    International Nuclear Information System (INIS)

    Kabaya, Takuro; Kojima, Nobuyuki; Arai, Masashi

    2015-01-01

    This paper provides method of an elasto-plastic analysis for practical seismic design of nuclear piping system. JSME started up the task to establish method of an elasto-plastic analysis for nuclear piping system. The benchmark analyses have been performed in the task to investigate on method of an elasto-plastic analysis. And our company has participated in the benchmark analyses. As a result, we have settled on the method which simulates the result of piping exciting test accurately. Therefore the recommended method of an elasto-plastic analysis is shown as follows; 1) An elasto-plastic analysis is composed of dynamic analysis of piping system modeled by using beam elements and static analysis of deformed elbow modeled by using shell elements. 2) Bi-linear is applied as an elasto-plastic property. Yield point is standardized yield point multiplied by 1.2 times, and second gradient is 1/100 young's modulus. Kinematic hardening is used as a hardening rule. 3) The fatigue life is evaluated on strain ranges obtained by elasto-plastic analysis, by using the rain flow method and the fatigue curve of previous studies. (author)

  2. The development of the design method of nuclear piping system supported by elasto-plastic support structures (Part 1)

    International Nuclear Information System (INIS)

    Endo, R.; Murota, M.; Kawahata, J.-I.; Sato, T.; Mekomoto, Y.; Takayama, Y.; Kobayashi, H.; Hirose, J.

    1993-01-01

    The conventional aseismic design method of nuclear piping system is very conservative because of the accumulation of various safety factors in the design process, and nuclear piping systems are thought to have a large safety margin. Considering this situation, we promoted research to further rationalize nuclear power plants by reducing the amount of support structures and reducing the piping seismic response through vibration energy absorption resulting from the elasto-plastic behavior of piping support structures. The research has the following three stages. In the first stage, we select conventional piping support structures in Japanese light-water reactors that exhibit elasto-plastic behavior, and study the displacement dependency and the vibration frequency dependency on the stiffness and the energy absorption by testing their model. In the second stage, we make a piping test model with support structures whose characteristics have already been obtained, and perform vibration tests on a shaking table. In this way, we analyze the piping vibration characteristics by sinusoidal wave sweep tests and the piping response characteristics by seismic wave vibration tests, when the support structures are in an elasto-plastic condition. In the third stage, a general method is developed to evaluate the characteristics of the support structures obtained in the tests and it is applied to the evaluation of the characteristics of general support structures. A simplified analysis method is developed to evaluate the piping seismic response using the piping model test result. To expand the results mentioned above, we are developing a seismic design method of piping systems that allows support structures to have elasto-plastic behaviour. This paper reports the results of experiments conducted under the joint research program of Japanese electric power companies with support elements in the first stage and those with piping models in the second stage

  3. Key quality aspects for a new metallic composite pipe: corrosion testing, welding, weld inspection and manufacturing

    Energy Technology Data Exchange (ETDEWEB)

    Conder, Robert J.; Felton, Peter [Xodus Group Ltd., Aberdeen (United Kingdom); Smith, Richard [Shell Global Solutions Inc., Houston, TX (United States); Burke, Raymond [Pipestream Inc., Houston, TX (United States); Dikstra, Frits; Deleye, Xavier [Applus RTD Ltd., Rotterdam (Netherlands)

    2010-07-01

    XPipeTM is a new metallic composite pipe. This paper discusses three aspects of this new technology. The first subject is determination of the probability of hydrogen embrittlement by the XPipeTM manufacturing method. Two materials were analyzed in three tests: slow strain rate test, constant load test and notched tensile test. The results showed that the high strength steels used do not appear to be susceptible to hydrogen embrittlement. The second subject of this article is weld inspection. A non-destructive testing method of girth welds is developed to allow inspection of the thin-walled austenitic liner pipe. The results demonstrated that the welds can be inspected using the creeping wave technique. The third subject is quality control systems using the SCADA system, which maintains traceability of the materials and monitors and records all parameters during the production process. This system appears to be efficient in ensuring that the product pipe meets recognized quality standards.

  4. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M Kothari; Gerard T. Pittard

    2004-07-01

    ) continued with additional in-pipe testing required to optimize the design of the robot elements and surface control electronics and software. Task 6 (Design & Build Surface Control and Monitoring System) has been completed with the control and computer display functions being operated through LabView. Task 7 (Design & Fabricate Large Diameter Live Access System) progressed to completing the detailed design of the entry fitting for 12-inch diameter cast iron pipe. The fitting is now being manufactured. The 12-inch ball valve for allowing no-blow access was also procured. Task 8 (System Integration and Laboratory Validation) continued with the development of the robot module inter-connects and of a master LabView-based system display and control software.

  5. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M. Kothari; Gerard T. Pittard

    2004-04-01

    Utilities in the U.S. operate over 75,000 km (47,000 miles) of old cast-iron pipes for gas distribution. The bell-and-spigot joints that connect pipe sections together tend to leak as these pipes age. Current repair practices are costly and highly disruptive. The objective of this program is to design, test and commercialize a robotic system capable of sealing multiple cast-iron bell and spigot joints from a single pipe entry point. The proposed system will perform repairs while the pipe remains in service by traveling through the pipe, cleaning each joint surface, and installing a stainless-steel sleeve lined with an epoxy-impregnated felt across the joint. This approach will save considerable time and labor, avoid traffic disruption, and eliminate any requirement to interrupt service to customers (which would result in enormous expense to utilities). Technical challenges include: (1) repair sleeves must compensate for diametric variation and eccentricity of cast-iron pipes; (2) the assembly must travel long distances through pipes containing debris; (3) the pipe wall must be effectively cleaned in the immediate area of the joint to assure good bonding of the sleeve; and (4) an innovative bolt-on entry fitting is required to conduct repair operations on live mains. The development effort is divided into eleven tasks. Task 1--Program Management and Task 2--were completed in prior quarters while Task 3--Design and Fabricate Ratcheting Stainless-Steel Repair Sleeves has progressed to installing prototype sleeves in cast iron test pipe segments. Efforts in this quarter continued to focus on Tasks 4--8, with significant progress made in each. Task 4 (Design, Fabricate and Test Patch Setting Robotic Train) progressed to the design of the control electronics and pneumatic system to inflate the bladder robotic patch setting module. Task 5 (Design & Fabricate Pipe-Wall Cleaning Robot Train with Pan/Zoom/Tilt Camera) continued with additional in-pipe testing required to

  6. Identification and reduction of piping-vibrations in plants

    International Nuclear Information System (INIS)

    Kerkhof, K.

    2012-01-01

    Safe operation, availability and lifetime assessment of piping systems are of utmost concern for plant operators. The use of tuned mass dampers is a rather new approach for reducing vibrations to avoid high cycle fatigue in a large chemical piping system. The investigated piping system is supported by a tall structure fixed at the base. As a result, the steel building stiffness decreases with height. Furthermore large piping-elbow forces act at the top of the building, which lead to large vibration amplitudes. Since both piping system and supporting structure exhibited these large vibration amplitudes, dampers or shock absorbers placed between them would prove ineffective. Therefore, special vibration absorbers were developed for such piping systems. The paper presents the design process, starting with an extensive system investigation up to the passive multi-axial vibration absorber design parameters. This includes: Laboratory tests with a mock-up pipe system, where the first design ideas for new passive vibration absorbers were investigated. Vibration measurements were carried out to investigate the current state of the vibration behaviour. The piping system was inspected; strain gauges were used to identify stress concentrations at welds and other notches due to ovalization. Finite element calculations were performed, first as a combined beam and shell model for the pipe without the support structure. A detailed model for the combined steel construction and pipe system was created. Model-updating was done to fit the calculated model to the experimental modal analysis data. Loading assumptions describing excitation forces from the mass flow were checked. Harmonic frequency analysis was performed. On the basis of these calculations design parameters for the passive vibration absorber were determined. Finally, a solution for the design of two passive vibration absorbers will be presented.

  7. Identification and reduction of piping-vibrations in plants

    Energy Technology Data Exchange (ETDEWEB)

    Kerkhof, K. [Stuttgart Univ. (Germany). MPA

    2012-07-01

    Safe operation, availability and lifetime assessment of piping systems are of utmost concern for plant operators. The use of tuned mass dampers is a rather new approach for reducing vibrations to avoid high cycle fatigue in a large chemical piping system. The investigated piping system is supported by a tall structure fixed at the base. As a result, the steel building stiffness decreases with height. Furthermore large piping-elbow forces act at the top of the building, which lead to large vibration amplitudes. Since both piping system and supporting structure exhibited these large vibration amplitudes, dampers or shock absorbers placed between them would prove ineffective. Therefore, special vibration absorbers were developed for such piping systems. The paper presents the design process, starting with an extensive system investigation up to the passive multi-axial vibration absorber design parameters. This includes: Laboratory tests with a mock-up pipe system, where the first design ideas for new passive vibration absorbers were investigated. Vibration measurements were carried out to investigate the current state of the vibration behaviour. The piping system was inspected; strain gauges were used to identify stress concentrations at welds and other notches due to ovalization. Finite element calculations were performed, first as a combined beam and shell model for the pipe without the support structure. A detailed model for the combined steel construction and pipe system was created. Model-updating was done to fit the calculated model to the experimental modal analysis data. Loading assumptions describing excitation forces from the mass flow were checked. Harmonic frequency analysis was performed. On the basis of these calculations design parameters for the passive vibration absorber were determined. Finally, a solution for the design of two passive vibration absorbers will be presented.

  8. Restart Testing Program for piping following steam generator replacement at North Anna Unit 1

    International Nuclear Information System (INIS)

    Bain, R.A.; Bayer, R.K.

    1993-01-01

    In order to provide assurance that the effects of performing steam generator replacement (SGR) at North Anna unit 1 had no adverse impact on plant piping systems, a cold functional verification restart testing program was developed. This restart testing program was implemented in lieu of a hot functional testing program normally used during the initial startup of a nuclear plant. A review of North Anna plant-specific and generic U.S. Nuclear Regulatory Commission requirements for restart testing was performed to ensure that no mandatory hot functional testing was required. This was determined to be the case, and the development of a cold functional test program was initiated. The cold functional test had inherent advantages as compared to the hot functional testing, while still providing assurance of piping system adequacy. The advantages of the cold verification program included reducing risk to personnel from hot piping, increasing the accuracy of measurements with the improvement in work conditions, eliminating engineering activities during the heatup process, and being able to record measurements as construction work was completed allowing for rework or repair of components if required. To ensure the effectiveness of the cold verification program, a project procedure was generated to identify the personnel, equipment, and measurement requirements. An engineering calculation was issued to document the scope of the restart test program, and an additional calculation was developed to provide acceptance criteria for the critical commodity measurements

  9. Numerical evaluation of cracked pipes under dynamic loading

    International Nuclear Information System (INIS)

    Petit, M.; Jamet, P.

    1989-01-01

    In order to apply the leak-before-break concept to piping systems, the behavior of cracked pipes under dynamic, and especially seismic, loadings must be studied. A simple finite element model of a cracked pipe has been developed and implemented in the general purpose computer code CASTEM 2000. The model is a generalization of the approach proposed by Paris and Tada (1). Considered loads are bending moment and axial force (representing thermal expansion and internal pressure.) The elastic characteristics of the model are determined using the Zahoor formulae for the geometry-dependent factors. Owing to the material behabior plasticity must be taken into account. To represent the crack growth, the material is defined by two characteristic values: J 1c which is the level of energy corresponding to crack initiation and the tearing modulus, T, which governs the length of propagation of the crack. For dynamic loads, unilateral conditions are imposed to represent crack closure. The model has been used for the design of dynamic tests to be conducted on shaking tables. Test principle is briefly described and numerical results are presented. Finally evaluation of margin, due to plasticity, in comparison with the standard design procedure is made

  10. Eddy Current Testing with Giant Magnetoresistance (GMR) Sensors and a Pipe-Encircling Excitation for Evaluation of Corrosion under Insulation.

    Science.gov (United States)

    Bailey, Joseph; Long, Nicholas; Hunze, Arvid

    2017-09-28

    This work investigates an eddy current-based non-destructive testing (NDT) method to characterize corrosion of pipes under thermal insulation, one of the leading failure mechanisms for insulated pipe infrastructure. Artificial defects were machined into the pipe surface to simulate the effect of corrosion wall loss. We show that by using a giant magnetoresistance (GMR) sensor array and a high current (300 A), single sinusoidal low frequency (5-200 Hz) pipe-encircling excitation scheme it is possible to quantify wall loss defects without removing the insulation or weather shield. An analysis of the magnetic field distribution and induced currents was undertaken using the finite element method (FEM) and analytical calculations. Simple algorithms to remove spurious measured field variations not associated with defects were developed and applied. The influence of an aluminium weather shield with discontinuities and dents was ascertained and found to be small for excitation frequency values below 40 Hz. The signal dependence on the defect dimensions was analysed in detail. The excitation frequency at which the maximum field amplitude change occurred increased linearly with the depth of the defect by about 3 Hz/mm defect depth. The change in magnetic field amplitude due to defects for sensors aligned in the azimuthal and radial directions were measured and found to be linearly dependent on the defect volume between 4400-30,800 mm³ with 1.2 × 10 -3 -1.6 × 10 -3 µT/mm³. The results show that our approach is well suited for measuring wall loss defects similar to the defects from corrosion under insulation.

  11. Development of bore tools for pipe inspection

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Nakahira, Masataka; Taguchi, Kou; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Reactor (ITER), replacement and maintenance on in-vessel components requires that all cooling pipes connected be cut and removed, that a new component be installed, and that all cooling pipes be rewelded. After welding is completed, welded area must be inspected for soundness. These tasks require a new work concept for securing shielded area and access from narrow ports. Tools had to be developed for nondestructive inspection and leak testing to evaluate pipe welding soundness by accessing areas from inside pipes using autonomous locomotion welding and cutting tools. A system was proposed for nondestructive inspection of branch pipes and the main pipe after passing through pipe curves, the same as for welding and cutting tool development. Nondestructive inspection and leak testing sensors were developed and the basic parameters were obtained. In addition, the inspection systems which can move inside pipes and conduct the nondestructive inspection and the leak testing were developed. In this paper, an introduction will be given to the current situation concerning the development of nondestructive inspection and leak testing machines for the branch pipes. (author)

  12. Analysis of nuclear piping system seismic tests with conventional and energy absorbing supports

    International Nuclear Information System (INIS)

    Park, Y.; DeGrassi, G.; Hofmayer, C.; Bezler, P.; Chokshi, N.

    1997-01-01

    Large-scale models of main steam and feedwater piping systems were tested on the shaking table by the Nuclear Power Engineering Cooperation (NUPEC) of Japan, as part of the Seismic Proving Test Program. This paper describes the linear and nonlinear analyses performed by NRC/BNL and compares the results to the test data

  13. Piping vibrations measured during FFTF startup

    International Nuclear Information System (INIS)

    Anderson, M.J.

    1981-03-01

    An extensive vibration survey was conducted on the Fast Flux Test Facility piping during the plant acceptance test program. The purpose was to verify that both mechanical and flow induced vibration amplitudes were of sufficiently low level so that pipe and pipe support integrity would not be compromised over the plant design lifetime. Excitation sources included main heat transport sodium pumps, reciprocating auxiliary system pumps, EM pumps, and flow oscillations. Pipe sizes varied from one-inch to twenty-eight-inches in diameter. This paper describes the test plan; the instrumentation and procedures utilized; and the test results

  14. Development of a Skewed Pipe Shear Connector for Precast Concrete Structures.

    Science.gov (United States)

    Kim, Sang-Hyo; Choi, Jae-Gu; Park, Sejun; Lee, Hyunmin; Heo, And Won-Ho

    2017-05-13

    Joint connection methods, such as shear key and loop bar, improve the structural performance of precast concrete structures; consequently, there is usually decreased workability or constructional efficiency. This paper proposes a high-efficiency skewed pipe shear connector. To resist shear and pull-out forces, the proposed connectors are placed diagonally between precast concrete segments and a cast-in-place concrete joint part on a girder. Design variables (such as the pipe diameter, length, and insertion angle) have been examined to investigate the connection performance of the proposed connector. The results of our testing indicate that the skewed pipe shear connectors have 50% higher ductility and a 15% higher ratio of maximum load to yield strength as compared to the corresponding parameters of the loop bar. Finite element analysis was used for validation. The resulting validation indicates that, compared to the loop bar, the skewed pipe shear connector has a higher ultimate shear and pull-out resistance. These results indicate that the skewed pipe shear connector demonstrates more idealized behavior than the loop bar in precast concrete structures.

  15. Stability of cracked pipe under inertial stresses. Subtask 1.1 final report

    International Nuclear Information System (INIS)

    Scott, P.; Wilson, M.; Olson, R.; Marschall, C.; Schmidt, R.; Wilkowski, G.

    1994-08-01

    This report presents the results of the pipe fracture experiments, analyses, and material characterization efforts performed within Subtask 1.1 of the IPIRG Program. The objective of Subtask 1.1 was to experimentally verify the analysis methodologies for circumferentially cracked pipe subjected primarily to inertial stresses. Eight cracked-pipe experiments were conducted on 6-inch nominal diameter TP304 and A106B pipe. The experimental procedure was developed using nonlinear time-history finite element analyses which included the nonlinear behavior due to the crack. The model did an excellent job of predicting the displacements, forces, and times to maximum moment. The comparison of the experimental loads to the predicted loads by the Net-Section-Collapse (NSC), Dimensionless Plastic-Zone Parameter, J-estimation schemes, R6, and ASME Section XI in-service flaw assessment criteria tended to underpredict the measured bending moments except for the NSC analysis of the A106B pipe. The effects of flaw geometry and loading history on toughness were evaluated by calculating the toughness from the pipe tests and comparing these results to C(l) values. These effects were found to be variable. The surface-crack geometry tended to increase the toughness (relative to CM results), whereas a negative load-ratio significantly decreased the TP304 stainless steel surface-cracked pipe apparent toughness. The inertial experiments tended to achieve complete failure within a few cycles after reaching maximum load in these relatively small diameter pipe experiments. Hence, a load-controlled fracture mechanics analysis may be more appropriate than a displacement-controlled analysis for these tests

  16. Vibration analysis for IHTS piping system of LMR conveying hot liquid sodium

    International Nuclear Information System (INIS)

    Koo, Gyeong Hoi; Lee, Hyeong Yeon; Lee, Jae Han

    2001-01-01

    In this paper, the vibration characteristics of IHTS(Intermediate Heat Transfer System) piping system of LMR(Liquid Metal Reactor) conveying hot liquid sodium are investigated to eliminate the pipe supports for economic reasons. To do this, a 3-dimensional straight pipe element and a curved pipe element conveying fluid are formulated using the dynamic stiffness method of the wave approach and coded to be applied to any complex piping system. Using this method, the dynamic characteristics including the natural frequency, the frequency response functions, and the dynamic instability due to the pipe internal flow velocity are analyzed. As one of the design parameters, the vibration energy flow is also analyzed to investigate the disturbance transmission paths for the resonant excitation and the non-resonant excitations

  17. Sensitivity Analysis on Elbow Piping Components in Seismically Isolated NPP under Seismic Loading

    Energy Technology Data Exchange (ETDEWEB)

    Ju, Hee Kun; Hahm, Dae Gi; Kim, Min Kyu [KAERI, Daejeon (Korea, Republic of); Jeon, Bub Gyu; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of)

    2016-05-15

    In this study, the FE model is verified using specimen test results and simulation with parameter variations are conducted. Effective parameters will randomly sampled and used as input values for simulations to be applied to the fragility analysis. pipelines are representative of them because they could undergo larger displacements when they are supported on both isolated and non-isolated structures simultaneously. Especially elbows are critical components of pipes under severed loading conditions such as earthquake action because strain is accumulated on them during the repeated bending of the pipe. Therefore, seismic performance of pipe elbow components should be examined thoroughly based on the fragility analysis. Fragility assessment of interface pipe should take different sources of uncertainty into account. However, selection of important sources and repeated tests with many random input values are very time consuming and expensive, so numerical analysis is commonly used. In the present study, finite element (FE) model of elbow component will be validated using the dynamic test results of elbow components. Using the verified model, sensitivity analysis will be implemented as a preliminary process of seismic fragility of piping system. Several important input parameters are selected and how the uncertainty of them are apportioned to the uncertainty of the elbow response is to be studied. Piping elbows are critical components under cyclic loading conditions as they are subjected large displacement. In a seismically isolated NPP, seismic capacity of piping system should be evaluated with caution. Seismic fragility assessment preliminarily needs parameter sensitivity analysis about the output of interest with different input parameter values.

  18. Development of bore tools for pipe welding and cutting

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji

    1998-01-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

  19. Development of bore tools for pipe welding and cutting

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Kiyoshi; Ito, Akira; Takiguchi, Yuji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    In the International Thermonuclear Experimental Reactor (ITER), in-vessel components replacement and maintenance requires that connected cooling pipes be cut and removed beforehand and that new components be installed to which cooling pipes must be rewelded. All welding must be inspected for soundness after completion. These tasks require a new task concept for ensuring shielded areas and access from narrow ports. Thus, it became necessary to develop autonomous locomotion welding and cutting tools for branch and main pipes to weld pipes by in-pipe access; a system was proposed that cut and welded branch and main pipes after passing inside pipe curves, and elemental technologies developed. This paper introduces current development in tools for welding and cutting branch pipes and other tools for welding and cutting the main pipe. (author)

  20. The modal analysis of a pipe elbow with realistic boundary conditions

    International Nuclear Information System (INIS)

    Carneiro, J.O.; Melo, F.J.Q. de; Rodrigues, J.F.D.; Lopes, H.; Teixeira, V.

    2005-01-01

    A vibration analysis for the determination of the natural frequencies and the associated eigenmodes of a pipe elbow with end-flanges or tangent terminations was performed. A numerical investigation of this problem was achieved with a semi-analytic definition finite ring element and a commercial finite element code. To assess the accuracy of the numerical solution for the elbow vibration, an experimental modal analysis was performed on a curved and on a straight pipe. The responses were processed by a data acquisition system which performs a fast Fourier transform on the time histories to convert them from a time to frequency domain, these leading to the extraction of natural frequencies and mode shapes associated with the test-specimen. The results were compared with the corresponding ones from the numerical approach and discussion about the results completes the paper

  1. Heat exchanger nozzle stresses due to pipe vibration

    International Nuclear Information System (INIS)

    Wolgemuth, G.A.

    1983-01-01

    A large diameter pipe in a heavy water production plant was excited into a low frequency vibration due to void collapse of the pipe contents at a sharp vertical drop in the pipe run. Fears that this vibration would fatigue the inlet nozzle to the heat exchanger prompted the introduction of a flow of cold water into the pipe to prevent the two-phase flow from developing but at the cost of reduced heat exchanger efficiency. An investigation was carried out to determine the stress levels in the nozzle with the quenching flow off and suggest means of reducing them if excessive. A finite element dynamic simulation of the pipe run was performed to determine the likely mode shapes. This information was used to optimize the placement of velocity probes on the pipe. Field measurements of vibration were taken for several operating conditions. This data was analyzed and the results used to refine the support stiffness used in the finite element simulation. The finite element model was then used to predict the nozzle forces and moments. In turn this data was used to determine the local stresses in the nozzle. The ASME Section III code was used to determine the allowable fully reversing stresses for the unit in question. It was found that the endurance limit of 83 MPa was exceeded in the analysis only when using the most conservative estimates for each uncertainty. It was recommended that if the safety factor was not deemed high enough, the nozzle should be built up with a reinforcing pad no thicker than 12 mm

  2. On the behavior of pressurized pipings under excessive-stresses caused by earthquake loadings

    International Nuclear Information System (INIS)

    Udoguchi, Y.; Akino, K.; Shibata, H.

    1975-01-01

    Five types of breaking experiments on pipe elements and piping structures had been carried out from 1971 to 1973 by the technical sub-committee of the Japan Electric Association under the leadership taken by Y. Udoguchi, one of the authors. One of the fruitful results was to realize the guillotine-type rupture of pipe element on a shaking table. However, it was also shown that the margin for the design is enough, and allowable stresses under earthquake loading are obtained by modifying those of the Emergency Condition of the ASME Code. The experiments effected were as follows: straight pipe elements, curved pipes and T-branch pipe connections, made of both ferritic and austenitic steels, were subjected to repeated bending moment, torsional moment and combined under pressurized condition. The pressure corresponded to their design value, but the stresses caused by such moments exceeded over their allowable stress of the Faulted Condition of the ASME Code. The wave patterns were both sinusoidal and natural earthquake records

  3. Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow

    International Nuclear Information System (INIS)

    Kuzay, T.M.; Kasza, K.E.

    1985-01-01

    A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90 0 sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions

  4. Proposal of failure criterion applicable to finite element analysis results for wall-thinned pipes under bending load

    Energy Technology Data Exchange (ETDEWEB)

    Meshii, Toshiyuki, E-mail: meshii@u-fukui.ac.jp [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui (Japan); Ito, Yoshiaki [Graduate School of Engineering, University of Fukui, 3-9-1 Bunkyo, Fukui (Japan)

    2012-01-15

    Highlights: Black-Right-Pointing-Pointer Limit bending load (LBL) of wall-thinned pipe by large strain FEA was considered. Black-Right-Pointing-Pointer Net section yield load had sufficient margin to LBL. Black-Right-Pointing-Pointer LBL for collapse was the load when volume with nominal thickness yielded. Black-Right-Pointing-Pointer LBL for cracking was the load when flawed section stress exceeded tensile strength. Black-Right-Pointing-Pointer Failure criterion considering above was named Domain Collapse Criterion. - Abstract: In this work, a failure criterion applicable to large strain Finite Element Analysis (FEA) results was proposed in order to predict both the fracture mode (collapse or cracking) and the limit bending load of wall-thinned straight pipes. This work was motivated from the recent experimental results of ; that is, fracture mode is not always collapse, and the fracture mode affects the limit bending load. The key finding in comparing their test results and a detailed large strain FEA results was that the Mises stress distribution at the limit bending load of a flawed cylinder was similar to that of a flawless cylinder; specifically, in case of collapse, the Mises stress exceeded the true yield stress of a material for the whole 'volume' of a cylinder with a nominal wall thickness. Based on this finding, a failure criterion applicable to large strain FEA results of wall-thinned straight pipes under a bending load that can predict both fracture mode and limit bending load was proposed and was named the Domain Collapse Criterion (DCC). DCC predicts the limit bending load as the lower value of either the M{sub c}{sup FEA}, which is the load at which the Mises stress exceeds the true yield strength of a straight pipe for the whole 'volume' with a nominal wall thickness (fracture mode: collapse), or the M{sub c}{sup FEAb}, which is the load at which the Mises stress in a section of the flaw ligament exceeds the true tensile stress

  5. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1994-01-01

    It is approximately 10 years since the Third Edition of Heat Pipes was published and the text is now established as the standard work on the subject. This new edition has been extensively updated, with revisions to most chapters. The introduction of new working fluids and extended life test data have been taken into account in chapter 3. A number of new types of heat pipes have become popular, and others have proved less effective. This is reflected in the contents of chapter 5. Heat pipes are employed in a wide range of applications, including electronics cooling, diecasting and injection mo

  6. An integrated heat pipe-thermal storage design for a solar receiver

    Science.gov (United States)

    Keddy, E.; Sena, J. T.; Woloshun, K.; Merrigan, M. A.; Heidenreich, G.

    Light-weight heat pipe wall elements that incorporate a thermal storage subassembly within the vapor space are being developed as part of the Organic Rankine Cycle Solar Dynamic Power System (ORC-SDPS) receiver for the Space Station application. The operating temperature of the heat pipe elements is in the 770 to 810 K range with a design power throughput of 4.8 kW per pipe. The total heat pipe length is 1.9 M. The Rankine cycle boiler heat transfer surfaces are positioned within the heat pipe vapor space, providing a relatively constant temperature input to the vaporizer. The heat pipe design employs axial arteries and distribution wicked thermal storage units with potassium as the working fluid. Performance predictions for this configuration have been conducted and the design characterized as a function of artery geometry, distribution wick thickness, porosity, pore size, and permeability.

  7. Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report

    International Nuclear Information System (INIS)

    Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

    1982-04-01

    This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

  8. Pipe Overpack Container Fire Testing: Phase I II & III.

    Energy Technology Data Exchange (ETDEWEB)

    Figueroa, Victor G. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ammerman, Douglas J. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Lopez, Carlos [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gill, Walter [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2018-02-01

    The Pipe Overpack Container (POC) was developed at Rocky Flats to transport plutonium residues with higher levels of plutonium than standard transuranic (TRU) waste to the Waste Isolation Pilot Plant (WIPP) for disposal. In 1996 Sandia National Laboratories (SNL) conducted a series of tests to determine the degree of protection POCs provided during storage accident events. One of these tests exposed four of the POCs to a 30-minute engulfing pool fire, resulting in one of the 7A drum overpacks generating sufficient internal pressure to pop off its lid and expose the top of the pipe container (PC) to the fire environment. The initial contents of the POCs were inert materials, which would not generate large internal pressure within the PC if heated. POCs are now being used to store combustible TRU waste at Department of Energy (DOE) sites. At the request of DOE’s Office of Environmental Management (EM) and National Nuclear Security Administration (NNSA), starting in 2015 SNL conducted a series of fire tests to examine whether PCs with combustibles would reach a temperature that would result in (1) decomposition of inner contents and (2) subsequent generation of sufficient gas to cause the PC to over-pressurize and release its inner content. Tests conducted during 2015 and 2016 were done in three phases. The goal of the first phase was to see if the PC would reach high enough temperatures to decompose typical combustible materials inside the PC. The goal of the second test phase was to determine under what heating loads (i.e., incident heat fluxes) the 7A drum lid pops off from the POC drum. The goal of the third phase was to see if surrogate aerosol gets released from the PC when the drum lid is off. This report will describe the various tests conducted in phase I, II, and III, present preliminary results from these tests, and discuss implications for the POCs.

  9. Test and evaluation about damping characteristics of hanger supports for nuclear power plant piping systems (Seismic Damping Ratio Evaluation Program)

    International Nuclear Information System (INIS)

    Shibata, H.; Ito, A.; Tanaka, K.; Niino, T.; Gotoh, N.

    1981-01-01

    Generally, damping phenomena of structures and equipments is caused by very complex energy dissipation. Especially, as piping systems are composed of many components, it is very difficult to evaluate damping characteristics of its system theoretically. On the other hand, the damping value for aseismic design of nuclear power plants is very important design factor to decide seismic response loads of structures, equipments and piping systems. The very extensive studies titled SDREP (Seismic Damping Ratio Evaluation Program) were performed to establish proper damping values for seismic design of piping as a joint work among a university, electric companies and plant makers. In SDREP, various systematic vibration tests were conducted to investigate factors which may contribute to damping characteristics of piping systems and to supplement the data of the pre-operating tests. This study is related to the component damping characteristics tests of that program. The object of this study is to clarify damping characteristics and mechanism of hanger supports used in piping systems, and to establish the evaluation technique of dispersing energy at hanger support points and its effect to the total damping ability of piping system. (orig./WL)

  10. Nonlinear dynamic analysis of piping systems using the pseudo force method

    International Nuclear Information System (INIS)

    Prachuktam, S.; Bezler, P.; Hartzman, M.

    1979-01-01

    Simple piping systems are composed of linear elastic elements and can be analyzed using conventional linear methods. The introduction of constraint springs separated from the pipe with clearance gaps to such systems to cope with the pipe whip or other extreme excitation conditions introduces nonlinearities to the system, the nonlinearities being associated with the gaps. Since these spring-damper constraints are usually limited in number, descretely located, and produce only weak nonlinearities, the analysis of linear systems including these nonlinearities can be carried out by using modified linear methods. In particular, the application of pseudo force methods wherein the nonlinearities are treated as displacement dependent forcing functions acting on the linear system were investigated. The nonlinearities induced by the constraints are taken into account as generalized pseudo forces on the right-hand side of the governing dynamic equilibrium equations. Then an existing linear elastic finite element piping code, EPIPE, was modified to permit application of the procedure. This option was inserted such that the analyses could be performed using either the direct integration method or via a modal superposition method, the Newmark-Beta integration procedure being employed in both methods. The modified code was proof tested against several problems taken from the literature or developed with the nonlinear dynamics code OSCIL. The problems included a simple pipe loop, cantilever beam, and lumped mass system subjected to pulsed and periodic forcing functions. The problems were selected to gage the overall accuracy of the method and to insure that it properly predicted the jump phenomena associated with nonlinear systems. (orig.)

  11. Residual stress improvement for pipe weld by means of induction heating pre-flawed pipe

    International Nuclear Information System (INIS)

    Umemoto, T.; Yoshida, K.; Okamoto, A.

    1980-01-01

    The intergranular stress corrosion cracking (IGSCC) has been found in type 304 stainless steel piping of several BWR plants. It is already well known that IGSCC is most likely to occur when three essential factors, material sensitization, high tensile stress and corrosive environment, are present. If the welding residual stress is sufficiently high (200 to approximately 400 MPa) in the inside piping surface near the welded joint, then it may be one of the biggest contributors to IGSCC. If the residual stress is reduced or reversed by some way, the IGSCC will be effectively mitigated. In this paper a method to improve the residual stress named IHSI (Induction Heating Stress Improvement) is explained. IHSI aims to improve the condition of residual stress in the inside pipe surface using the thermal stress induced by the temperature difference in pipe wall, that is produced when the pipe is heated from the outside surface by an induction heating coil and cooled on the inside surface by water simultaneously. This method becomes more attractive when it can be successfully applied to in-service piping which might have some pre-flaw. In order to verify the validity of IHSI for such piping, some experiments and calculations using finite element method were conducted. These results are mainly discussed in this paper from the view-points of residual stress, flaw behaviour during IHSI and material deterioration. (author)

  12. Remote controlled in-pipe manipulators for milling, welding and EC-testing, for application in BWRS

    International Nuclear Information System (INIS)

    Seeberger, E.K.

    2000-01-01

    Many pipes in power plants and industrial facilities have piping sections, which are not accessible from the outside or which are difficult to access. Accordingly, remote controlled pipe machining manipulators have been built which enable in-pipe inspection and repair. Since the 1980s, defects have been found at the Inconel welds of the RPV nozzles of boiling water reactors throughout the world. These defects comprise cracks caused by stress corrosion cracking in areas of manual welds made using the weld filler metal Inconel 182. The cracks were found in Inconel-182 buttering at the ferritic nozzles as well as in the welded joints connecting to the fully-austenitic safe ends (Inconel 600 and stainless steel). These welds are not accessible from outside. The ferritic nozzle is cladded with austenitic material on the inside. The adjacent buttering was applied manually using the weld filler metal Inconel 182. The safe end made of Inconel 600 was welded to the nozzle also using Inconel 182 as the filler metal. The repair problems for inside were solved with remote-controlled in-pipe manipulators which enable in-pipe inspection and repair. A complete systems of manipulators has been developed and qualified for application in nuclear power plants. The tasks that must be performed with this set of in-pipe manipulator are as follows: 1st step - Insertion of the milling/ET manipulator into piping to the work location; 2nd step Detection of the transition line with the ferritic measurement probe; 3rd step - Performance of a surface crack examination by eddy current (ET) method; 4th step - Milling of the groove and preparation for weld backlay and, in case of ET indications, elimination of such flaws also by milling. 5th step - Welding of backlay and/or repair weld using the GTA pulsed arc technique; 6th step - After welding it is necessary to prepare the surface for eddy current testing. A final milling inside the pipe is done with the milling manipulator to adjust the

  13. Pipe elbow stiffness coefficients including shear and bend flexibility factors for use in direct stiffness codes

    International Nuclear Information System (INIS)

    Perry, R.F.

    1977-01-01

    Historically, developments of computer codes used for piping analysis were based upon the flexibility method of structural analysis. Because of the specialized techniques employed in this method, the codes handled systems composed of only piping elements. Over the past ten years, the direct stiffness method has gained great popularity because of its systematic solution procedure regardless of the type of structural elements composing the system. A great advantage is realized with a direct stiffness code that combines piping elements along with other structural elements such as beams, plates, and shells, in a single model. One common problem, however, has been the lack of an accurate pipe elbow element that would adequately represent the effects of transverse shear and bend flexibility factors. The purpose of the present paper is to present a systematic derivation of the required 12x12 stiffness matrix and load vectors for a three dimensional pipe elbow element which includes the effects of transverse shear and pipe bend flexibility according to the ASME Boiler and Pressure Vessel Code, Section III. The results are presented analytically and as FORTRAN subroutines to be directly incorporated into existing direct stiffness codes. (Auth.)

  14. 46 CFR 61.15-5 - Steam piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam piping. 61.15-5 Section 61.15-5 Shipping COAST... Periodic Tests of Piping Systems § 61.15-5 Steam piping. (a) Main steam piping shall be subjected to a... removed and the piping thoroughly examined. (b) All steam piping subject to pressure from the main boiler...

  15. Pipe inspection using the pipe crawler. Innovative technology summary report

    International Nuclear Information System (INIS)

    1999-05-01

    The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned

  16. Pipe inspection using the pipe crawler. Innovative technology summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-05-01

    The US Department of Energy (DOE) continually seeks safer and more cost-effective remediation technologies for use in the decontamination and decommissioning (D and D) of nuclear facilities. In several of the buildings at the Fernald Site, there is piping that was used to transport process materials. As the demolition of these buildings occur, disposal of this piping has become a costly issue. Currently, all process piping is cut into ten-foot or less sections, the ends of the piping are wrapped and taped to prevent the release of any potential contaminants into the air, and the piping is placed in roll off boxes for eventual repackaging and shipment to the Nevada Test Site (NTS) for disposal. Alternatives that allow for the onsite disposal of process piping are greatly desired due to the potential for dramatic savings in current offsite disposal costs. No means is currently employed to allow for the adequate inspection of the interior of piping, and consequently, process piping has been assumed to be internally contaminated and thus routinely disposed of at NTS. The BTX-II system incorporates a high-resolution micro color camera with lightheads, cabling, a monitor, and a video recorder. The complete probe is capable of inspecting pipes with an internal diameter (ID) as small as 1.4 inches. By using readily interchangeable lightheads, the same system is capable of inspecting piping up to 24 inches in ID. The original development of the BTX system was for inspection of boiler tubes and small diameter pipes for build-up, pitting, and corrosion. However, the system is well suited for inspecting the interior of most types of piping and other small, confined areas. The report describes the technology, its performance, uses, cost, regulatory and policy issues, and lessons learned.

  17. Erosion corrosion in power plant piping systems - Calculation code for predicting wall thinning

    International Nuclear Information System (INIS)

    Kastner, W.; Erve, M.; Henzel, N.; Stellwag, B.

    1990-01-01

    Extensive experimental and theoretical investigations have been performed to develop a calculation code for wall thinning due to erosion corrosion in power plant piping systems. The so-called WATHEC code can be applied to single-phase water flow as well as to two-phase water/steam flow. Only input data which are available to the operator of the plant are taken into consideration. Together with a continuously updated erosion corrosion data base the calculation code forms one element of a weak point analysis for power plant piping systems which can be applied to minimize material loss due to erosion corrosion, reduce non-destructive testing and curtail monitoring programs for piping systems, recommend life-extending measures. (author). 12 refs, 17 figs

  18. Flow induced vibrations of piping

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    In order to design the supports of piping systems, estimations of the vibrations induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary to calculate the model parameters of liquid containing pipes. In most computer codes, fluid effects are accounted for just by adding the fuid mass to the structure. This may lead to serious errors.- Inertial effects from the fluid are not correctly evaluated especially in the case of bended or of non-uniform section pipes. Fluid boundary conditions are simply ignored. - In many practical problems fluid compressibility cannot be negelcted, even in the low frequencies domain which corresponds to efficient excitation by turbulent sources of the flow. This paper presents a method to take into account these efects, by solving a coupled mechanical acoustical problem: the computer code TEDEL of the C.E.A./D.E.M.T. System, based on the finite-elements method, has been extended to calculate simultaneously the pressure fluctuations in the fluid and the vibrations of the pipe. (Auth.)

  19. Resolution of thermal striping issue downstream of a horizontal pipe elbow in stratified pipe flow. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Kuzay, T.M.; Kasza, K.E.

    1985-01-01

    A thermally stratified pipe flow produced by a thermal transient when passing through a horizontal elbow as a result of secondary flow gives rise to large thermal fluctuations on the inner curvature wall of the downstream piping. These fluctuations were measured in a specially instrumented horizontal pipe and elbow system on a test set-up using water in the Mixing Components Technology Facility (MCTF) at Argonne National Laboratory (ANL). This study is part of a larger program which is studying the influence of thermal buoyancy on general reactor component performance. This paper discusses the influence of pipe flow generated thermal oscillations on the thermal stresses induced in the pipe walls. The instrumentation was concentrated around the exit plane of the 90/sup 0/ sweep elbow, since prior tests had indicated that the largest thermal fluctuations would occur within about one hydraulic diameter downstream of the elbow exit. The thermocouples were located along the inner curvature of the piping and measured the near surface fluid temperature. The test matrix involved thermal downramps under turbulent flow conditions.

  20. Force-deflection analysis of offset indentations on pressurised pipes

    International Nuclear Information System (INIS)

    Hyde, T.H.; Luo, R.; Becker, A.A.

    2007-01-01

    The indenter force vs. deflection characteristics of pressurised pipes with long offset indentations under plane strain conditions have been investigated using finite element (FE) and analytical methods with four experimental tests performed on aluminium rings. Two different materials and five different geometries were used to investigate their effects on the elastic-plastic behaviour. A comparison of the experimental, FE and the analytical results indicates that the analytical formulation developed in this paper, for predicting the force-deflection curves for pressurised pipes with offset indenters, is reasonably accurate. Also, all of the analyses presented in this paper indicate that by using a representative flow stress, which is defined as the average of the yield and ultimate tensile stresses, the analytical method can accurately predict the force-deflection curves

  1. Experimental benchmark for piping system dynamic-response analyses

    International Nuclear Information System (INIS)

    1981-01-01

    This paper describes the scope and status of a piping system dynamics test program. A 0.20 m(8 in.) nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Particular attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed

  2. Experimental benchmark for piping system dynamic response analyses

    International Nuclear Information System (INIS)

    Schott, G.A.; Mallett, R.H.

    1981-01-01

    The scope and status of a piping system dynamics test program are described. A 0.20-m nominal diameter test piping specimen is designed to be representative of main heat transport system piping of LMFBR plants. Attention is given to representing piping restraints. Applied loadings consider component-induced vibration as well as seismic excitation. The principal objective of the program is to provide a benchmark for verification of piping design methods by correlation of predicted and measured responses. Pre-test analysis results and correlation methods are discussed. 3 refs

  3. Development of new damping devices for piping

    International Nuclear Information System (INIS)

    Kobayashi, Hiroe

    1991-01-01

    An increase of the damping ratio is known to be very effective for the seismic design of a piping system. Increasing the damping ratio and reducing the seismic response of the piping system, the following three types of damping devices for piping systems are introduced: (1) visco-elastic damper, (2) elasto-plastic damper and (3) compact dynamic damper. The dynamic characteristics of these damping devices were investigated by the component test and the applicability of them to the piping system was confirmed by the vibration test using a three dimensional piping model. These damping devices are more effective than mechanical snubbers to reduce the vibration of the piping system. (author)

  4. Butt-welding technology for double walled Polyethylene pipe

    International Nuclear Information System (INIS)

    Lee, Bo-Young; Kim, Jae-Seong; Lee, Sang-Yul; Kim, Yeong K.

    2012-01-01

    Highlights: ► We developed a butt welding apparatus for doubled walled Polyethylene pipe. ► We design the welding process by analyzing thermal behaviors of the material. ► We performed the welding and tested the welded structural performances. ► We also applied the same technology to PVC pipes. ► We verified the butt welding was successful and effective for the pipes with irregular sections. -- Abstract: In this study, mechanical analyses of a butt welding technology for joining Polyethylene pipe are presented. The pipe had unique structure with double wall, and its section topology was not flat. For an effective repair of leakage and replacements of the pipe, the butt welding technology was developed and tested. For the material characterizations, thermodynamic analyses such as thermal gravimetric analysis and differential scanning calorimetry were performed. Based on the test results, the process temperature and time were determined to ensure safe joining of the pipes using a hot plate apparatus. The welding process was carefully monitored by measuring the temperature. Then, the joined pipes were tested by various methods to evaluate the quality. The analyses results showed the detail process mechanism during the joining process, and the test results demonstrated the successful application of the technology to the sewage pipe repairs.

  5. Analytical study for frequency effects on the EPRI/USNRC piping component tests. Part 1: Theoretical basis and model development

    International Nuclear Information System (INIS)

    Adams, T.M.; Branch, E.B.; Tagart, S.W. Jr.

    1994-01-01

    As part of the engineering effort for the Advanced Light Water Reactor the Advanced Reactor Corporation formed a Piping Technical Core Group to develop a set of improved ASME Boiler and Pressure Vessel Code, Section III design rules and approaches for ALWR plant piping and support design. The technical basis for the proposed changes to the ASME Boiler and Pressure Vessel Code developed by Technical Core Group for the design of piping relies heavily on the failure margins determined from the EPRI/USNRC piping component test program. The majority of the component tests forming the basis for the reported margins against failure were run with input frequency to natural frequency ratios (Ω/ω) in the range of 0.74 to 0.87. One concern investigated by the Technical Core Group was the effect which could exist on measured margins if the tests had been run at higher or lower frequency ratios than those in the limited frequency ratio range tested. Specifically, the concern investigated was that the proposed Technical Core Group Piping Stress Criteria will allow piping to be designed in the low frequency range (Ω/ω ≥ 2.0) for which there is little test data from the EPRI/USNRC test program. The purpose of this analytical study was to: (1) evaluate the potential for margin variation as a function of the frequency ratio (R ω = Ω/ω, where Ω is the forcing frequency and ω is the natural component frequency), (2) recommend a margin reduction factor (MRF) that could be applied to margins determined from the EPRI/USNRC test program to adjust those margins for potential margin variation with frequency ratio. Presented in this paper is the analytical approach and methodology, which are inelastic analysis, which was the basis of the study. Also, discussed is the development of the analytical model, the procedure used to benchmark the model to actual test results, and the various parameter studies conducted

  6. Elemental composition at different points of the rainwater harvesting system

    International Nuclear Information System (INIS)

    Morrow, A.C.; Dunstan, R.H.; Coombes, P.J.

    2010-01-01

    Entry of contaminants, such as metals and non-metals, into rainwater harvesting systems can occur directly from rainfall with contributions from collection surfaces, accumulated debris and leachate from storage systems, pipes and taps. Ten rainwater harvesting systems on the east coast of Australia were selected for sampling of roof runoff, storage systems and tap outlets to investigate the variations in rainwater composition as it moved throughout the system, and to identify potential points of contribution to elemental loads. A total of 26 elements were screened at each site. Iron was the only element which was present in significantly higher concentrations in roof runoff samples compared with tank tap samples (P < 0.05). At one case study site, results suggested that piping and tap material can contribute to contaminant loads of harvested rainwater. Increased loads of copper were observed in hot tap samples supplied by the rainwater harvesting system via copper piping and a storage hot water system (P < 0.05). Similarly, zinc, lead, arsenic, strontium and molybdenum were significantly elevated in samples collected from a polyvinyl chloride pipe sampling point that does not supply household uses, compared with corresponding roof runoff samples (P < 0.05). Elemental composition was also found to vary significantly between the tank tap and an internal cold tap at one of the sites investigated, with several elements fluctuating significantly between the two outlets of interest at this site, including potassium, zinc, manganese, barium, copper, vanadium, chromium and arsenic. These results highlighted the variability in the elemental composition of collected rainwater between different study sites and between different sampling points. Atmospheric deposition was not a major contributor to the rainwater contaminant load at the sites tested. Piping materials, however, were shown to contribute significantly to the total elemental load at some locations.

  7. Instability predictions for circumferentially cracked Type-304 stainless-steel pipes under dynamic loading. Final report. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

    1982-04-01

    This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

  8. Study on pipe deflection by using numerical method

    Science.gov (United States)

    Husaini; Zaki Mubarak, Amir; Agustiar, Rizki

    2018-05-01

    Piping systems are widely used in a refinery or oil and gas industry. The piping system must be properly designed to avoid failure or leakage. Pipe stress analysis is conducted to analyze the loads and critical stress occurred, so that the failure of the pipe can be avoided. In this research, it is analyzed the deflection of a pipe by using Finite Element Method. The pipe is made of A358 / 304SS SCH10S Stainless Steel. It is 16 inches in size with the distance between supports is 10 meters. The fluid flown is Liquid Natural Gas (LNG) with the range of temperature of -120 ° C to -170 ° C, and a density of 461.1 kg / m 3. The flow of LNG causes deflection of the pipe. The pipe deflection must be within the permissible tolerable range. The objective is to analyze the deflection occurred in the piping system. Based on the calculation and simulation, the deflection is 4.4983 mm, which is below the maximum limit of deflection allowed, which is 20.3 mm.

  9. Laboratory piping system vibration tests to determine parametric effects on damping in the seismic frequency range

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-01-01

    A pipe damping research program is being conducted for the United States Nuclear Regulatory Commission at the Idaho National Engineering Laboratory to establish more realistic, best-estimate damping values for use in dynamic structural analyses of piping systems. As part of this program, tests were conducted on a 5-in. (128 mm ID) laboratory piping system to determine the effects of pressure, support configuration, insulation and response amplitude on damping. The tests were designed to produce a wide range of damping values, from very low damping in lightly excited uninsulated systems with few supports, to higher damping under conditions of either/or insulation, high level excitation, and various support arrangements. The effect of pressure at representative seismic levels was considered to be minimal. The supports influence damping at all excitation levels; damping was highest when a mechanical snubber was present in the system. The addition of insulation produced a large increase in damping for the hydraulic shaker excitation tests, but there was no comparable increase for the snapback excitation tests. Once a response amplitude of approximately one-half yield stress was reached, overall damping increased to relatively high levels (>10% of critical)

  10. Simulation and analysis on ultrasonic testing for the cement grouting defects of the corrugated pipe

    Energy Technology Data Exchange (ETDEWEB)

    Qingbang, Han; Ling, Chen; Changping, Zhu [Changzhou Key Laboratory of Sensor Networks and Environmental Sensing, College of IOT, Hohai University Changzhou, Jiangsu, 213022 (China)

    2014-02-18

    The defects exist in the cement grouting process of prestressed corrugated pipe may directly impair the bridge safety. In this paper, sound fields propagation in concrete structures with corrugated pipes and the influence of various different defects are simulated and analyzed using finite element method. The simulation results demonstrate a much complex propagation characteristic due to multiple reflection, refraction and scattering, where the scattering signals caused by metal are very strong, while the signals scattered by an air bubble are weaker. The influence of defect both in time and frequency domain are found through deconvolution treatment. In the time domain, the deconvolution signals correspond to larger defect display a larger head wave amplitude and shorter arrive time than those of smaller defects; in the frequency domain, larger defect also shows a stronger amplitude, lower center frequency and lower cutoff frequency.

  11. Simulation and analysis on ultrasonic testing for the cement grouting defects of the corrugated pipe

    International Nuclear Information System (INIS)

    Qingbang, Han; Ling, Chen; Changping, Zhu

    2014-01-01

    The defects exist in the cement grouting process of prestressed corrugated pipe may directly impair the bridge safety. In this paper, sound fields propagation in concrete structures with corrugated pipes and the influence of various different defects are simulated and analyzed using finite element method. The simulation results demonstrate a much complex propagation characteristic due to multiple reflection, refraction and scattering, where the scattering signals caused by metal are very strong, while the signals scattered by an air bubble are weaker. The influence of defect both in time and frequency domain are found through deconvolution treatment. In the time domain, the deconvolution signals correspond to larger defect display a larger head wave amplitude and shorter arrive time than those of smaller defects; in the frequency domain, larger defect also shows a stronger amplitude, lower center frequency and lower cutoff frequency

  12. Finite Element Limit Pressures for Circumferential Through-Wall Cracks in the Interface between Elbow and Pipe

    International Nuclear Information System (INIS)

    Jang, Yoon-Young; Han, Tae-Song; Huh, Nam-Su; Jeong, Jae-Uk

    2014-01-01

    Among integrity assessment method based on a fracture mechanics concept for piping system, a limit load method is one of the important way to predict a maximum load carrying capacity in the materials with high ductility in the sense that it is used to either assess directly structural integrity of pipe based on fully plastic fracture mechanics or calculate elasticplastic fracture mechanics parameters based on reference stress concept. In nuclear power plants, piping system often involves elbows welded to straight pipe. Since welded regions are vulnerable to cracking, it is important to predict an accurate limit load for pipes with a crack in the interface between elbows and attached pipes. However, although extensive works have been made for developing limit analysis methods for cracked pipes, they were mainly for straight pipes. Recently, limit moment solutions for elbow that is attached to straight pipe with a circumferential through-wall crack(TWC) in the interface were proposed, whereas limit pressure for this geometry is not suggested yet. In this context, plastic limit pressures of circumferential TWCs between elbow and straight pipe were calculated in the present study considering geometric parameters such as an elbow curvature, a pipe size and a crack length. In the present study, the FE plastic limit analyses for circumferential TWC in the interface between elbow and pipe under internal pressure were conducted based on elastic perfectly plastic assumption. Based on the present FE results, it is found that plastic limit pressures of straight pipes with circumferential TWC are not appropriate for predicting plastic limit pressures of circumferential TWC in the interface between elbow and pipe for shorter crack length

  13. Analytical and numerical modeling for flexible pipes

    Science.gov (United States)

    Wang, Wei; Chen, Geng

    2011-12-01

    The unbonded flexible pipe of eight layers, in which all the layers except the carcass layer are assumed to have isotropic properties, has been analyzed. Specifically, the carcass layer shows the orthotropic characteristics. The effective elastic moduli of the carcass layer have been developed in terms of the influence of deformation to stiffness. With consideration of the effective elastic moduli, the structure can be properly analyzed. Also the relative movements of tendons and relative displacements of wires in helical armour layer have been investigated. A three-dimensional nonlinear finite element model has been presented to predict the response of flexible pipes under axial force and torque. Further, the friction and contact of interlayer have been considered. Comparison between the finite element model and experimental results obtained in literature has been given and discussed, which might provide practical and technical support for the application of unbonded flexible pipes.

  14. Computerized tomography used in non-destructive testing of welded pipes

    Energy Technology Data Exchange (ETDEWEB)

    Iovea, M; Rizescu, C; Georgescu, G; Marinescu, A; Chitescu, P; Sava, T; Neagu, M; Avram, D [Institute of Research and Design for Electrical Engineering, ICPE - Electrostatica Splaiul Unirii 313, Sect. 3, R-74204 Bucharest (Romania)

    1997-12-31

    High quality standards in operation of National Power System is ensured by the use of high performance techniques and systems for Non-Destructive Testing (NDT). In recent years a number of new developments of the non-conventional technologies in the field of NDT have been achieved. In our laboratory there have been developed two computerized technologies using {gamma}-ray computed tomography and ultrasonic imaging methods. The standard techniques for imaging from projection data is computerized tomography. The industrial computerized tomography methods consist in the measurement of thin X - or {gamma}-ray beam attenuation when passing through some selected surface of the tested object, along several directions, so that by means of an adequate mathematical algorithm, a map of linear attenuation coefficients for the scanned surface is obtained. In fact, this map gives the density of materials occurring in the surface plane. Computerized tomography equipment, in various constructive versions, are intended for the following applications: (1) NDT in those fields requiring strict control of product quality, as for instance the nuclear energy, military industry, aeronautics, transportation fields, etc., (2) research in field of materials technology, machine engineering, metallurgy, welding, etc. This paper presents the applications of Computerized Tomography in NDT, by showing the results obtained on welded pipes, as well as the facilities offered by this method. In the final part, the paper presents the concept of a mobile tomography system for industrial pipes testing. (author). 2 figs., 7 refs.

  15. Computerized tomography used in non-destructive testing of welded pipes

    International Nuclear Information System (INIS)

    Iovea, M.; Rizescu, C.; Georgescu, G.; Marinescu, A.; Chitescu, P.; Sava, T.; Neagu, M.; Avram, D.

    1996-01-01

    High quality standards in operation of National Power System is ensured by the use of high performance techniques and systems for Non-Destructive Testing (NDT). In recent years a number of new developments of the non-conventional technologies in the field of NDT have been achieved. In our laboratory there have been developed two computerized technologies using γ-ray computed tomography and ultrasonic imaging methods. The standard techniques for imaging from projection data is computerized tomography. The industrial computerized tomography methods consist in the measurement of thin X - or γ-ray beam attenuation when passing through some selected surface of the tested object, along several directions, so that by means of an adequate mathematical algorithm, a map of linear attenuation coefficients for the scanned surface is obtained. In fact, this map gives the density of materials occurring in the surface plane. Computerized tomography equipment, in various constructive versions, are intended for the following applications: 1) NDT in those fields requiring strict control of product quality, as for instance the nuclear energy, military industry, aeronautics, transportation fields, etc., 2) research in field of materials technology, machine engineering, metallurgy, welding, etc. This paper presents the applications of Computerized Tomography in NDT, by showing the results obtained on welded pipes, as well as the facilities offered by this method. In the final part, the paper presents the concept of a mobile tomography system for industrial pipes testing. (author). 2 figs., 7 refs

  16. Long time durability tests of fabric inlet stratification pipes

    DEFF Research Database (Denmark)

    Andersen, Elsa; Furbo, Simon

    2008-01-01

    and that this destroys the capability of building up thermal stratification for the fabric inlet stratification pipe. The results also show that although dirt, algae etc. are deposited in the fabric pipes in the space heating tank, the capability of the fabric inlet stratifiers to build up thermal stratification...

  17. Piping system damping data at higher frequencies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-01-01

    Research has been performed at the Idaho National Engineering Laboratory (INEL) for the United States Nuclear Regulatory Commission (USNRC) to determine best-estimate damping values for dynamic analyses of nuclear piping systems excited in the 20 to 100 Hz frequency range. Vibrations in this frequency range are typical of fluid-induced transients, for which no formal pipe damping guidelines exist. The available data found in the open literature and the USNRC/INEL nuclear piping damping data bank were reviewed, and a series of tests on a straight 3-in. (76-mm) piping system and a 5-in. (127-mm) system with several bends and elbows were conducted as part of this research program. These two systems were supported with typical nuclear piping supports that could be changed from test to test during the series. The resulting damping values were ≥ those of the Pressure Vessel Research Committee (PVRC) proposal for unisulated piping. Extending the PVRC damping curve from 20 to 100 Hz at 3% of critical damping would give a satisfactory representation of the test data. This position has been endorsed by the PVRC Technical Committee on Piping Systems. 14 refs

  18. Development of a piping thickness monitoring system using equipotential switching direct current potential drop method

    International Nuclear Information System (INIS)

    Kyung Ha, Ryu; Na Young, Lee; Il Soon, Hwang

    2007-01-01

    As nuclear power plants age, low alloy steel piping undergoes wall thickness reduction due to Flow Accelerated Corrosion (FAC). Persisting pipe rupture accidents prompted thinned pipe management programs. As a consequence extensive inspection activities are made based on the Ultrasonic Technique (UT). As the inspection points increase, time is needed to cover required inspection areas. In this paper, we present the Wide Range Monitoring (WiRM) concept with Equipotential Switching Direct Current Potential Drop (ES-DCPD) method by which FAC-active areas can be screened for detailed UT inspections. To apply ES-DCPD, we developed an electric resistance network model and electric field model based on Finite Element Analysis (FEA) to verify its feasibility. Experimentally we measured DCPD of the pipe elbow and confirmed the validity using UT inspections. For a more realistic validation test, we designed a high temperature flow test loop with environmental parameters turned for FAC simulation in the laboratory. Using electrochemical monitoring of water chemistry and local flow velocity prediction by computational fluid dynamic model, FAC rate is estimated. Based on the FAC prediction model and the simulation loop test, we plan to demonstrate the applicability of ES-DCPD in the PWR secondary environment. (authors)

  19. Replaceable liquid nitrogen piping

    International Nuclear Information System (INIS)

    Yasujima, Yasuo; Sato, Kiyoshi; Sato, Masataka; Hongo, Toshio

    1982-01-01

    This liquid nitrogen piping with total length of about 50 m was made and installed to supply the liquid nitrogen for heat insulating shield to three superconducting magnets for deflection and large super-conducting magnet for detection in the π-meson beam line used for high energy physics experiment in the National Laboratory for High Energy Physics. The points considered in the design and manufacture stages are reported. In order to minimize the consumption of liquid nitrogen during transport, vacuum heat insulation method was adopted. The construction period and cost were reduced by the standardization of the components, the improvement of welding works and the elimination of ineffective works. For simplifying the maintenance, spare parts are always prepared. The construction and the procedure of assembling of the liquid nitrogen piping are described. The piping is of double-walled construction, and its low temperature part was made of SUS 316L. The super-insulation by aluminum vacuum evaporation and active carbon were attached on the external surface of the internal pipe. The final leak test and the heating degassing were performed. The tests on evacuation, transport capacity and heat entry are reported. By making the internal pipe into smaller size, the piping may be more efficient. (Kako, I.)

  20. Introduction to Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

    This is the presentation file for the short course Introduction to Heat Pipes, to be conducted at the 2015 Thermal Fluids and Analysis Workshop, August 3-7, 2015, Silver Spring, Maryland. NCTS 21070-15. Course Description: This course will present operating principles of the heat pipe with emphases on the underlying physical processes and requirements of pressure and energy balance. Performance characterizations and design considerations of the heat pipe will be highlighted. Guidelines for thermal engineers in the selection of heat pipes as part of the spacecraft thermal control system, testing methodology, and analytical modeling will also be discussed.

  1. Literature review and experimental investigation of heat pipes

    Science.gov (United States)

    Barsch, W. O.; Schoenhals, R. J.; Viskanta, R.; Winter, E. R. F.

    1971-01-01

    Tests on heat pipes determine operational limits, external boundary conditions, noncondensable gas effects, startup behavior, and geometric configurations. Experiment consists of design, construction, and testing of an apparatus for measuring wick properties, conventional heat pipes and coplanar heat pipes.

  2. Nonlinear dynamic analysis of high energy line pipe whip

    International Nuclear Information System (INIS)

    Hsu, L.C.; Kuo, A.Y.; Tang, H.T.

    1983-01-01

    To facilitate potential cost savings in pipe whip protection design, TVA conducted a 1'' high pressure line break test to investigate the pipe whip behavior. The test results are available to EPRI as a data base for a generic study on nonlinear dynamic behavior of piping systems and pipe whip phenomena. This paper describes a nonlinear dynamic analysis of the TVA high energy line tests using ABAQUS-EPGEN code. The analysis considers the effects of large deformation and high strain rate on resisting moment and energy absorption capability of the analyzed piping system. The numerical results of impact forces, impact velocities, and reaction forces at pipe supports are compared to the TVA test data. The pipe whip impact time and forces have also been calculated per the current NRC guidelines and compared. The calculated pipe support reaction forces prior to impact have been found to be in good agreement with the TVA test data except for some peak values at the very beginning of the pipe break. These peaks are believed to be due to stress wave propagation which cannot be addressed by the ABAQUS code. Both the effects of elbow crushing and strain rate have been approximately simulated. The results are found to be important on pipe whip impact evaluation. (orig.)

  3. Development of a Short-term Failure Assessment of High Density Polyethylene Pipe Welds - Application of the Limit Load Analysis -

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Ho-Wan; Han, Jae-Jun; Kim, Yun-Jae [Korea University, Seoul (Korea, Republic of); Kim, Jong-Sung [Sunchon National University, Suncheon (Korea, Republic of); Kim, Jeong-Hyeon; Jang, Chang-Heui [KAIST, Daejeon (Korea, Republic of)

    2015-04-15

    In the US, the number of cases of subterranean water contamination from tritium leaking through a damaged buried nuclear power plant pipe continues to increase, and the degradation of the buried metal piping is emerging as a major issue. A pipe blocked from corrosion and/or degradation can lead to loss of cooling capacity in safety-related piping resulting in critical issues related to the safety and integrity of nuclear power plant operation. The ASME Boiler and Pressure Vessel Codes Committee (BPVC) has recently approved Code Case N-755 that describes the requirements for the use of polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. This paper contains tensile and slow crack growth (SCG) test results for high-density polyethylene (HDPE) pipe welds under the environmental conditions of a nuclear power plant. Based on these tests, the fracture surface of the PENT specimen was analyzed, and the fracture mechanisms of each fracture area were determined. Finally, by using 3D finite element analysis, limit loads of HDPE related to premature failure were verified.

  4. Refined inelastic analysis of piping systems using a beam-type program

    International Nuclear Information System (INIS)

    Millard, A.; Hoffmann, A.

    1981-08-01

    A finite element for inelastic piping analysis has been presented, which enables accounting for local effects like thermal gradients and supplies local states of stresses and strains, while keeping all the advantages of a classical beam type program (easy to use, simple boundary conditions, cost effectiveness). Thanks to the local description of the cross section, geometrical non-linearity due to inertia modification can be introduced together with material non-linearity. The element can also be degenerated into a straight pipe element

  5. Performance correlations for high temperature potassium heat pipes

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Keddy, E.S.; Sena, J.T.

    1987-01-01

    Potassium heat pipes designed for operation at a nominal temperature of 775K have been developed for use in a heat pipe cooled reactor design. The heat pipes operate in a gravity assist mode with a maximum required power throughput of approximately 16 kW per heat pipe. Based on a series of sub-scale experiments with 2.12 and 3.2 cm diameter heat pipes the prototypic heat pipe diameter was set at 5.7 cm with a simple knurled wall wick used in the interests of mechanical simplicity. The performance levels required for this design had been demonstrated in prior work with gutter assisted wicks and emphasis in the present work was on the attainment of similar performance with a simplified wick structure. The wick structure used in the experiment consisted of a pattern of knurled grooves in the internal wall of the heat pipe. The knurl depth required for the planned heat pipe performance was determined by scaling of wick characteristic data from the sub-scale tests. These tests indicated that the maximum performance limits of the test heat pipes did not follow normal entrainment limit predictions for textured wall gravity assist heat pipes. Test data was therefore scaled to the prototype design based on the assumption that the performance was controlled by an entrainment parameter based on the liquid flow depth in the groove structure. This correlation provided a reasonable fit to the sub-scale test data and was used in scale up of the design from the 8.0 cm 2 cross section of the largest sub-scale heat pipe to the 25.5 cm 2 cross section prototype. Correlation of the model predictions with test data from the prototype is discussed

  6. Evaluation of piping fracture analysis method by benchmark study, 1

    International Nuclear Information System (INIS)

    Takahashi, Yukio; Kashima, Koichi; Kuwabara, Kazuo

    1987-01-01

    Importance of strength evaluation methods for cracked piping is growing with the progress of the rationalization of the nuclear piping system based on the leak-before-break concept. As an analytical tool, finite element method is principally used. To obtain the reliable solutions by the finite element programs, it is important to grasp the influences of various factors on the solutions. In this study, benchmark analysis is carried out for a stainless steel pipe with a circumferential through-wall crack subjected to four-point bending loading. Eight solutions obtained by using five finite element programs are compared with each other. Good agreement is obtained between the solutions on the deformation characteristics as well as fracture mechanics parameters. It is found through this study that the influence of the difference in the solution technique is generally small. (author)

  7. Pressurized water-reactor feedwater piping response to water hammer

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

    The nuclear power industry is interested in steam-generator water hammer because it has damaged the piping and components at pressurized water reactors (PWRs). Water hammer arises when rapid steam condensation in the steam-generator feedwater inlet of a PWR causes depressurization, water-slug acceleration, and slug impact at the nearest pipe elbow. The resulting pressure pulse causes the pipe system to shake, sometimes violently. The objective of this study is to evaluate the potential structural effects of steam-generator water hammer on feedwater piping. This was accomplished by finite-element computation of the response of two sections of a typical feedwater pipe system to four representative water-hammer pulses. All four pulses produced high shear and bending stresses in both sections of pipe. Maximum calculated pipe stresses varied because the sections had different characteristics and were sensitive to boundary-condition modeling

  8. Failure behavior of a pipe system with a circumferentially orientated flaw - analytical and experimental investigations

    International Nuclear Information System (INIS)

    Mikkola, T.P.J.; Diem, H.; Blind, D.; Hunger, H.

    1989-01-01

    At the german HDR-test-facility a pipe failure experiment was performed at a fullsize feedwater piping system under operating conditions of T=240 0 C, p=10.6 MPa and with an elevated oxygen content in the pressure medium. The loading was internal pressure and a cyclic varying bending moment with an R-ratio of 0.5. The in form of a circumferentially orientated notch initially weakened piping system failed after a total number of 4773 loaded cycles with different frequencies in form of a small leak. The analyses of the fracture surface indicated the strongly growing influence of corrosion effects on the crack propagation rate with decreasing loading frequency. The cyclic crack growth and the leak-before-break behavior of the piping system could be explained on the basis of results of finite element calculations using ADINA-code. (orig.)

  9. Pipe whip analysis using the TEDEL code

    International Nuclear Information System (INIS)

    Millard, D.; Hoffmann, A.

    1985-02-01

    In view of their abundance, piping systems are one of the main components in power industries and in particular in nuclear power plants. They must be designed for normal as well as faulted conditions, for safety requirements. The prediction of the dynamic behaviour of the free pipe requires accounting for several nonlinearities. For this purpose, a beam type finite element program (TEDEL) has been used. The aim of this paper is to enlight the main features of this program, when applied to pipe whip analysis. An example of application to a real case will also be presented

  10. The Geometry Optimisation of a Triple Branch Pipe Using Finite Element Method

    Directory of Open Access Journals (Sweden)

    Dorian Nedelcu

    2008-01-01

    Full Text Available The paper presents the geometrical optimization of a triple branch pipesubmitted to an internal pressure. The goal of the optimization was todetermine the optimum thickness of piping and branch pipe ribs, in thecondition of reaching admissible values of the stress and displacement.The resistance calculus was realized with Cosmos DesignStar softwareand the geometry was modeled with Microstation Modeler software.

  11. Ultrasonic guided waves in eccentric annular pipes

    International Nuclear Information System (INIS)

    Pattanayak, Roson Kumar; Balasubramaniam, Krishnan; Rajagopal, Prabhu

    2014-01-01

    This paper studies the feasibility of using ultrasonic guided waves to rapidly inspect tubes and pipes for possible eccentricity. While guided waves are well established in the long range inspection of structures such as pipes and plates, studies for more complex cross sections are limited and analytical solutions are often difficult to obtain. Recent developments have made the Semi Analytical Finite Element (SAFE) method widely accessible for researchers to study guided wave properties in complex structures. Here the SAFE method is used to study the effect of eccentricity on the modal structures and velocities of lower order guided wave modes in thin pipes of diameters typically of interest to the industry. Results are validated using experiments. The paper demonstrates that even a small eccentricity in the pipe can strongly affect guided wave mode structures and velocities and hence shows potential for pipe eccentricity inspection

  12. INEL/USNRC pipe damping experiments and studies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1987-08-01

    Since the previous paper on this subject presented at the 8th SMiRT Conference, the Idaho National Engineering Laboratory (INEL) has conducted further research on piping system damping for the United States Nuclear Regulatory Commission (USNRC). These efforts have included vibration tests on two laboratory piping systems at response frequencies up to 100 Hz, and damping data calculations from both of these two systems and from a third laboratory piping system test series. In addition, a statistical analysis was performed on piping system damping data from tests representative of seismic and hydrodynamic events of greater than minimal excitation. The results of this program will be used to assist regulators in establishing suitable damping values for use in dynamic analyses of nuclear piping systems, and in revising USNRC Regulatory Guide (RG) 1.61

  13. On the computer simulation of LMFBR piping systems

    International Nuclear Information System (INIS)

    A-Moneim, M.T.; Chang, Y.-W.; Fistedis, S.H.

    1977-01-01

    A two-dimensional coupled hydrodynamic-structural response analysis of piping systems is described. Implicit Continuous-Fluid Eulerian (ICE) technique is utilized in the hydrodynamics while a finite-element technique is used in the structural analysis. Different piping components such as elbows, valves, reducers, expansions, heat exchangers, and tees are modelled and coupled with the straight pipe model. An axisymmetric general component model that can be used in modelling valves, reducers, expansions, and heat exchangers is described. At the inlet and outlet region of such component the cross-sectional area may change suddently or gradually, or many not change at all. Among the options available in this model are deformable exterior walls, interior rigid wall simulation, and tube bundle effect. Exterior walls of pipes and components are treated as thin axisymmetric shell. A convected coordinate explicit finite-element scheme for large displacement small strain, elastic-plastic material behavior in which membrane and bending strengths are accounted for is employed. The strains are linearly related to the displacement of the element relative to its convective coordinates, and similarly, the nodal forces are linearly related to the elements stresses. The coupling of the hydrodynamics and structural problems is done in such a way that the hydrodynamics supplies the structure with a pressure loading and the structure supplying the hydrodynamics with a moving boundary condition. Because of the difficulties of handling interior walls that may occupy partial zones, the walls are assumed rigid and limited in their orientation to be parallel to the radial or axial directions, their position to zone boundaries, and their thickness to zero

  14. Flexibility of trunnion piping elbows

    International Nuclear Information System (INIS)

    Lewis, G.D.; Chao, Y.J.

    1987-01-01

    Flexibility factors and stress indices for piping component such as straight pipe, elbows, butt-welding tees, branch connections, and butt-welding reducers are contained in the code, but many of the less common piping components, like the trunnion elbow, do not have flexibility factors or stress indices defined. The purpose of this paper is to identify the in-plane and out-of-plane flexibility factors in accordance with code procedures for welded trunnions attached to the tangent centerlines of long radius elbows. This work utilized the finite element method as applicable to plates and shells for calculating the relative rotations of the trunnion elbow-ends for in-plane and out-of-plane elbow moment loadings. These rotations are used to derive the corresponding in-plane and out-of-plane flexibility factors. (orig./GL)

  15. Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system

    International Nuclear Information System (INIS)

    Lowry, W.

    1994-01-01

    The objective for the development of the Pipe Explorer trademark radiological characterization system is to achieve a cost effective, low risk means of characterizing gamma radioactivity on the inside surface of pipes. The unique feature of this inspection system is the use of a pneumatically inflated impermeable membrane which transports the detector into the pipe as it inverts. The membrane's internal air pressure tows the detector and tether through the pipe. This mechanism isolates the detector and its cabling from the contaminated surface, yet allows measurement of radioactive emissions which can readily penetrate the thin plastic membrane material (such as gamma and high energy beta emissions). In Phase 1, an initial survey of DOE facilities was conducted to determine the physical and radiological characteristics of piping systems. The inverting membrane deployment system was designed and extensively tested in the laboratory. A range of membrane materials was tested to evaluate their ruggedness and deployment characteristics. Two different sizes of gamma scintillation detectors were procured and tested with calibrated sources. Radiation transport modeling evaluated the measurement system's sensitivity to detector position relative to the contaminated surface, the distribution of the contamination, background gamma levels, and gamma source energy levels. In the culmination of Phase 1, a field demonstration was conducted at the Idaho National Engineering Laboratory's Idaho Chemical Processing Plant. The project is currently in transition from Phase 1 to Phase 2, where more extensive demonstrations will occur at several sites. Results to date are discussed

  16. Leak-Before-Break assessment of a welded piping based on 3D finite element method

    International Nuclear Information System (INIS)

    Chen, Mingya; Yu, Weiwei; Chen, Zhilin; Qian, Guian; Lu, Feng; Xue, Fei

    2017-01-01

    Highlights: • The effects of load reduction, strength match, welding width, load level, crack size and constraint are studied. • The results show that the LBB margin is dependent on the load level. • The results show that higher strength-match of WPJs will have higher crack-front constraints. • The results show that the engineering method has a high precision only if the width of weld is comparable to the crack depth. - Abstract: The paper studies the effects of the load reduction (discrepancy between designing and real loadings), strength match of the welded piping joint (WPJ), welding width, crack size and crack tip constraint on the Leak-Before-Break (LBB) assessment of a welded piping. The 3D finite element (FE) method is used in the study of a surge line of the steam generator in a nuclear power plant. It is demonstrated that the LBB margin is dependent on the loading level and the load reduction effect should be considered. When the loading is high enough, there is a quite large deviation between the J-integral calculated based on the real material property of WPJ and that calculated based on the engineering method, e.g. Zahoor handbook of Electric Power Research Institute (EPRI). The engineering method assumes that the whole piping is made of the unique welding material in the calculation. As the influence of the strength matching and welding width is ignored in the engineering method for J-integral calculation, the engineering method has a sufficient precision only if the width of welding is comparable to the crack depth. Narrower welding width leads to higher constraint of the plastic deformation in the welding and larger high stress areas in the base for the low strength-match WPJ. Higher strength matching of WPJs has higher crack-front constraints.

  17. Study of the performances of acoustic emission testing for glass fibre reinforced plastic pipes containing defects

    International Nuclear Information System (INIS)

    Villard, D.; Vidal, M.C.

    1995-08-01

    Glass fibre reinforced plastic pipes are more and more often used, in nuclear power plants, for building or replacement of water pipings classified 'nuclear safety'. Tests have been performed to evaluate the performances of acoustic emission testing for in service inspection of these components. The tests were focused on glass fibre reinforced polyester and vinyl-ester pipes, in as received conditions or containing impacts, and intentionally introduced defects. They have been carried out by CETIM, following the ASTM Standard E 1118 (code CARP), to a maximum pressure lever of 25 Bar The results show that the CARP procedure can be used for detection of defects and evaluation of their noxiousness towards internal pressure: most of the tubes containing low energy impacts could not be distinguished from tubes without defect; on the other hand the important noxiousness of lacks of impregnation of roving layer appeared clearly. Complementary tests have been performed on some tubes at a more important pressure lever, for which the damage of the tubes in enough to deteriorate there elastic properties. The results showed that CARP procedure give valuable informations on damage level. It would be interesting to evaluate acoustic emission on tubes containing realistic in-service degradations. (author). 11 refs., 15 figs., 6 tabs., 2 appends

  18. Development of testing system for the thermo-mechanical fatigue crack analysis of nuclear power plant pipes

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Kim, Maan Won; Lee, Bong Sang

    2003-12-01

    Fatigue crack growth analysis plays an important role in the structural integrity assessment or the service life calculation of the nuclear power plant pipes. To obtain the material properties as a basic data to achieve an accurate crack growth analysis, a lot of tests and numerical crack growth simulations have been done for decades. The BS 7910 or the ASME Boiler and Pressure Vessel Code Section XI, generally used to evaluate crack growth behavior, were made under the based on simple stress states or at the evaluated isothermal temperature. It is well known that the ASME code could sometimes give so conservative results in some cases of which the cracked components are experiencing with cyclic thermal shock. In this report, we suggested a method for the life assessment of a crack embedded in nuclear power plant pipes under the thermal-mechanical fatigue loads. We here use the numerical method to get the temperature history for thermal- mechanical fatigue crack growth test. And then we can calculate the remaining life time of the pipe by using the fracture mechanics and the test results together. For this purpose, we constructed a thermal-mechanical fatigue crack growth testing system. We also gave a lot of review about recent researches in the experimental field of thermal-mechanical fatigue analysis

  19. Application of the cracked pipe element to creep crack growth prediction

    Energy Technology Data Exchange (ETDEWEB)

    Brochard, J.; Charras, T. [C.E.A.-C.E.-Saclay DRN/DMT, Gif Sur Yvette (France); Ghoudi, M. [C.E.A.-C.E.-Saclay, Gif Sur Yvette (France)

    1997-04-01

    Modifications to a computer code for ductile fracture assessment of piping systems with postulated circumferential through-wall cracks under static or dynamic loading are very briefly described. The modifications extend the capabilities of the CASTEM2000 code to the determination of fracture parameters under creep conditions. The main advantage of the approach is that thermal loads can be evaluated as secondary stresses. The code is applicable to piping systems for which crack propagation predictions differ significantly depending on whether thermal stresses are considered as primary or secondary stresses.

  20. Evaluation of stress histories of reactor coolant loop piping for pipe rupture prediction

    International Nuclear Information System (INIS)

    Lu, S.C.; Larder, R.A.; Ma, S.M.

    1981-01-01

    This paper describes the analyses used to evaluate stress histories in the primary coolant loop piping of a selected four-loop PNR power station. In order to make the simulation as realistic as possible, best estimates rather than conservative assumptions were considered throughout. The best estimate solution, however, was aided by a sensitivity study to assess the possible variation of outcomes resulted from uncertainties associated with these assumptions. Sources of stresses considered in the evaluation were pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, pump vibrations, and finally seismic excitations. The best estimates of pressure and thermal transient histories arising from plant operations were based on actual plant operation records supplemented by specified plant design conditions. Seismic motions were generated from response spectrum curves developed specifically for the region surrounding the plant site. Stresses due to dead weight and thermal expansion were computed from a three dimensional finite element model which used a combination of pipe, truss, and beam elements to represent the coolant loop piping, the pressure vessel, coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients were obtained by closed form solutions. Seismic stress calculations considered the soil structure interaction, the coupling effect between the containment structure and the reactor coolant system. A time history method was employed for the seismic analysis. Calculations of residual stresses accounted for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation were estimated by a dynamic analysis using existing measurements of pump vibrations. (orig./HP)

  1. Determination of Secondary Encasement Pipe Design Pressure

    Energy Technology Data Exchange (ETDEWEB)

    TEDESCHI, A.R.

    2000-10-26

    This document published results of iterative calculations for maximum tank farm transfer secondary pipe (encasement) pressure upon failure of the primary pipe. The maximum pressure was calculated from a primary pipe guillotine break. Results show encasement pipeline design or testing pressures can be significantly lower than primary pipe pressure criteria.

  2. Structural Dynamic Assessment of the GN2 Piping System for NASA's New and Powerful Reverberant Acoustic Test Facility

    Science.gov (United States)

    McNelis, Mark E.; Staab, Lucas D.; Akers, James C.; Hughes, WIlliam O.; Chang, Li, C.; Hozman, Aron D.; Henry, Michael W.

    2012-01-01

    The National Aeronautics and Space Administration (NASA) Glenn Research Center (GRC) has led the design and build of the new world-class vibroacoustic test capabilities at the NASA GRC's Plum Brook Station in Sandusky, Ohio, USA from 2007-2011. SAIC-Benham has completed construction of a new reverberant acoustic test facility to support the future testing needs of NASA's space exploration program and commercial customers. The large Reverberant Acoustic Test Facility (RATF) is approximately 101,000 cu ft in volume and was designed to operate at a maximum empty chamber acoustic overall sound pressure level (OASPL) of 163 dB. This combination of size and acoustic power is unprecedented amongst the world's known active reverberant acoustic test facilities. Initial checkout acoustic testing was performed on March 2011 by SAIC-Benham at test levels up to 161 dB OASPL. During testing, several branches of the gaseous nitrogen (GN2) piping system, which supply the fluid to the noise generating acoustic modulators, failed at their "t-junctions" connecting the 12 inch supply line to their respective 4 inch branch lines. The problem was initially detected when the oxygen sensors in the horn room indicated a lower than expected oxygen level from which was inferred GN2 leaks in the piping system. In subsequent follow up inspections, cracks were identified in the failed "t-junction" connections through non-destructive evaluation testing . Through structural dynamic modeling of the piping system, the root cause of the "t-junction" connection failures was determined. The structural dynamic assessment identified several possible corrective design improvements to the horn room piping system. The effectiveness of the chosen design repairs were subsequently evaluated in September 2011 during acoustic verification testing to 161 dB OASPL.

  3. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M. Kothari; Gerard T. Pittard

    2003-06-01

    range of coiled-tubing string designs and pipe environments. Work conducted in the second quarter consisted of: (1) selecting a preferred pan/zoom/tilt camera; (2) initiating design of the digital control electronics and switching power supply for the control and operation of the in-pipe robotic modules; (3) continuing design of the repair sleeve and (4) initial testing of the wall-cleaning device. Activities in the third quarter included: (1) development of the system's pan/zoom/tilt camera control electronics and operating software, and implementing these in the surface and downhole modules and (2) further testing of the wall-cleaning elements used to clean the inside of the bell and spigot joints. Most recently, fourth quarter developments were centered on designing and testing the pipe-wall cleaning device including the selection of the drive motor and its control electronics. In addition, efforts were also focused on the design of the repair sleeve. Details of these activities are described in the body of the report along with a summary of events scheduled for the next quarter.

  4. FSI analysis of piping systems under seismic excitation

    International Nuclear Information System (INIS)

    Uras, R.A.; Ma, D.C.; Chang, Yao W.; Liu, Wing Kam

    1991-01-01

    A formulation which accounts for fluid-structure interaction of piping system under seismic excitation is presented. The governing equations of the fluid and the structure to model the pipe are stated. Using the finite element method the discretized equations are obtained. A transformation procedure for proper assembly of matrices is introduced. A solution algorithm is described. 9 refs., 2 figs

  5. Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report

    International Nuclear Information System (INIS)

    Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

    1982-04-01

    This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models

  6. Standard practice for ultrasonic testing of the Weld Zone of welded pipe and tubing

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2010-01-01

    1.1 This practice describes general ultrasonic testing procedures for the detection of discontinuities in the weld and adjacent heat affected zones of welded pipe and tubing by scanning with relative motion between the search unit and pipe or tube. When contact or unfocused immersion search units are employed, this practice is intended for tubular products having specified outside diameters ≥2 in. (≥50 mm) and specified wall thicknesses of 1/8to 11/16 in. (3 to 27 mm). When properly focused immersion search units are employed, this practice may also be applied to material of smaller diameter and thinner wall. Note 1—When contact or unfocused immersion search units are used, precautions should be exercised when examining pipes or tubes near the lower specified limits. Certain combinations of search unit size, frequency, thin–wall thicknesses, and small diameters could cause generation of unwanted sound waves that may produce erroneous examination results. 1.2 All surfaces of material to be examined in ...

  7. Base-plate effects on pipe-support stiffness

    International Nuclear Information System (INIS)

    Winkel, B.V.; LaSalle, F.R.

    1981-01-01

    Present nuclear power plant design methods require that pipe support spring rates be considered in the seismic design of piping systems. Base plate flexibility can have a significant effect on the spring rates of these support structures. This paper describes the field inspection, test, and analytical techniques used to identify and correct excessively flexible base plates on the Fast Flux Test Facility pipe support structures

  8. A comparison of time-history elastic plastic piping analysis with measurement

    International Nuclear Information System (INIS)

    Scavuzzo, R.J.; Sansalone, K.H.

    1992-01-01

    The GE/ETEC Green piping system was subjected to high seismic inputs from hydraulic sleds at each pipe foundation. These inputs were high enough to force bending stresses into the plastic regime. Strain gages recorded the pipe response at various positions within the system. The ABAQUS finite element code was used to model this piping system and the dynamic input. Problems associated with the dynamic input are discussed. Various types of finite elements were evaluated for accurancy. Both an elastic time-history analysis and an elastic-plastic time-history analysis of the system were conducted. Results of these analyses are compared to each other and the experimental data. These comparisons indicated that elastic analysis of dynamic strains are conservative at all points of comparison and that there is good agreement between the nonlinear elastic-plastic analysis and experimental data. (orig.)

  9. Thermal expansion movements of piping during FFTF plant startup

    International Nuclear Information System (INIS)

    Lindquist, M.R.

    1981-03-01

    FFTF liquid metal piping exhibits significant displacements during heatup of the plant heat transport system. Verification of correct piping movements is important to assure that no restraints are present and to provide data for additional piping design/analysis validation. A test program is described in which a series of measurements were taken at selected piping locations. These data were obtained during Plant Acceptance Testing involving system heatup cycles to approximately 800 0 F(427 0 C). Typical test data are shown and compared to analytical predictions. Two piping system problems that were identified as a result of the testing are described along with resolutions thereof. Establishment of final baseline data is discussed

  10. Theory and application of a three-dimensional code SHAPS to complex piping systems

    International Nuclear Information System (INIS)

    Wang, C.Y.

    1983-01-01

    This paper describes the theory and application of a three-dimensional computer code SHAPS to the complex piping systems. The code utilizes a two-dimensional implicit Eulerian method for the hydrodynamic analysis together with a three-dimensional elastic-plastic finite-element program for the structural calculation. A three-dimensional pipe element with eight degrees of freedom is employed to account for the hoop, flexural, axial, and the torsional mode of the piping system. In the SHAPS analysis the hydrodynamic equations are modified to include the global piping motion. Coupling between fluid and structure is achieved by enforcing the free-slip boundary conditions. Also, the response of the piping network generated by the seismic excitation can be included. A thermal transient capability is also provided in SHAPS. To illustrate the methodology, many sample problems dealing with the hydrodynamic, structural, and thermal analyses of reactor-piping systems are given. Validation of the SHAPS code with experimental data is also presented

  11. ON-POWER DETECTION OF PIPE WALL-THINNED DEFECTS USING IR THERMOGRAPHY IN NPPS

    Directory of Open Access Journals (Sweden)

    JU HYUN KIM

    2014-04-01

    Full Text Available Wall-thinned defects caused by accelerated corrosion due to fluid flow in the inner pipe appear in many structures of the secondary systems in nuclear power plants (NPPs and are a major factor in degrading the integrity of pipes. Wall-thinned defects need to be managed not only when the NPP is under maintenance but also when the NPP is in normal operation. To this end, a test technique was developed in this study to detect such wall-thinned defects based on the temperature difference on the surface of a hot pipe using infrared (IR thermography and a cooling device. Finite element analysis (FEA was conducted to examine the tendency and experimental conditions for the cooling experiment. Based on the FEA results, the equipment was configured before the cooling experiment was conducted. The IR camera was then used to detect defects in the inner pipe of the pipe specimen that had artificially induced defects. The IR thermography developed in this study is expected to help resolve the issues related to the limitations of non-destructive inspection techniques that are currently conducted for NPP secondary systems and is expected to be very useful on the NPPs site.

  12. Analysis and Optimisation of Carcass Production for Flexible Pipes

    DEFF Research Database (Denmark)

    Nielsen, Peter Søe

    Un-bonded flexible pipes are used in the offshore oil and gas industry worldwide transporting hydrocarbons from seafloor to floating production vessels topside. Flexible pipes are advantageous over rigid pipelines in dynamic applications and during installation as they are delivered in full length......-axial tension FLC points were attained. Analysis of weld fracture of duplex stainless steel EN 1.4162 is carried out determining strains with GOM ARAMIS automated strain measurement system, which shows that strain increases faster in the weld zone than the global strain of the parent material. Fracture...... is the analysis and optimisation of the carcass manufacturing process by means of a fundamental investigation in the fields of formability, failure modes / mechanisms, Finite Element Analysis (FEA), simulative testing and tribology. A study of failure mechanisms in carcass production is performed by being present...

  13. Calculations of Edwards' pipe blowdown tests using the code TRAC P1

    International Nuclear Information System (INIS)

    O'Mahoney, R.

    1979-05-01

    The paper describes the results obtained using the non-thermal equilibrium LOCA code TRAC-P1 for two of a series of Pipe Blowdown Tests. Comparisons are made with the experimental values and RELAP-UK Mark IV predictions. Some discrepancies between prediction and experiment are observed, and certain aspects of the model are considered to warrant possible further attention. (U.K.)

  14. Comparative study on deformation and mechanical behavior of corroded pipe: Part I–Numerical simulation and experimental investigation under impact load

    Directory of Open Access Journals (Sweden)

    Dong-Man Ryu

    2017-09-01

    Full Text Available Experiments and a numerical simulation were conducted to investigate the deformation and impact behavior of a corroded pipe, as corrosion, fatigue, and collision phenomena frequently occur in subsea pipelines. This study focuses on the deformation of the corrosion region and the variation of the geometry of the pipe under impact loading. The experiments for the impact behavior of the corroded pipe were performed using an impact test apparatus to validate the results of the simulation. In addition, during the simulation, material tests were performed, and the results were applied to the simulation. The ABAQUS explicit finite element analysis program was used to perform numerical simulations for the parametric study, as well as experiment scenarios, to investigate the effects of defects under impact loading. In addition, the modified ASME B31.8 code formula was proposed to define the damage range for the dented pipe.

  15. Experimental and numerical approach on fracture behaviour of four inches diameter carbon-manganese cracked welded pipes in four point bending

    International Nuclear Information System (INIS)

    Semete, P.; Faidy, C.; Lautier, J.L.

    2001-01-01

    EDF has conducted a research programme to demonstrate the fracture resistance of carbon-manganese welded pipes. The main task of this programme consisted of testing three four inches diameter (114.3 mm O.D.) thin welded pipes (8.56 mm thick) which are representative of those of the sites. The three pipes were loaded under four point bending at a quasi-static rate at -20 C till their maximum bending moment was reached. This paper presents the experimental results, finite element calculations and their comparison with the simplified fracture assessment method of the RSE-M Code. (author)

  16. Fatigue evaluation of piping connections under thermal transients

    International Nuclear Information System (INIS)

    Aquino, C.T.E. de; Maneschy, J.E.

    1993-01-01

    In designing nuclear power plant piping, thermal transients, caused by non-steady operation conditions, should be considered. These events may reduce considerably the lifetime of the pipes, creating the necessity of using structural elements designed in such a way to minimize the acting thermal stresses. Typical examples of the usage of these elements are the connections between pipes of small and large diameters, in which it is usually used a weldolet. Nevertheless, in some situations, the thermal stresses caused by the transients are greater than the allowable limits, being, in this case, an alternative for best results, the introduction of a special fitting replacing the weldolet. Such a fitting is designed in a way to permit a better distribution of the stresses, reducing its maximum value to acceptable levels. This paper intends to present a fatigue evaluation of a connection, using the above mentioned fitting, when subjected to a load expressed in terms of a step thermal gradient, varying from 263 deg to 40 deg C. Two different methodologies are used in this analysis: (a) Determination of the temperature distribution from the heat transfer equations for piping, being the stresses calculated according to ASME III NB-3600. (b) Thermal and stress analyses using axisymmetric elements, according to the rules presented at ASME III NB-3200. In the first case, named simplified analysis, the computer code used is the PIPESTRESS, while in the second case, the ANSYS program was adopted

  17. Effect of the weld groove shape and pass number on residual stresses in butt-welded pipes

    International Nuclear Information System (INIS)

    Sattari-Far, I.; Farahani, M.R.

    2009-01-01

    This study used finite element techniques to analyze the thermo-mechanical behaviour and residual stresses in butt-welded pipes. The residual stresses were also measured in some welds by using the Hole-Drilling method. The results of the finite element analysis were compared with experimentally measured data to evaluate the accuracy of the finite element modelling. Based on this study, a finite element modelling procedure with reasonable accuracy was developed. The developed FE modelling was used to study the effects of weld groove shape and weld pass number on welding residual stresses in butt-welded pipes. The hoop and axial residual stresses in pipe joints of 6 and 10 mm thickness of different groove shapes and pass number were studied. It is shown that these two parameters may have significant effects on magnitude and distribution of residual stresses in welded pipes.

  18. Experimental basis for parameters contributing to energy dissipation in piping systems

    International Nuclear Information System (INIS)

    Ibanez, P.; Ware, A.G.

    1985-01-01

    The paper reviews several pipe testing programs to suggest the phenomena causing energy dissipation in piping systems. Such phenomena include material damping, plasticity, collision in gaps and between pipes, water dynamics, insulation straining, coupling slippage, restraints (snubbers, struts, etc.), and pipe/structure interaction. These observations are supported by a large experimental data base. Data are available from in-situ and laboratory tests (pipe diameters up to about 20 inches, response levels from milli-g's to responses causing yielding, and from excitation wave forms including sinusoid, snapback, random, and seismic). A variety of pipe configurations have been tested, including simple, bare, straight sections and complex lines with bends, snubbers, struts, and insulation. Tests have been performed with and without water and at zero to operating pressure. Both light water reactor and LMFBR piping have been tested

  19. Application of LBB to a nozzle-pipe interface

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Y.J.; Sohn, G.H.; Kim, Y.J. [and others

    1997-04-01

    Typical LBB (Leak-Before-Break) analysis is performed for the highest stress location for each different type of material in the high energy pipe line. In most cases, the highest stress occurs at the nozzle and pipe interface location at the terminal end. The standard finite element analysis approach to calculate J-Integral values at the crack tip utilizes symmetry conditions when modeling near the nozzle as well as away from the nozzle region to minimize the model size and simplify the calculation of J-integral values at the crack tip. A factor of two is typically applied to the J-integral value to account for symmetric conditions. This simplified analysis can lead to conservative results especially for small diameter pipes where the asymmetry of the nozzle-pipe interface is ignored. The stiffness of the residual piping system and non-symmetries of geometry along with different material for the nozzle, safe end and pipe are usually omitted in current LBB methodology. In this paper, the effects of non-symmetries due to geometry and material at the pipe-nozzle interface are presented. Various LBB analyses are performed for a small diameter piping system to evaluate the effect a nozzle has on the J-integral calculation, crack opening area and crack stability. In addition, material differences between the nozzle and pipe are evaluated. Comparison is made between a pipe model and a nozzle-pipe interface model, and a LBB PED (Piping Evaluation Diagram) curve is developed to summarize the results for use by piping designers.

  20. Further considerations for damping in heavily insulated pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

    1985-01-01

    Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Guide 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have how been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design of heavily insulated pipe systems are then recommended

  1. Further considerations for damping in heavily insulated pipe systems

    International Nuclear Information System (INIS)

    Anderson, M.J.; Lindquist, M.R.; Severud, L.K.

    1985-01-01

    Over the past several years a body of test data has been accumulated which demonstrates that damping in small diameter heavily insulated pipe systems is much larger than presently recommended by Regulatory Code 1.61. This data is generally based on pipe systems using a stand-off insulation design with a heater annulus. Additional tests have now been completed on similar pipe systems using a strap-on insulation design without the heater annulus. Results indicate some reduction in damping over the stand-off designs. Test data has also been obtained on a larger sixteen-inch diameter heavily insulated pipe system. Results of these two additional test series are presented. Revised damping values for seismic design for heavily insulated pipe systems are then recommended

  2. Design of a cavity heat pipe receiver experiment

    Science.gov (United States)

    Schneider, Michael G.; Brege, Mark H.; Greenlee, William J.

    1992-01-01

    A cavity heat pipe experiment has been designed to test the critical issues involved with incorporating thermal energy storage canisters into a heat pipe. The experiment is a replication of the operation of a heat receiver for a Brayton solar dynamic power cycle. The heat receiver is composed of a cylindrical receptor surface and an annular heat pipe with thermal energy storage canisters and gaseous working fluid heat exchanger tubes surrounding it. Hardware for the cavity heat pipe experiment will consist of a sector of the heat pipe, complete with gas tube and thermal energy storage canisters. Thermal cycling tests will be performed on the heat pipe sector to simulate the normal energy charge/discharge cycle of the receiver in a spacecraft application.

  3. Pipe damping studies

    International Nuclear Information System (INIS)

    Ware, A.G.

    1986-01-01

    The Idaho National Engineering Laboratory (INEL) is conducting a research program to assist the United States Nuclear Regulatory Commission (USNRC) in determining best-estimate damping values for use in the dynamic analysis of nuclear power plant piping systems. This paper describes four tasks in the program that were undertaken in FY-86. In the first task, tests were conducted on a 5-in. INEL laboratory piping system and data were analyzed from a 6-in. laboratory system at the ANCO Engineers facility to investigate the parameters influencing damping in the seismic frequency range. Further tests were conducted on 3- and 5-in. INEL laboratory piping systems as the second task to determine damping values representative of vibrations in the 33 to 100 Hz range, typical of hydrodynamic transients. In the third task a statistical evaluation of the available damping data was conduted to determine probability distributions suitable for use in probabilistic risk assessments (PRAs), and the final task evaluated damping data at high strain levels

  4. Large-bore pipe decontamination

    International Nuclear Information System (INIS)

    Ebadian, M.A.

    1998-01-01

    The decontamination and decommissioning (D and D) of 1200 buildings within the US Department of Energy-Office of Environmental Management (DOE-EM) Complex will require the disposition of miles of pipe. The disposition of large-bore pipe, in particular, presents difficulties in the area of decontamination and characterization. The pipe is potentially contaminated internally as well as externally. This situation requires a system capable of decontaminating and characterizing both the inside and outside of the pipe. Current decontamination and characterization systems are not designed for application to this geometry, making the direct disposal of piping systems necessary in many cases. The pipe often creates voids in the disposal cell, which requires the pipe to be cut in half or filled with a grout material. These methods are labor intensive and costly to perform on large volumes of pipe. Direct disposal does not take advantage of recycling, which could provide monetary dividends. To facilitate the decontamination and characterization of large-bore piping and thereby reduce the volume of piping required for disposal, a detailed analysis will be conducted to document the pipe remediation problem set; determine potential technologies to solve this remediation problem set; design and laboratory test potential decontamination and characterization technologies; fabricate a prototype system; provide a cost-benefit analysis of the proposed system; and transfer the technology to industry. This report summarizes the activities performed during fiscal year 1997 and describes the planned activities for fiscal year 1998. Accomplishments for FY97 include the development of the applicable and relevant and appropriate regulations, the screening of decontamination and characterization technologies, and the selection and initial design of the decontamination system

  5. Instability predictions for circumferentially cracked Type-304 stainless steel pipes under dynamic loading. Volume 2. Appendixes. Final report. [BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zahoor, A.; Wilkowski, G.; Abou-Sayed, I.; Marschall, C.; Broek, D.; Sampath, S.; Rhee, H.; Ahmad, J.

    1982-04-01

    This report provides methods to predict margins of safety for circumferentially cracked Type 304 stainless steel pipes subjected to applied bending loads. An integrated combination of experimentation and analysis research was pursued. Two types of experiments were performed: (1) laboratory-scale tests on center-cracked panels and bend specimens to establish the basic mechanical and fracture properties of Type 304 stainless steel, and (2) full-scale pipe fracture tests under quasi-static and dynamic loadings to assess the analysis procedures. Analyses were based upon the simple plastic collapse criterion, a J-estimation procedure, and elastic-plastic large-deformation finite element models.

  6. Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in 'like-new' conditions

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1994-09-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the 'like-new' condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed

  7. Evaluation of thermal aging effect on primary pipe material in nuclear power plant by micro hardness test method

    International Nuclear Information System (INIS)

    Xue Fei; Yu Weiwei; Wang Zhaoxi; Ma Qinzheng; Liu Wei

    2012-01-01

    The investigation was carried out on the changes in mechanical properties of the primary pipe material Z3CN20.09M after 10000 h aging at 400℃ by using micro- Vickers and impact testing machine. The results show that the impact energy of testing material decreases. However, the micro-Vickers hardness of ferrite phase and austenite phase which constitute the testing material increase and keep constant, respectively. The intrinsic relations were analyzed between the micro-Vickers hardness and the impact energy to make an attempt to present the micro-Vickers hardness measurement as a method applicable to evaluating the thermal aging of the primary pipe material. (authors)

  8. Experimental results from containment piping bellows subjected to severe accident conditions: Results from bellows tested in corroded conditions. Volume 2

    International Nuclear Information System (INIS)

    Lambert, L.D.; Parks, M.B.

    1995-10-01

    Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of nineteen bellows have been tested. Thirteen bellows were tested in ''like-new'' condition (results reported in Volume 1), and six were tested in a corroded condition. The tests showed that bellows in ''like-new'' condition are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage, while those in a corroded condition did not perform as well, depending on the amount of corrosion. The corroded bellows test program and results are presented in this report

  9. Preliminary piping layout and integration of European test blanket modules subsystems in ITER CVCS area

    Energy Technology Data Exchange (ETDEWEB)

    Tarallo, Andrea, E-mail: andrea.tarallo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Mozzillo, Rocco; Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio, 80, 80125 Naples (Italy); Aiello, Antonio; Utili, Marco [ENEA UTIS, C.R. Brasimone, Bacino del Brasimone, I-40032 Camugnano, BO (Italy); Ricapito, Italo [TBM& MD Project, Fusion for Energy, EU Commission, Carrer J. Pla, 2, Building B3, 08019 Barcelona (Spain)

    2015-04-15

    Highlights: • The use of human modeling tools for piping design in view of maintenance is discussed. • A possible preliminary layout for TBM subsystems in CVCS area has been designed with CATIA. • A DHM-based method to quickly check for maintainability of piping systems is suggested. - Abstract: This paper explores a possible integration of some ancillary systems of helium-cooled lithium lead (HCLL) and helium-cooled pebble-bed (HCPB) test blanket modules in ITER CVCS area. Computer-aided design and ergonomics simulation tools have been fundamental not only to define suitable routes for pipes, but also to quickly check for maintainability of equipment and in-line components. In particular, accessibility of equipment and systems has been investigated from the very first stages of the design using digital human models. In some cases, the digital simulations have resulted in changes in the initial space reservations.

  10. Fuel element for a nuclear reactor

    International Nuclear Information System (INIS)

    Tanihiro, Yasunori; Sumita, Isao.

    1970-01-01

    An improved fuel element of the heat pipe type is disclosed in which the fuel element itself is given a heat pipe structure and filled with a coated particle fuel at the section thereof having a capillary tube construction, whereby the particular advantages of heat pipes and coated fuels are combined and utilized to enhance thermal control and reactor efficiency. In an embodiment, the fuel element of the present invention is filled at its lower capillary tube section with coated fuel and at its upper section with a granurated neutron absorber. Both sections are partitioned from the central shaft by a cylindrically shaped wire mesh defining a channel through which the working liquid is vaporized from below and condensed by the coolant external to the fuel element. If the wire mesh is chosen to have a melting point lower than that of the fuel but higher than that of the operating temperature of the heat pipe, the mesh will melt and release the neutron absorbing particles should hot spots develop, thus terminating fission. (Owens, K. J.)

  11. Study on unstable fracture characteristics of light water reactor piping

    International Nuclear Information System (INIS)

    Kurihara, Ryoichi

    1998-08-01

    Many testing studies have been conducted to validate the applicability of the leak before break (LBB) concept for the light water reactor piping in the world. It is especially important among them to clarify the condition that an inside surface crack of the piping wall does not cause an unstable fracture but ends in a stable fracture propagating only in the pipe thickness direction, even if the excessive loading works to the pipe. Pipe unstable fracture tests performed in Japan Atomic Energy Research Institute had been planned under such background, and clarified the condition for the cracked pipe to cause the unstable fracture under monotonous increase loading or cyclic loading by using test pipes with the inside circumferential surface crack. This paper examines the pipe unstable fracture by dividing it into two parts. One is the static unstable fracture that breaks the pipe with the inside circumferential surface crack by increasing load monotonously. Another is the dynamic unstable fracture that breaks the pipe by the cyclic loading. (author). 79 refs

  12. Finned Carbon-Carbon Heat Pipe with Potassium Working Fluid

    Science.gov (United States)

    Juhasz, Albert J.

    2010-01-01

    This elemental space radiator heat pipe is designed to operate in the 700 to 875 K temperature range. It consists of a C-C (carbon-carbon) shell made from poly-acrylonitride fibers that are woven in an angle interlock pattern and densified with pitch at high process temperature with integrally woven fins. The fins are 2.5 cm long and 1 mm thick, and provide an extended radiating surface at the colder condenser section of the heat pipe. The weave pattern features a continuous fiber bath from the inner tube surface to the outside edges of the fins to maximize the thermal conductance, and to thus minimize the temperature drop at the condenser end. The heat pipe and radiator element together are less than one-third the mass of conventional heat pipes of the same heat rejection surface area. To prevent the molten potassium working fluid from eroding the C C heat pipe wall, the shell is lined with a thin-walled, metallic tube liner (Nb-1 wt.% Zr), which is an integral part of a hermetic metal subassembly which is furnace-brazed to the inner surface of the C-C tube. The hermetic metal liner subassembly includes end caps and fill tubes fabricated from the same Nb-1Zr alloy. A combination of laser and electron beam methods is used to weld the end caps and fill tubes. A tungsten/inert gas weld seals the fill tubes after cleaning and charging the heat pipes with potassium. The external section of this liner, which was formed by a "Uniscan" rolling process, transitions to a larger wall thickness. This section, which protrudes beyond the C-C shell, constitutes the "evaporator" part of the heat pipe, while the section inside the shell constitutes the condenser of the heat pipe (see figure).

  13. Heat Pipes

    Science.gov (United States)

    1990-01-01

    Bobs Candies, Inc. produces some 24 million pounds of candy a year, much of it 'Christmas candy.' To meet Christmas demand, it must produce year-round. Thousands of cases of candy must be stored a good part of the year in two huge warehouses. The candy is very sensitive to temperature. The warehouses must be maintained at temperatures of 78-80 degrees Fahrenheit with relative humidities of 38- 42 percent. Such precise climate control of enormous buildings can be very expensive. In 1985, energy costs for the single warehouse ran to more than 57,000 for the year. NASA and the Florida Solar Energy Center (FSEC) were adapting heat pipe technology to control humidity in building environments. The heat pipes handle the jobs of precooling and reheating without using energy. The company contacted a FSEC systems engineer and from that contact eventually emerged a cooperative test project to install a heat pipe system at Bobs' warehouses, operate it for a period of time to determine accurately the cost benefits, and gather data applicable to development of future heat pipe systems. Installation was completed in mid-1987 and data collection is still in progress. In 1989, total energy cost for two warehouses, with the heat pipes complementing the air conditioning system was 28,706, and that figures out to a cost reduction.

  14. Load-deflection characteristics of small bore insulated pipe clamps

    International Nuclear Information System (INIS)

    Severud, L.K.; Clark, G.L.

    1982-01-01

    High temperature LMFBR piping is subject to rapid temperature changes during transient events. Typically, this pipe is supported by specially designed insulated pipe clamps to prevent excessive thermal stress from developing during these transients. The special insulated clamps used on both FFTF and CRBR piping utilize a Belleville spring arrangement to compensate for pipe thermal expansion. Analysis indicates that this produces a non-linear, directionally sensitive clamp spring rate. Since these spring rates influence the seismic response of a supported piping system, it was deemed necessary to evaluate them further by test. This has been accomplished for the FFTF clamps. A more standard insulated pipe clamp, which does not incorporate Belleville springs to accommodate thermal expansion, was also tested. This type clamp is simple in design, and economically attractive. It may have wide application prospects for use in LMFBR small bore auxiliary piping operating at temperatures below 427 0 C. Load deflection tests were conducted on 2.54 CM and 7.62 CM diameter samples of these commercial clamps

  15. Participation of the GRS in the 'Degraded piping program' of the USNRC

    International Nuclear Information System (INIS)

    Azodi, D.; Hoefler, A.; Sievers, J.

    1989-01-01

    During the course of the Degraded Piping Program, Battelle was the organizer of three round-robin activities for advanced elasto-plastic fracture mechanics calculations (e.g. finite element method and J-estimation scheme). GRS participated in all of them and submitted finite element and J-estimation scheme results. As a main result of round-robin on elasto-plastic fracture mechanics calculations: Based on finite element calculations, the J-integral method (energy release rate) provided the ability to describe the fracture behaviour of flawed piping even in a very ductile material. (orig./HP) [de

  16. Low Cost High Performance Generator Technology Program. Volume 5. Heat pipe topical, appendices

    International Nuclear Information System (INIS)

    1975-07-01

    Work performed by Dynatherm Corporation for Teledyne Isotopes during a program entitled ''Heat Pipe Fabrication, Associated Technical Support and Reporting'' is reported. The program was initiated on November 29, 1972; the main objectives were accomplished with the delivery of the heat pipes for the HPG. Life testing of selected heat pipe specimens is continuing to and beyond the present date. The program consisted of the following tasks: Heat Pipe Development of Process Definition; Prototype Heat Pipes for Fin Segment Test; HPG Heat Pipe Fabrication and Testing; Controlled Heat Pipe Life Test; and Heat Pipe Film Coefficient Determination

  17. Pile load test on large diameter steel pipe piles in Timan-Pechora, Russia

    Energy Technology Data Exchange (ETDEWEB)

    McKeown, S. [Golder Associates Inc., Houston, TX (United States); Tart, B. [Golder Associates Inc., Anchorage, AK (United States); Swartz, R. [Paragon Engineering Services Inc., Houston, TX (United States)

    1994-12-31

    Pile load testing conducted in May and June of 1993 at the Polar Lights Ardalin project in Arkangelsk province, Russia, was documented. Pile load testing was conducted to determine the ultimate and allowable pile loads for varying pile lengths and ground temperature conditions and to provide creep test data for deformation under constant load. The piles consisted of 20 inch diameter steel pipe piles driven open ended through prebored holes into the permafrost soils. Ultimate pile capacities, adfreeze bond, and creep properties observed were discussed. 10 figs., 4 tabs.

  18. Thinned pipe management program of Korean NPPs

    International Nuclear Information System (INIS)

    Lee, S.H.; Kim, T.R.; Jeon, S.C.; Hwang, K.M.

    2003-01-01

    Wall thinning of carbon steel pipe components due to Flow-Accelerated Corrosion (FAC) is one of the most serious threats to the integrity of steam cycle systems in Nuclear Power Plants (NPP). If the thickness of a pipe component is reduced below the critical level, it cannot sustain stress and consequently results in leakage or rupture. In order to minimize the possibility of excessive wall thinning, Thinned Pipe Management Program (TPMP) has been set up and being implemented to all Korean NPPs. Important elements of the TPMP include the prediction of the FAC rate for each component based on model analysis, prioritization of pipe components for inspection, thickness measurement, calculation of wear and wear rate for each component. Additionally, decision making associated with replacement or continuous service for thinned pipe components and establishment of long-term strategic management plan based on diagnosis of plant condition regarding overall wall thinning also are essential part of the TPMP. From pre-service inspection data, it has been found that initial thickness is varies, which influences wear and wear rate calculations. (author)

  19. Heat pipe development

    Science.gov (United States)

    Bienart, W. B.

    1973-01-01

    The objective of this program was to investigate analytically and experimentally the performance of heat pipes with composite wicks--specifically, those having pedestal arteries and screwthread circumferential grooves. An analytical model was developed to describe the effects of screwthreads and screen secondary wicks on the transport capability of the artery. The model describes the hydrodynamics of the circumferential flow in triangular grooves with azimuthally varying capillary menisci and liquid cross-sections. Normalized results were obtained which give the influence of evaporator heat flux on the axial heat transport capability of the arterial wick. In order to evaluate the priming behavior of composite wicks under actual load conditions, an 'inverted' glass heat pipe was designed and constructed. The results obtained from the analysis and from the tests with the glass heat pipe were applied to the OAO-C Level 5 heat pipe, and an improved correlation between predicted and measured evaporator and transport performance were obtained.

  20. Structural analyses on piping systems of sodium reactors. 2. Eigenvalue analyses of hot-leg pipelines of large scale sodium reactors

    International Nuclear Information System (INIS)

    Furuhashi, Ichiro; Kasahara, Naoto

    2002-01-01

    Two types of finite element models analyzed eigenvalues of hot-leg pipelines of a large-scale sodium reactor. One is a beam element model, which is usual for pipe analyses. The other is a shell element model to evaluate particular modes in thin pipes with large diameters. Summary of analysis results: (1) A beam element model and a order natural frequency. A beam element model is available to get the first order vibration mode. (2) The maximum difference ratio of beam mode natural frequencies was 14% between a beam element model with no shear deformations and a shell element model. However, its difference becomes very small, when shear deformations are considered in beam element. (3) In the first order horizontal mode, the Y-piece acts like a pendulum, and the elbow acts like the hinge. The natural frequency is strongly affected by the bending and shear rigidities of the outer supporting pipe. (4) In the first order vertical mode, the vertical sections of the outer and inner pipes moves in the axial-directional piston mode, the horizontal section of inner pipe behaves like the cantilever, and the elbow acts like the hinge. The natural frequency is strongly affected by the axial rigidity of outer supporting pipe. (5) Both effective masses and participation factors were small for particular shell modes. (author)

  1. Task force activity to take the effect of elastic-plastic behaviour into account on the seismic safety evaluation of nuclear piping systems

    International Nuclear Information System (INIS)

    Nakamura, Izumi; Shiratori, Masaki; Morishita, Masaki; Otani, Akihito; Shibutani, Tadahito

    2015-01-01

    According to investigations of several nuclear power plants (NPPs) hit by actual seismic events and a number of experimental researches on the failure behavior of piping systems under seismic loads, it is recognized that piping systems used in NPPs include a large seismic safety margin until boundary failure. Since the stress assessment based on the elastic analysis does not reflect actual seismic capability of piping systems including plastic region, it is necessary to develop a rational procedures to estimate the elastic-plastic behavior of piping systems under a large seismic load. With the aim of establishing a procedure that takes into account the elastic-plastic behavior effect in the seismic safety estimation of nuclear piping systems, a task force activity has been planned. Through the activity, the authors intend to establish guidelines to estimate the elastic-plastic behavior of piping systems rationally and conservatively, and to provide new rational seismic safety criteria taking the effect of elastic-plastic behavior into account. As the first step of making out the analysis guideline, benchmark analyses are conducted for a pipe element test and a piping system test. In this paper, the outline of the research activity and the preliminary results of benchmark analyses are described. (author)

  2. Application of 1 D Finite Element Method in Combination with Laminar Solution Method for Pipe Network Analysis

    Science.gov (United States)

    Dudar, O. I.; Dudar, E. S.

    2017-11-01

    The features of application of the 1D dimensional finite element method (FEM) in combination with the laminar solutions method (LSM) for the calculation of underground ventilating networks are considered. In this case the processes of heat and mass transfer change the properties of a fluid (binary vapour-air mix). Under the action of gravitational forces it leads to such phenomena as natural draft, local circulation, etc. The FEM relations considering the action of gravity, the mass conservation law, the dependence of vapour-air mix properties on the thermodynamic parameters are derived so that it allows one to model the mentioned phenomena. The analogy of the elastic and plastic rod deformation processes to the processes of laminar and turbulent flow in a pipe is described. Owing to this analogy, the guaranteed convergence of the elastic solutions method for the materials of plastic type means the guaranteed convergence of the LSM for any regime of a turbulent flow in a rough pipe. By means of numerical experiments the convergence rate of the FEM - LSM is investigated. This convergence rate appeared much higher than the convergence rate of the Cross - Andriyashev method. Data of other authors on the convergence rate comparison for the finite element method, the Newton method and the method of gradient are provided. These data allow one to conclude that the FEM in combination with the LSM is one of the most effective methods of calculation of hydraulic and ventilating networks. The FEM - LSM has been used for creation of the research application programme package “MineClimate” allowing to calculate the microclimate parameters in the underground ventilating networks.

  3. Study on heat pipe assisted thermoelectric power generation system from exhaust gas

    Science.gov (United States)

    Chi, Ri-Guang; Park, Jong-Chan; Rhi, Seok-Ho; Lee, Kye-Bock

    2017-11-01

    Currently, most fuel consumed by vehicles is released to the environment as thermal energy through the exhaust pipe. Environmentally friendly vehicle technology needs new methods to increase the recycling efficiency of waste exhaust thermal energy. The present study investigated how to improve the maximum power output of a TEG (Thermoelectric generator) system assisted with a heat pipe. Conventionally, the driving energy efficiency of an internal combustion engine is approximately less than 35%. TEG with Seebeck elements is a new idea for recycling waste exhaust heat energy. The TEG system can efficiently utilize low temperature waste heat, such as industrial waste heat and solar energy. In addition, the heat pipe can transfer heat from the automobile's exhaust gas to a TEG. To improve the efficiency of the thermal power generation system with a heat pipe, effects of various parameters, such as inclination angle, charged amount of the heat pipe, condenser temperature, and size of the TEM (thermoelectric element), were investigated. Experimental studies, CFD simulation, and the theoretical approach to thermoelectric modules were carried out, and the TEG system with heat pipe (15-20% charged, 20°-30° inclined configuration) showed the best performance.

  4. Crack propagation and arrest simulation of X90 gas pipe

    International Nuclear Information System (INIS)

    Yang, Fengping; Huo, Chunyong; Luo, Jinheng; Li, He; Li, Yang

    2017-01-01

    To determine whether X90 steel pipe has enough crack arrest toughness or not, a damage model was suggested as crack arrest criterion with material parameters of plastic uniform percentage elongation and damage strain energy per volume. Fracture characteristic length which characterizes fracture zone size was suggested to be the largest mesh size on expected cracking path. Plastic uniform percentage elongation, damage strain energy per volume and fracture characteristic length of X90 were obtained by five kinds of tensile tests. Based on this criterion, a length of 24 m, Φ1219 × 16.3 mm pipe segment model with 12 MPa internal gas pressure was built and computed with fluid-structure coupling method in ABAQUS. Ideal gas state equation was used to describe lean gas behavior. Euler grid was used to mesh gas zone inside the pipe while Lagrangian shell element was used to mesh pipe. Crack propagation speed and gas decompression speed were got after computation. The result shows that, when plastic uniform percentage elongation is equal to 0.054 and damage strain energy per volume is equal to 0.64 J/mm"3, crack propagation speed is less than gas decompression speed, which means the simulated X90 gas pipe with 12 MPa internal pressure can arrest cracking itself. - Highlights: • A damage model was suggested as crack arrest criterion. • Plastic uniform elongation and damage strain energy density are material parameters. • Fracture characteristic length is suggested to be largest mesh size in cracking path. • Crack propagating simulation with coupling of pipe and gas was realized in ABAQUS. • A Chinese X90 steel pipe with 12 MPa internal pressure can arrest cracking itself.

  5. Vortex induced vibrations in gapped restrainted pipes

    International Nuclear Information System (INIS)

    Veloso, P. de A.A.; Loula, A.F.D.

    1984-01-01

    The vortex induced vibration problem of gapped restrained piping is solved numerically. The model proposed by Skop-Griffin is used to describe the pipe-fluid interaction. The variational formulation is obtained modeling the gapped restraints as non-linear elastic springs. The regularized problem is solved using a finite element discretization for the spatial domain. In the time domain a finite difference discretization is used for the lift coefficient equatin and a Newmark discretization for the equation of motion. (Author) [pt

  6. Numerical simulation of water hammer in low pressurized pipe: comparison of SimHydraulics and Lax-Wendroff method with experiment

    Science.gov (United States)

    Himr, D.

    2013-04-01

    Article describes simulation of unsteady flow during water hammer with two programs, which use different numerical approaches to solve ordinary one dimensional differential equations describing the dynamics of hydraulic elements and pipes. First one is Matlab-Simulink-SimHydraulics, which is a commercial software developed to solve the dynamics of general hydraulic systems. It defines them with block elements. The other software is called HYDRA and it is based on the Lax-Wendrff numerical method, which serves as a tool to solve the momentum and continuity equations. This program was developed in Matlab by Brno University of Technology. Experimental measurements were performed on a simple test rig, which consists of an elastic pipe with strong damping connecting two reservoirs. Water hammer is induced with fast closing the valve. Physical properties of liquid and pipe elasticity parameters were considered in both simulations, which are in very good agreement and differences in comparison with experimental data are minimal.

  7. Numerical simulation of water hammer in low pressurized pipe: comparison of SimHydraulics and Lax-Wendroff method with experiment

    Directory of Open Access Journals (Sweden)

    Himr D.

    2013-04-01

    Full Text Available Article describes simulation of unsteady flow during water hammer with two programs, which use different numerical approaches to solve ordinary one dimensional differential equations describing the dynamics of hydraulic elements and pipes. First one is Matlab-Simulink-SimHydraulics, which is a commercial software developed to solve the dynamics of general hydraulic systems. It defines them with block elements. The other software is called HYDRA and it is based on the Lax-Wendrff numerical method, which serves as a tool to solve the momentum and continuity equations. This program was developed in Matlab by Brno University of Technology. Experimental measurements were performed on a simple test rig, which consists of an elastic pipe with strong damping connecting two reservoirs. Water hammer is induced with fast closing the valve. Physical properties of liquid and pipe elasticity parameters were considered in both simulations, which are in very good agreement and differences in comparison with experimental data are minimal.

  8. Development of a software for the ASME code qualification of class-I nuclear piping systems

    International Nuclear Information System (INIS)

    Mishra, Rajesh; Umashankar, C.; Soni, R.S.; Kushwaha, H.S.; Venkat Raj, V.

    1999-11-01

    In nuclear industry, the designer often comes across the requirements of Class-1 piping systems which need to be qualified for various normal and abnormal loading conditions. In order to have quick design changes and the design reviews at various stages of design, it is quite helpful if a dedicated software is available for the qualification of Class-1 piping systems. BARC has already purchased a piping analysis software CAESAR-II and has used it for the life extension of heavy water plant, Kota. CAESAR-II facilitates the qualification of Class-2 and Class-3 piping systems among others. However, the present version of CAESAR-II does not have the capability to perform stress checks for the ASME Class-1 nuclear piping systems. With this requirement in mind and the prohibitive costs of commercially available software for the Class-1 piping analyses, it was decided to develop a separate software for this class of piping in such a way that the input and output details of the piping from the CAESAR-II software can be made use of. This report principally contains the details regarding development of a software for codal qualification of Class-1 nuclear piping as per ASME code section-III, NB-3600. The entire work was carried out in three phases. The first phase consisted of development of the routines for reading the output files obtained from the CAESAR-II software, and converting them into required format for further processing. In this phase, the nodewise informations available from the CAESAR-II output file were converted into element-wise informations. The second phase was to develop a general subroutine for reading the various input parameters such as diameter, wall thickness, corrosion allowance, bend radius and also to recognize the bend elements based on the bend radius, directly from the input file of CAESAR-II software. The third phase was regarding the incorporation of the required steps for performing the ASME codal checks as per NB-3600 for Class-1 piping

  9. The influences of soil and nearby structures on dispersion characteristics of wave propagating along buried plastic pipes

    Science.gov (United States)

    Liu, Shuyong; Jiang, J.; Parr, Nicola

    2016-09-01

    Water loss in distribution systems is a global problem for the water industry and governments. According to the international water supply association (IWSA), as a result of leaks from distribution pipes, 20% to 30% of water is lost while in transit from treatment plants to consumers. Although governments have tried to push the water industry to reduce the water leaks, a lot of experts have pointed out that a wide use of plastic pipes instead of metal pipes in recent years has caused difficulties in the detection of leaks using current acoustic technology. Leaks from plastic pipes are much quieter than traditional metal pipes and comparing to metal pipes the plastic pipes have very different coupling characteristics with soil, water and surrounding structures, such as other pipes, road surface and building foundations. The dispersion characteristics of wave propagating along buried plastic pipes are investigated in this paper using finite element and boundary element based models. Both empty and water- filled pipes were considered. Influences from nearby pipes and building foundations were carefully studied. The results showed that soil condition and nearby structures have significant influences on the dispersion characteristics of wave propagating along buried plastic pipes.

  10. Effect of cross section on collapse load in pipe bends subjected to in ...

    African Journals Online (AJOL)

    user

    Also various researchers have estimated the plastic loads of pipe bends with cracks (Hong et al, ... In reality, the pipe bend exists with shape imperfections namely ovality and ... C t. −. = ×. (3). 3. Finite element limit analysis. Figure 1. Pipe bend with ..... Chattopadhyay J., Natahani D. K., Dutta B. K. and Kushwaha H. S. 2000.

  11. MARVIKEN-CFT, Marviken Full Scale Critical Flow Tests

    International Nuclear Information System (INIS)

    1998-01-01

    1 - Description of test facility: The major components of the facility are: - The pressure vessel with a net volume = 425 m 3 , a maximum design pressure = 5.75 MPa and a maximum design temperature = 545 K. - The discharge pipe attached to the pressure vessel bottom. The discharge pipe consists of a ball valve and pipe spools which house the instrumentation upstream of the test nozzle. - The test nozzles and rupture disc assemblies. A set of test nozzles of specified lengths and diameters (tubular section lengths from 0.166 to 1.809 m and tubular section diameters from 0.2 to 0.509 m) were used to which the rupture disc assemblies were attached. - The containment and exhaust pipes, consisting of the drywell, the wet-well, fuel element transport hall, the ground level exhaust pipe, and the upper exhaust pipe. 2 - Description of the tests: Before each test the vessel was filled with degassed water up to a certain level which varied between the tests (from about 16.5 to 20 m above vessel bottom). The pressure was around 5 MPa (one test around 4 MPa). A pre-test warm-up period produced a temperature profile along the vessel height. After a stabilizing period of several hours, the test was initiated by failing the discs contained in the rupture disc assembly. Measurements recorded in the vessel, the discharge pipe, and the test nozzle, provided the data, while the vessel fluid was discharged through the test nozzle into the containment and further through the exhaust pipes to the ambient atmosphere. The test was terminated when the ball valve begun to close or when pure steam was entered the discharge pipe. 3 - Experimental limitations or shortcomings: Initial maximum pressure was limited to about 5 MPa

  12. Fabrication of lithium/C-103 alloy heat pipes for sharp leading edge cooling

    Science.gov (United States)

    Ai, Bangcheng; Chen, Siyuan; Yu, Jijun; Lu, Qin; Han, Hantao; Hu, Longfei

    2018-05-01

    In this study, lithium/C-103 alloys heat pipes are proposed for sharp leading edge cooling. Three models of lithium/C-103 alloy heat pipes were fabricated. And their startup properties were tested by radiant heat tests and aerothermal tests. It is found that the startup temperature of lithium heat pipe was about 860 °C. At 1000 °C radiant heat tests, the operating temperature of lithium/C-103 alloy heat pipe is lower than 860 °C. Thus, startup failure occurs. At 1100 °C radiant heat tests and aerothermal tests, the operating temperature of lithium/C-103 alloy heat pipe is higher than 860 °C, and the heat pipe starts up successfully. The startup of lithium/C-103 alloy heat pipe decreases the leading edge temperature effectively, which endows itself good ablation resistance. After radiant heat tests and aerothermal tests, all the heat pipe models are severely oxidized because of the C-103 poor oxidation resistance. Therefore, protective coatings are required for further applications of lithium/C-103 alloy heat pipes.

  13. Nitrogen heat pipe for cryocooler thermal shunt

    International Nuclear Information System (INIS)

    Prenger F.C.; Hill, D.D.; Daney, D.E.

    1996-01-01

    A nitrogen heat pipe was designed, built and tested for the purpose of providing a thermal shunt between the two stages of a Gifford-McMahan (GM) cryocooler during cooldown. The nitrogen heat pipe has an operating temperature range between 63 and 123 K. While the heat pipe is in this temperature range during the system cooldown, it acts as a thermal shunt between the first and second stage of the cryocooler. The heat pipe increases the heat transfer to the first stage of the cryocooler, thereby reducing the cooldown time of the system. When the heat pipe temperature drops below the triple point, the nitrogen working fluid freezes, effectively stopping the heat pipe operation. A small heat leak between cryocooler stages remains because of axial conduction along the heat pipe wall. As long as the heat pipe remains below 63 K, the heat pipe remains inactive. Heat pipe performance limits were measured and the optimum fluid charge was determined

  14. PPOOLEX experiments with two parallel blowdown pipes

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2011-01-15

    This report summarizes the results of the experiments with two transparent blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through either one or two vertical transparent blowdown pipes to the condensation pool. Five experiments with one pipe and six with two parallel pipes were carried out. The main purpose of the experiments was to study loads caused by chugging (rapid condensation) while steam is discharged into the condensation pool filled with sub-cooled water. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. In the experiments the initial temperature of the condensation pool water varied from 12 deg. C to 55 deg. C, the steam flow rate from 40 g/s to 1 300 g/s and the temperature of incoming steam from 120 deg. C to 185 deg. C. In the experiments with only one transparent blowdown pipe chugging phenomenon didn't occur as intensified as in the preceding experiments carried out with a DN200 stainless steel pipe. With the steel blowdown pipe even 10 times higher pressure pulses were registered inside the pipe. Meanwhile, loads registered in the pool didn't indicate significant differences between the steel and polycarbonate pipe experiments. In the experiments with two transparent blowdown pipes, the steamwater interface moved almost synchronously up and down inside both pipes. Chugging was stronger than in the one pipe experiments and even two times higher loads were measured inside the pipes. The loads at the blowdown pipe outlet were approximately the same as in the one pipe cases. Other registered loads around the pool were about 50-100 % higher than with one pipe. The experiments with two parallel blowdown pipes gave contradictory results compared to the earlier studies dealing with chugging loads in case of multiple pipes. Contributing

  15. Crack initiation through vibration fatigue of small-diameter pipes

    International Nuclear Information System (INIS)

    Comby, R.; Thebault, Y.; Papaconstantinou, T.

    2002-01-01

    Socket welds are used extensively for small bore piping connections in nuclear power plant systems. Numerous fatigue-related failures occurred in the past ten years mainly on safeguard systems and continue to occur frequently, showing that corrective actions did not take into account all aspects of the problem. Destructive examination of cracked small bore piping connections allowed a better understanding of failure mechanisms and a prediction of crack initiation site depending on nozzle fittings such as run pipe and small bore pipe thickness. A three-dimensional finite element model confirmed the conclusions of the lab examinations. For thick run pipes, it was shown that the failure tend to initiate predominantly at the socket weld toe or at the root, depending on the respective thickness of coupling and small bore pipe. Some additional studies, based on RSE-M code, are in progress in order to determine the maximum stresses location. Lessons learned through these investigations led to optimise the in-service inspection scope and to define solutions to be carried out to prevent failure of ''susceptible'' small bore pipe connections. Since July 2000, a large program is in progress to select all ''susceptible'' small bore pipes in safety-related systems and to apply corrective measures such as piping modifications or system operational modifications. (authors)

  16. Effect on mechanical properties of glass reinforced epoxy (GRE) pipe filled with different geopolymer filler molarity for piping application

    Science.gov (United States)

    Hashim, M. F. Abu; Abdullah, M. M. A.; Ghazali, C. M. R.; Hussin, K.; Binhussain, M.

    2017-04-01

    This study investigated the use of a novel white clay geopolymer as a filler to produce high strength glass reinforced epoxy pipe. It was found that using white clay geopolymer as filler gives better compressive strength to the glass reinforced epoxy pipe. The disadvantages of current glass reinforced epoxy pipes such low compressive strength which can be replaced by the composite pipes. Geopolymerization is an innovative technology that can transform several aluminosilicate materials into useful products called geopolymers or inorganic polymers. A series of glass reinforced epoxy pipe and glass reinforced epoxy pipe filled with 10 - 40 weight percentages white clay geopolymer filler with 4 Molarity and 8 Molarity were prepared. Morphology of white clay geopolymer filler surface was indicates using scanning electron microscopy. The additions of white clay geopolymer filler for both 4 Molarity and 8 Molarity show higher compressive strength than glass reinforced epoxy pipe without any geopolymer filler. The compressive test of these epoxy geopolymer pipe samples was determined using Instron Universal Testing under compression mode. Nonetheless, the compressive strength of glass reinforced epoxy pipe with white clay geopolymer filler continues to drop when added to 40 wt% of the geopolymer filler loading for both 4 Molarity and 8 Molarity. These outcomes showed that the mixing of geopolymer materials in epoxy system can be attained in this research.

  17. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1986-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-in. and a pressurized 6-in. diameter carbon steel nuclear pipe systems subjected to high level shaking have been accomplished. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occurred in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate very well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules may be appropriate to cover the ratchet-fatigue failure mode

  18. Pre-irradiation testing of experimental fuel elements

    International Nuclear Information System (INIS)

    Basova, B.G.; Davydov, E.F.; Dvoretskij, V.G.; Ivanov, V.B.; Syuzev, V.N.; Timofeev, G.A.; Tsykanov, V.A.

    1979-01-01

    The problems of testing of experimental fuel elements of nuclear reactors on the basis of complex accountancy of the factors defining operating capacity of the fuel elements are considered. The classification of the parameters under control and the methods of initial technological testing, including testing of the fuel product, cladding and fished fuel element, is given. The requirements to the apparatus used for complex testing are formulated. One of the possible variants of representation of the information obtained in the form of the input certificate of a single fuel element under study is proposed. The processing flowsheet of the gathered information using the computer is given. The approach under consideration is a methodological basis of investigation of fuel element operating life at the testing stage of the experimental fuel elements

  19. Testing in support of transportation of residues in the pipe overpack container

    International Nuclear Information System (INIS)

    Ammerman, D.J.; Bobbe, J.G.; Arviso, M.; Bronowski, D.R.

    1997-04-01

    The disposition of the large back-log of plutonium residues at the Rocky Flats Environmental Technology Site (Rocky Flats) will require interim storage and subsequent shipment to a waste repository. Current plants call for disposal at the Waste Isolation Pilot Plant (WIPP) and the transportation to WIPP in the TRUPACT-II. The transportation phase will require the residues to be packaged in a container that is more robust than a standard 55-gallon waste drum. Rocky Flats has designed the Pipe Overpack Container to meet this need. The tests described here were performed to qualify the Pipe Overpack Container as a waste container for shipment in the TRUPACT-II. Using a more robust container will assure the fissile materials in each container can not be mixed with the fissile material from the other containers and will provide criticality control. This will allow an increase in the payload of the TRUPACT-II from 325 fissile gram equivalents to 2,800 fissile gram equivalents

  20. Criteria for accepting piping vibrations measured during FFTF plant startup

    International Nuclear Information System (INIS)

    Huang, S.N.

    1981-03-01

    Piping in the Fast Flux Test Facility is subjected to low-amplitude, high cycle vibration over the plant lifetime. Excitation sources include the mechanical vibration induced by main centrifugal pumps, auxiliary reciprocating pumps, EM pumps and possible flow oscillations. Vibration acceptance criteria must be established which will prevent excessive pipe and support fatigue damage when satified. This paper describes the preparation of such criteria against pipe failure used for acceptance testing of the Fast Flux Test Facility main heat transport piping

  1. Crack stability in a representative piping system under combined inertial and seismic/dynamic displacement-controlled stresses. Subtask 1.3 final report

    International Nuclear Information System (INIS)

    Scott, P.; Olson, R.; Wilkowski, O.G.; Marschall, C.; Schmidt, R.

    1997-06-01

    This report presents the results from Subtask 1.3 of the International Piping Integrity Research Group (IPIRG) program. The objective of Subtask 1.3 is to develop data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system under combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The piping system evaluated is an expansion loop with over 30 meters of 16-inch diameter Schedule 100 pipe. The experimental facility is equipped with special hardware to ensure system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe-system experiments. The uncracked experiment was conducted to evaluate piping system damping and natural frequency characteristics. The cracked-pipe experiments evaluated the fracture behavior, pipe system response, and stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Results from all pipe-system experiments and material characterization efforts are presented. Results of fracture mechanics analyses, dynamic finite element stress analyses, and stability analyses are presented and compared with experimental results

  2. Crack stability in a representative piping system under combined inertial and seismic/dynamic displacement-controlled stresses. Subtask 1.3 final report

    Energy Technology Data Exchange (ETDEWEB)

    Scott, P.; Olson, R.; Wilkowski, O.G.; Marschall, C.; Schmidt, R.

    1997-06-01

    This report presents the results from Subtask 1.3 of the International Piping Integrity Research Group (IPIRG) program. The objective of Subtask 1.3 is to develop data to assess analysis methodologies for characterizing the fracture behavior of circumferentially cracked pipe in a representative piping system under combined inertial and displacement-controlled stresses. A unique experimental facility was designed and constructed. The piping system evaluated is an expansion loop with over 30 meters of 16-inch diameter Schedule 100 pipe. The experimental facility is equipped with special hardware to ensure system boundary conditions could be appropriately modeled. The test matrix involved one uncracked and five cracked dynamic pipe-system experiments. The uncracked experiment was conducted to evaluate piping system damping and natural frequency characteristics. The cracked-pipe experiments evaluated the fracture behavior, pipe system response, and stability characteristics of five different materials. All cracked-pipe experiments were conducted at PWR conditions. Material characterization efforts provided tensile and fracture toughness properties of the different pipe materials at various strain rates and temperatures. Results from all pipe-system experiments and material characterization efforts are presented. Results of fracture mechanics analyses, dynamic finite element stress analyses, and stability analyses are presented and compared with experimental results.

  3. The influence of gouge defects on failure pressure of steel pipes

    International Nuclear Information System (INIS)

    Alang, N A; Razak, N A; Zulfadli, M R

    2013-01-01

    Failure pressure of API X42 steel pipes with gouge defects was estimated through a nonlinear finite element (FE) analysis. The effect of gouge length on failure pressure of different pipe diameters was investigated. Stress modified critical strain (SMCS) model was applied as in predicting the failure of the pipe. The model uses strain based criteria to predict the failure. For validation of the model, the FE results were compared to experimental data in literature showing overall good agreement. The results show that the gouge length has significant influence on failure pressure. A smaller pipe diameter gives highest value of failure pressure

  4. Analysis of a piping system for requalification

    International Nuclear Information System (INIS)

    Hsieh, B.J.; Tang, Yu.

    1992-01-01

    This paper discusses the global stress analysis required for the seismic/structural requalification of a reactor secondary piping system in which minor defects (flaws) were discovered during a detailed inspection. The flaws in question consisted of weld imperfections. Specifically, it was necessary to establish that the stresses at the flawed sections did not exceed the allowables and that the fatigue life remained within acceptable limits. At the same time the piping system had to be qualified for higher earthquake loads than those used in the original design. To accomplish these objectives the nominal stress distributions in the piping system under the various loads (dead load, thermal load, wind load and seismic load) were determined. First a best estimate finite element model was developed and calculations were performed using the piping analysis modules of the ANSYS Computer Code. Parameter studies were then performed to assess the effect of physically reasonable variations in material, structural, and boundary condition characteristics. The nominal stresses and forces so determined, provided input for more detailed analyses of the flawed sections. Based on the reevaluation, the piping flaws were judged to be benign, i.e., the piping safety margins were acceptable inspite of the increased seismic demand. 13 refs

  5. Volatile organic components migrating from plastic pipes (HDPE, PEX and PVC) into drinking water.

    Science.gov (United States)

    Skjevrak, Ingun; Due, Anne; Gjerstad, Karl Olav; Herikstad, Hallgeir

    2003-04-01

    High-density polyethylene pipes (HDPE), crossbonded polyethylene pipes (PEX) and polyvinyl chloride (PVC) pipes for drinking water were tested with respect to migration of volatile organic components (VOC) to water. The odour of water in contact with plastic pipes was assessed according to the quantitative threshold odour number (TON) concept. A major migrating component from HDPE pipes was 2,4-di-tert-butyl-phenol (2,4-DTBP) which is a known degradation product from antioxidants such as Irgafos 168(R). In addition, a range of esters, aldehydes, ketones, aromatic hydrocarbons and terpenoids were identified as migration products from HDPE pipes. Water in contact with HDPE pipes was assessed with respect to TON, and values > or =4 were determined for five out of seven brands of HDPE pipes. The total amount of VOC released to water during three successive test periods were fairly constant for the HDPE pipes. Corresponding migration tests carried out for PEX pipes showed that VOC migrated in significant amounts into the test water, and TON >/=5 of the test water were observed in all tests. Several of the migrated VOC were not identified. Oxygenates predominated the identified VOC in the test water from PEX pipes. Migration tests of PVC pipes revealed few volatile migrants in the test samples and no significant odour of the test water.

  6. Effective applied moment in circumferential through-wall cracked pipes for leak-before-break evaluation considering pipe restraint effects

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Yeji; Hwang, Il-Soon [Seoul National University, Seoul 08826 (Korea, Republic of); Oh, Young-Jin, E-mail: yjoh2@kepco-enc.com [KEPCO Engineering and Construction Co. Inc., Gimcheon 39660 (Korea, Republic of)

    2016-05-15

    Highlights: • Effective applied moment at pipe cracked section considering the pipe restraint effect. • Verification of the proposed evaluation methods using finite element analyses. • Applicability for distributed external load of the proposed methods. - Abstract: In the leak-before-break (LBB) design of nuclear power plants, crack opening displacement (COD) is an essential element for determining the length of the leakage size crack. Recent researches regarding the evaluation of COD have indicated that the current practice of the LBB evaluation without consideration of the pressure induced bending (PIB) restraint overestimates COD, which in turn gives non-conservative results. Under a free-ended boundary condition, however, the applied moment at cracked section also can be overestimated, which has conservative effects on LBB evaluation. Therefore, it is necessary to evaluate pipe restraint effects on the applied moment as well as on COD to keep the constancy. In this paper, an evaluation method for the effect of the PIB restraint on COD and an effective applied moment (=crack driving force) at cracked section was developed. Both the linear elastic and elastic–plastic behaviors of the crack were considered. By comparing the behaviors with 3-D finite element analysis results from earlier studies, it was confirmed that the proposed methods make accurate estimations of the PIB restraint effect on COD. Next, the applicability of the proposed method to other types of external loading conditions was examined.

  7. Experimental analytical study on heat pipes

    International Nuclear Information System (INIS)

    Ismail, K.A.R.; Liu, C.Y.; Murcia, N.

    1981-01-01

    An analytical model is developed for optimizing the thickness distribution of the porous material in heat pipes. The method was used to calculate, design and construct heat pipes with internal geometrical changes. Ordinary pipes are also constructed and tested together with the modified ones. The results showed that modified tubes are superior in performance and that the analytical model can predict their performance to within 1.5% precision. (Author) [pt

  8. Space qualification of high capacity grooved heat pipes

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, M; Mullender, B; Druart, J [SABCA, Societe Anomyme Belgel de Construction Aeronautique (Belgium); Supper, W; Beddows, A [ESTEC-The (Netherlands)

    1997-12-31

    Based on the thermal requirements of the future telecommunication satellites, the development of a High Capacity Grooved Heat Pipe (HPG), was contracted by ESA to SABCA leading to an aluminium extruded heat pipe (outer diameter of 25 mm) based on a multi re-entrant grooves design. After an intensive acceptance test campaign whose results showed a good confidence in the design and the fulfillment of the required specifications of heat transport and on tilt capability (experimental maximum heat transport capability of 1500 Watt metres for a vapour temperature of 20 deg C), similar heat pipes have been developed with various outer diameters (11 mm, 15 mm and 20 mm) and with various shapes (circular outer shapes, integrated saddles). Several of these heat pipes were tested during two parabolic flight campaigns, by varying the heat loads during the micro-gravity periods. This HGP heat pipe family is now being submitted to a space qualification program according to ESA standards (ESA PSS-49), both in straight and bent configuration. Within this qualification, the heat pipes are submitted to an extended test campaign including environmental (random/sinus vibration, constant acceleration) and thermal tests (thermal performance, thermal cycle, thermal soak, ageing). (authors) 9 refs.

  9. Space qualification of high capacity grooved heat pipes

    Energy Technology Data Exchange (ETDEWEB)

    Dubois, M.; Mullender, B.; Druart, J. [SABCA, Societe Anomyme Belgel de Construction Aeronautique (Belgium); Supper, W.; Beddows, A. [ESTEC-The (Netherlands)

    1996-12-31

    Based on the thermal requirements of the future telecommunication satellites, the development of a High Capacity Grooved Heat Pipe (HPG), was contracted by ESA to SABCA leading to an aluminium extruded heat pipe (outer diameter of 25 mm) based on a multi re-entrant grooves design. After an intensive acceptance test campaign whose results showed a good confidence in the design and the fulfillment of the required specifications of heat transport and on tilt capability (experimental maximum heat transport capability of 1500 Watt metres for a vapour temperature of 20 deg C), similar heat pipes have been developed with various outer diameters (11 mm, 15 mm and 20 mm) and with various shapes (circular outer shapes, integrated saddles). Several of these heat pipes were tested during two parabolic flight campaigns, by varying the heat loads during the micro-gravity periods. This HGP heat pipe family is now being submitted to a space qualification program according to ESA standards (ESA PSS-49), both in straight and bent configuration. Within this qualification, the heat pipes are submitted to an extended test campaign including environmental (random/sinus vibration, constant acceleration) and thermal tests (thermal performance, thermal cycle, thermal soak, ageing). (authors) 9 refs.

  10. Assessment of NDE Methods on Inspection of HDPE Butt Fusion Piping Joints for Lack of Fusion with Validation from Mechanical Testing

    International Nuclear Information System (INIS)

    Anderson, Michael T.; Cinson, Anthony D.; Crawford, Susan L.; Doctor, Steven R.; Moran, Traci L.; Watts, Michael W.

    2010-01-01

    Studies at the Pacific Northwest National Laboratory (PNNL) in Richland, Washington, are being conducted to evaluate nondestructive examinations (NDE) coupled with mechanical testing of butt fusion joints in high-density polyethylene (HDPE) pipe for assessing lack of fusion. The work provides information to the U.S. Nuclear Regulatory Commission (NRC) on the effectiveness of volumetric inspection techniques of HDPE butt fusion joints in Section III, Division 1, Class 3, buried piping systems in nuclear power plants. This paper describes results from preliminary assessments using ultrasonic and microwave nondestructive techniques and mechanical testing with the high-speed tensile impact test and the side-bend test for determining joint integrity. A series of butt joints were fabricated in 3408, 12-in. IPS DR-11 HDPE material by varying the fusion parameters to create good joints and joints containing a range of lack-of-fusion conditions. Six of these butt joints were volumetrically examined with time-of-flight diffraction (TOFD), phased-array (PA) ultrasound, and the Evisive microwave system. The outer-diameter weld beads were removed for the microwave inspection. In two of the four pipes, both the outer and inner weld beads were removed and the pipe joints re-evaluated. The pipes were sectioned and the joints destructively evaluated with the side-bend test by cutting portions of the fusion joint into slices that were planed and bent. The last step in this limited study will be to correlate the fusion parameters, nondestructive, and destructive evaluation results to validate the effectiveness of what each NDE technology detects and what each does not detect. The results of the correlation will be used in identifying any future work that is needed.

  11. Load-deflection characteristics of small-bore insulated-pipe clamps

    International Nuclear Information System (INIS)

    Severud, L.K.; Clark, G.L.

    1981-12-01

    The special insulated clamps used on both FFTF and CRBR piping utilize a Belleville spring arrangement to compensate for pipe thermal expansion. Analysis indicates that this produces a non-linear, directionally sensitive clamp spring rate. Since these spring rates influence the seismic response of a supported piping system, it was deemed necessary to evaluate them further by test. This has been accomplished for the FFTF clamps. A more standard insulated pipe clamp, which does not incorporate Belleville springs to accommodate thermal expansion, was also tested. This type clamp is simple in design, and economically attractive. It may have wide application prospects for use in LMFBR small bore auxiliary piping operating at temperatures below 427 0 C. Load deflection tests were conducted on 2.54 CM and 7.62 CM diameter samples of these commercial clamps

  12. Noncondensable gas accumulation phenomena in nuclear power plant piping

    International Nuclear Information System (INIS)

    Yamamoto, Yasushi; Aoki, Kazuyoshi; Sato, Teruaki; Shida, Akira; Ichikawa, Nagayoshi; Nishikawa, Akira; Inagaki, Tetsuhiko

    2011-01-01

    In the case of the boiling water reactor, hydrogen and oxygen slightly exist in the main steam, because these noncondensable gases are generated by the radiolytic decomposition of the reactor water. BWR plants have taken measures to prevent noncondensable gas accumulation. However, in 2001, the detonation of noncondensable gases occurred at Hamaoka-1 and Brunsbuttel, resulting in ruptured piping. The accumulation phenomena of noncondensable gases in BWR closed piping must be investigated and understood in order to prevent similar events from occurring in the future. Therefore, an experimental study on noncondensable gas accumulation was carried out. The piping geometries for testing were classified and modeled after the piping of actual BWR plants. The test results showed that 1) noncondensable gases accumulate in vertical piping, 2) it is hard for noncondensable gases to accumulate in horizontal piping, and 3) noncondensable gases accumulate under low-pressure conditions. A simple accumulation analysis method was proposed. To evaluate noncondensable gas accumulation phenomena, the three component gases were treated as a mixture. It was assumed that the condensation amount of the vapor is small, because the piping is certainly wrapped with heat insulation material. Moreover, local thermal equilibrium was assumed. This analysis method was verified using the noncondensable gas accumulation test data on branch piping with a closed top. Moreover, an experimental study on drain trap piping was carried out. The test results showed that the noncondensable gases dissolved in the drain water were discharged from the drain trap, and Henry's law could be applied to evaluate the amount of dissolved noncondensable gases in the drain water. (author)

  13. A Lift-Off-Tolerant Magnetic Flux Leakage Testing Method for Drill Pipes at Wellhead

    OpenAIRE

    Wu, Jianbo; Fang, Hui; Li, Long; Wang, Jie; Huang, Xiaoming; Kang, Yihua; Sun, Yanhua; Tang, Chaoqing

    2017-01-01

    To meet the great needs for MFL (magnetic flux leakage) inspection of drill pipes at wellheads, a lift-off-tolerant MFL testing method is proposed and investigated in this paper. Firstly, a Helmholtz coil magnetization method and the whole MFL testing scheme are proposed. Then, based on the magnetic field focusing effect of ferrite cores, a lift-off-tolerant MFL sensor is developed and tested. It shows high sensitivity at a lift-off distance of 5.0 mm. Further, the follow-up high repeatabilit...

  14. Residual stress state in pipe cut ring specimens for fracture toughness testing

    Energy Technology Data Exchange (ETDEWEB)

    Damjanovic, Darko [J.J. Strossmayer Univ. of Osijek, Slavonski Brod (Croatia). Mechanical Engineering Faculty; Kozak, Drazan [Zagreb Univ. (Croatia). Dept. for Mechanical Design; Marsoner, Stefan [Materials Center, Leoben (Austria).; Gubeljak, Nenad [Maribor Univ. (Slovenia). Chair of Mechanics

    2017-07-01

    Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

  15. Residual stress state in pipe cut ring specimens for fracture toughness testing

    International Nuclear Information System (INIS)

    Damjanovic, Darko; Kozak, Drazan; Marsoner, Stefan; Gubeljak, Nenad

    2017-01-01

    Thin-walled pipes are not suitable for measuring fracture toughness parameters of vital importance because longitudinal crack failure is the most common failure mode in pipes. This is due to the impossibility to manufacture standard specimens for measuring fracture toughness, such as SENB or CT specimens, from the thin wall of the pipe. Previous works noticed this problem, but until now, a good and convenient solution has not been found or developed. To overcome this problem, very good alternative solution was proposed, the so-called pipe ring notched bend specimen (PRNB) [1-5]. Until now, only the idealized geometry PRNB specimen is analyzed, i. e., a specimen which is not cut out from an actual pipe but produced from steel plate. Based on that, residual stresses are neglected along with the imperfections in geometry (elliptical and eccentricity). The aim of this research is to estimate the residual stress state(s) in real pipes used in the boiler industry produced by hot rolling technique. These types of pipes are delivered only in normalized condition, but not stress relieved. Therefore, there are residual stresses present due to the manufacturing technique, but also due to uneven cooling after the production process. Within this paper, residual stresses are estimated by three methods: the incremental hole drilling method (IHMD), X-ray diffraction (XRD) and the splitting method (SM). Knowing the residual stress state in the ring specimen, it is possible to assess their impact on fracture toughness measured on the corresponding PRNB specimen(s).

  16. The insitu lining of cooling water piping

    International Nuclear Information System (INIS)

    Vaughan, W.K.; Oxner, K.B.

    1994-01-01

    The internal corrosion of cooling water piping as well as other industrial piping is becoming an increasing problem to system reliability. There are various alternatives being offered as solutions to the problem including water treatment, coatings, and piping replacement. The in-place lining of these pipes is becoming increasingly popular as a cost-effective method to control corrosion. A cured-in-place plastic composite system can be installed with minimal dismantling or excavation. This paper will examine case histories of the installations of this lining system in power plants at three (3) locations in the United States and one in France. It will also summarize testing that has been performed on the lining system and tests that are currently being performed

  17. limit loads for wall-thinning feeder pipes under combined bending and internal pressure

    International Nuclear Information System (INIS)

    Je, Jin Ho; Lee, Kuk Hee; Chung, Ha Joo; Kim, Ju Hee; Han, Jae Jun; Kim, Yun Jae

    2009-01-01

    Flow Accelerated Corrosion (FAC) during inservice conditions produces local wall-thinning in the feeder pipes of CANDU. The Wall-thinning in the feeder pipes is main degradation mechanisms affecting the integrity of piping systems. This paper discusses the integrity assessment of wall-thinned feeder pipes using limit load analysis. Based on finite element limit analyses, this paper compare limit loads for wall-thinning feeder pipes under combined bending and internal pressure with proposed limit loads. The limit loads are determined from limit analyses based on rectangular wall-thinning and elastic-perfectly-plastic materials using the large geometry change.

  18. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

    Slagis, G.C.; Moore, S.E.

    1994-01-01

    Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs

  19. Multiple blowdown pipe experiments with the PPOOLEX facility

    International Nuclear Information System (INIS)

    Puustinen, M.; Laine, J.; Raesaenen, A.

    2011-03-01

    This report summarizes the results of the experiments with two steel blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  20. Multiple blowdown pipe experiments with the PPOOLEX facility

    Energy Technology Data Exchange (ETDEWEB)

    Puustinen, M.; Laine, J.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2011-03-15

    This report summarizes the results of the experiments with two steel blowdown pipes carried out with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through the blowdown pipes to the condensation pool. The main purpose of the experiment series was to study chugging phenomena (rapid condensation) while steam is discharged through two parallel blowdown pipes into the condensation pool filled with sub-cooled water. Particularly, the aim was to study if the pipe material (polycarbonate) used in the earlier experiment series with two blowdown pipes has had an effect on the general chugging behaviour and measured loads. In the experiments the initial temperature of the pool water was 20 deg. C. The steam flow rate ranged from 220 g/s to 2 350 g/s and the temperature of incoming steam from 148 deg. C to 207 deg. C. The formation and collapse of steam bubbles and the movement of the steam/water interface inside the pipes was non-synchronous. There could be even a 70 ms time difference between the occurrences of steam bubble collapses at the outlets of the two pipes. There was no clear pattern in which pipe the steam bubble first starts to collapse. Several successive bubbles could collapse first in either pipe but then the order changed for a single or several cycles. High pressure loads were measured inside the blowdown pipes due to rapid condensation of the steam volumes in the pipes and resulting water hammer effects. The loads seemed to be higher in pipe 1 than in pipe 2. An explanation for this could be a possible unequal distribution of steam flow between the two pipes. The pipe material has an effect on the condensation phenomena inside the blowdown pipes. A huge difference in the measured pressure curves inside the pipes could be observed compared to the experiments with the polycarbonate pipes. With the same test conditions the amplitude of the

  1. Determination of flexibility factors in curved pipes with end restraints using a semi-analytic formulation

    International Nuclear Information System (INIS)

    Fonseca, E.M.M.; Melo, F.J.M.Q. de; Oliveira, C.A.M.

    2002-01-01

    Piping systems are structural sets used in the chemical industry, conventional or nuclear power plants and fluid transport in general-purpose process equipment. They include curved elements built as parts of toroidal thin-walled structures. The mechanical behaviour of such structural assemblies is of leading importance for satisfactory performance and safety standards of the installations. This paper presents a semi-analytic formulation based on Fourier trigonometric series for solving the pure bending problem in curved pipes. A pipe element is considered as a part of a toroidal shell. A displacement formulation pipe element was developed with Fourier series. The solution of this problem is solved from a system of differential equations using mathematical software. To build-up the solution, a simple but efficient deformation model, from a semi-membrane behaviour, was followed here, given the geometry and thin shell assumption. The flexibility factors are compared with the ASME code for some elbow dimensions adopted from ISO 1127. The stress field distribution was also calculated

  2. Modelling of Aquitaine II pipe whipping test with EUROPLEXUS fast dynamics code

    International Nuclear Information System (INIS)

    Potapov, S.

    2003-01-01

    To validate the modelling of multi-physics phenomena with EUROPLEXUS code we considered a pipe whipping problem occurring in thermal hydraulic conditions of a Loss of Coolant Accident in PWR primary circuit. Two numerical fluid-structure interaction (FSI) models, a simplified 'pipe-like' model and a mixed 1D/3D model, were used to simulate both the conditioning phase and a phase of whipping. The results of calculations were compared with existing experimental data. Analysis of numerical results shows that both models give a good prediction of global behaviour of the coupled fluid-structure system, namely for pipe displacements and stresses in the pipe walls, as well as for pressure and velocity in the fluid. By comparison with experimental data, we show that only the mixed EUROPLEXUS model, where the pipe elbow is discretized with shells, allows us to estimate correctly the time history and maximum value of the contact force between the pipe and the obstacle. The 1D model with reduced kinematics (rigid cross section hypothesis) does not allow the correct detection of contact phenomenon. This study shows that the use of mixed numerical models containing simplified and totally 3D parts duly interconnected allows a very efficient and CPU inexpensive numerical analysis which is able to take into account different global and local physical phenomena. (author)

  3. Evaluation of physical-protection elements for interior applications

    International Nuclear Information System (INIS)

    Scott, S.H.

    1983-01-01

    Considerable emphasis has been given in recent years to the threat of sabotage by an insider at nuclear facilities. This threat is inherently different from the outsider threat of theft or sabotage because of the insiders' unique knowledge and access to vital material and equipment. Thus, special safeguards elements are needed in order to counter the insider threat. In addition, insider physical protection system elements must be compatible with the operations, safety, and maintenance programs at the facility. To help identify elements which meet these needs, field evaluations were performed on an interior access control system, piping sensors, interior video motion detectors, and valve monitoring devices. These elements were tested in a realistic operating environment and both technical and operational evaluation data were obtained. Safeguards element descriptions and the results of the operational tests and evaluations are outlined

  4. Comparison of fracture toughness values from an IPIRG-1 large-scale pipe system test and C(T) specimens on wrought TP304 stainless steel

    International Nuclear Information System (INIS)

    Olson, R.J.; Scott, P.; Marschall, C.W.; Wilkowski, G.M.

    1994-01-01

    Within the First International Piping Integrity Research Group (IPIRG-1) program, pipe system experiments involving dynamic loading with intentionally circumferentially cracked pipe were conducted. The pipe system was fabricated from 406-mm (16-inch) diameter Schedule 100 pipe, and the experiments were conducted at a pressure of 15.5 MPa (2,250 psi) and 288 C (550 F). The loads consisted of pressure, dead-weight, thermal expansion, inertia, and dynamic anchor motion. Significant instrumentation was used to allow the material fracture resistance to be calculated from these large-scale experiments. Three independent analyses were used to calculate the toughness directly from one of these pipe experiments. A comparison of the toughness values from the stainless steel base metal pipe experiment to standard quasi-static and dynamic C(T) specimen tests showed the pipe toughness value was significantly lower than that obtained from C(T) specimens. It is hypothesized that the cyclic loading from inertial stresses in this pipe system experiment caused local degradation of the material toughness. Such effects are not considered in current LBB or pipe flaw evaluation criteria

  5. Evaluation of stresses in large diameter, thin walled piping at support locations

    International Nuclear Information System (INIS)

    Bryan, B.J.; Flanders, H.E. Jr.; Rawls, G.B. Jr.

    1992-01-01

    The highest stresses in many thin walled piping systems are the local stresses at the pipe supports. These secondary stresses are caused by saddles or other structural discontinuities that restrain pipe ovalization. A static analysis of a thin walled pipe supported on structural steel saddle under dead weight loading is presented. The finite element analysis is performed using a shell model with distributed gravity and hydrostatic pressure loading. Parametric studies on global and local stress are performed to determine the effect of the pipe diameter to thickness ratio. Two aspects of the saddle design are also investigated: the effect of saddle width, and the effect of saddle wrap angle. Additionally, the computed stresses are compared to closed form solutions

  6. The response of liquid-filled pipes to vapour collapse

    International Nuclear Information System (INIS)

    Tijsseling, A.S.; Fan, D.

    1991-01-01

    The collapse of vapour cavities in liquid is usually accompanied with almost instantaneous pressure rises. These pressure rises impose severe loads on liquid-conveying pipes whenever the cavities become sufficiently large. Due to the impact nature of loadings, movement of the pipe walls can be expected. Tests are performed in a water-filled closed pipe suspended by thin steel wires. Vaporous cavities are induced in the liquid by hitting the pipe axially by a steel rod. The volume of the cavities can be varied by changing the initial pressure of the water. The developing and collapsing of cavities in the liquid is inferred from pressure measurements. Strain gauges and a laser Doppler vibrometer are used to record the response of the pipe to these pressures. The test results are compared with predictions from a numerical model. The model describes 1) axial stress wave propagations in the pipe and 2) water hammer and cavitation phenomena in the liquid. Pipe and liquid interact via 1) the radial expansion and contraction of the pipe wall and 2) the closed ends of the pipe, where large vapour cavities may develop. (author)

  7. Elastic and inelastic methods of piping systems analysis: a preliminary review

    International Nuclear Information System (INIS)

    Reich, M.; Esztergar, E.P.; Spence, J.; Boyle, J.; Chang, T.Y.

    1975-02-01

    A preliminary review of the methods used for elastic and inelastic piping system analysis is presented. The following principal conclusions are reached: techniques for the analysis of complex piping systems operating in the high temperature creep regime should be further developed; accurate analysis of a complete pipework system in creep using the ''complete shell finite element methods'' is not feasible at the present, and the ''reduced shell finite element method'' still requires excessive computer time and also requires further investigation regarding the compatibility problems associated with the pipe bend element, particularly when applied to cases involving general loading conditions; and with the current size of proposed high temperature systems requiring the evaluation of long-term operating life (30 to 40 years), it is important to adopt a simplified analysis method. A design procedure for a simplified analysis method based on currently available techniques applied in a three-stage approach is outlined. The work required for implementation of these procedures together with desirable future developments are also briefly discussed. Other proposed simplified approximations also are reviewed in the text. 101 references. (U.S.)

  8. Interpretation, with respect to ASME code Case N-318, of limit moment and fatigue tests of lugs welded to pipe

    International Nuclear Information System (INIS)

    Foster, D.C.; Van Duyne, D.A.; Budlong, L.A.; Muffett, J.W.; Wais, E.A.; Streck, G.; Rodabaugh, E.C.

    1990-01-01

    Two nonmandatory ASME code cases have been used often in the evaluation of lugs on nuclear-power- plant piping systems. ASME Code Case N-318 provides guidance for evaluation of the design of rectangular cross-section attachments on Class 2 or 3 piping, and ASME Code Case N-122 provides guidance for evaluation of lugs on Class 1 piping. These code cases have been reviewed and evaluated based on available test data. The results indicate that the Code cases are overly conservative. Recommendations for revisions to the cases are presented which, if adopted, will reduce the overconservatism

  9. Flow induced vibrations of piping system (Vibration sources - Mechanical response of the pipes)

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.; Villard, B.

    1978-01-01

    In order to design the supports of piping system, an estimation of the vibration induced by the fluid conveyed through the pipes are generally needed. For that purpose it is necessary. To evaluate the power spectra of all the main sources generated by the flow. These sources are located at the singular points of the circuit (enlargements, bends, valves, etc. ...). To calculate the modal parameters of fluid containing pipes. This paper presents: a methodical study of the most current singularities. Inter-correlation spectra of local pressure fluctuation downstream from the singularity and correlation spectra of associated acoustical sources have been measured. A theory of noise generation by unsteady flow in internal acoustics has been developed. All these results are very useful for evaluating the source characteristics in most practical pipes. A comparison between the calculation and the results of an experimental test has shown a good agreement

  10. On the optimization of support positioning and stiffness for piping systems in power plants

    International Nuclear Information System (INIS)

    Collina, A.; Zanaboni, P.; Belloli, M.

    2005-01-01

    The optimal location of supports for the reduction of vibration in piping systems is an interesting structural problem, that can be approached also with the methods used in the updating parameters problems, especially when the cause of the vibration is due to a resonance or a ineffective damping level, and the shift of a critical frequency is an effective remedy. In the proposed paper a frequency domain method of Finite Element model updating is used in order to properly locate the natural frequencies of a given piping system, starting from the nominal condition. The method considers modal parameters, and enables to directly update the physical quantities of the finite element model, i.e., in the considered application, the stiffness and damping of the supporting devices. The model of the structure is made up by beam finite elements, lumped springs and masses, and rigid links. The general procedure is applied here updating the stiffness and location of the supports of the piping according to the comparison among the current frequencies of the piping systems structure and the ones that are required, according to specific requirements. Application to a system similar to a portion of an actual AP1000 (a new Advanced Passive Nuclear Reactor) piping-support layout is presented. (authors)

  11. Simplified inelastic (plastic and creep) analysis of pipe elbows subjected to inplane and out-of-plane bending

    International Nuclear Information System (INIS)

    Mello, R.M.; Scheller, J.D.

    1975-01-01

    The inelastic analysis of elbows typical of use in Liquid Metal Fast Reactor piping systems is described. Detailed information on stresses, plastic and creep strains, and deformations throughout the elbow bodies resulting from elastic/plastic, elastic/plastic/creep, and elastic/plastic-creep/relaxation material behavior was obtained for the cases of applied inplane and out-of-plane moment loading on the elbows. Some conclusions are made concerning the nature of the resulting stresses in the elbows. The simplified pipe-bend element in the MARC nonlinear finite element program is used as the analytical tool. This element permits nonlinear analysis of piping elbows and pipeline systems at realistic cost. 16 refs

  12. Investigation on vibrational evaluation criteria for small-bore pipe

    International Nuclear Information System (INIS)

    Tsuji, Takashi; Maekawa, Akira; Takahashi, Tsuneo; Kato, Minoru; Torigoe, Yuichi

    2013-01-01

    The well-known organization such as API and SwRI in USA developed criteria for piping vibrational evaluation. These criteria are targeted for main pipes, but not branch pipes with small bore. In this study, applicability of criteria of API and SwRI to branch pipes was investigated. Vibration test using piping system with small bore branch pipe was conducted and amplitudes of vibrational stress and displacement were measured for various exciting force. In comparison of the measurements with the two criteria, though the criteria of API and SwRI were applicable to small bore branch pipe, they made too conservative evaluation. (author)

  13. Control in fabrication of PWR and BWR type reactor fuel elements

    International Nuclear Information System (INIS)

    Gorskij, V.V.

    1981-01-01

    Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

  14. Solar heat-pipe wick modeling

    Energy Technology Data Exchange (ETDEWEB)

    Andraka, C.E.

    1999-07-01

    Stirling-cycle engines have been identified as a promising technology for the conversion of concentrated solar energy into usable electrical power. In previous experimental work, the author has demonstrated that a heat pipe receiver can significantly improve system performance over a directly-illuminated heater head. The design and operating conditions of a heat pipe receiver differ significantly from typical laboratory heat pipes. New wick structures have been developed to exploit the characteristics of the solar generation system. Typically, these wick structures allow vapor generation within the wick. Conventional heat pipe models do not handle this enhancement, yet it can more than double the performance of the wick. In this study, the author developed a steady-state model of a boiling-enhanced wick for a solar heat pipe receiver. The model is used for design-point calculations and is written in FORTRAN90. Some limited comparisons have been made with actual test data.

  15. Response of buried pipes to missile impact

    International Nuclear Information System (INIS)

    Vardanega, C.; Cremonini, M.G.; Mirone, M.; Luciani, A.

    1989-01-01

    This paper presents the methodology and results of the analyses carried out to determine an effective layout and the dynamic response of safety related cooling water pipes, buried in backfill, for the Alto Lazio Nuclear Power Plant in Italy, subjected to missile impact loading at the backfill surface. The pipes are composed of a steel plate encased in two layers of high-quality reinforced concrete. The methodology comprises three steps. The first step is the definition of the 'free-field' dynamic response of the backfill soil, not considering the presence of the pipes, through a dynamic finite element direct integration analysis utilizing an axisymmetric model. The second step is the pipe-soil interaction analysis, which is conducted by utilizing the soil displacement and stress time-histories obtained in the previous steps. Soil stress time-histories, combined with the geostatic and other operational stresses (such as those due to temperature and pressure), are used to obtain the actions in the pipe walls due to ring type deformation. For the third step, the analysis of the beam type response, a lumped parameter model is developed which accounts for the soil stiffness, the pipe characteristics and the position of the pipe with respect to the impact area. In addition, the effect of the presence of large concrete structures, such as tunnels, between the ground surface and the pipe is evaluated. The results of the structural analyses lead to defining the required steel thickness and also allow the choice of appropriate embedment depth and layout of redundant lines. The final results of the analysis is not only the strength verification of the pipe section, but also the definition of an effective layout of the lines in terms of position, depth, steel thickness and joint design. (orig.)

  16. Evaluation of underground pipe-structure interface for surface impact load

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Shen, E-mail: swang@terrapower.com

    2017-06-15

    Highlights: • A simple method is proposed for the evaluation of underground pipelines for surface impact load considering the effect of a nearby pipe-structure interface. • The proposed simple method can be used to evaluate the magnitude of damage within a short period of time after accidental drop occurs. • The proposed method is applied in a practical example and compared by using finite element analysis. - Abstract: Nuclear safety related buried pipelines need to be assessed for the effects of postulated surface impact loads. In published solutions, the buried pipe is often considered within an elastic half space without interference with other underground structures. In the case that a surface impact occurs in short distance from an underground pipe-structure interface, this boundary condition will further complicate the buried pipe evaluation. Neglecting such boundary effect in the assessment may lead to underestimating potential damage of buried pipeline, and jeopardizing safety of the nuclear power plant. Comprehensive analysis of such structure-pipe-soil system is often subjected to availability of state-of-art finite element tools, as well as costly and time consuming. Simple, but practical conservative techniques have not been established. In this study, a mechanics based solution is proposed in order to assess the magnitude of damage to a buried pipeline beneath a heavy surface impact considering the effect of a nearby pipe-structure interface. The proposed approach provides an easy to use tool in the early stage of evaluation before the decision of applying more costly technique can be made by owner of the nuclear facility.

  17. Evaluation of underground pipe-structure interface for surface impact load

    International Nuclear Information System (INIS)

    Wang, Shen

    2017-01-01

    Highlights: • A simple method is proposed for the evaluation of underground pipelines for surface impact load considering the effect of a nearby pipe-structure interface. • The proposed simple method can be used to evaluate the magnitude of damage within a short period of time after accidental drop occurs. • The proposed method is applied in a practical example and compared by using finite element analysis. - Abstract: Nuclear safety related buried pipelines need to be assessed for the effects of postulated surface impact loads. In published solutions, the buried pipe is often considered within an elastic half space without interference with other underground structures. In the case that a surface impact occurs in short distance from an underground pipe-structure interface, this boundary condition will further complicate the buried pipe evaluation. Neglecting such boundary effect in the assessment may lead to underestimating potential damage of buried pipeline, and jeopardizing safety of the nuclear power plant. Comprehensive analysis of such structure-pipe-soil system is often subjected to availability of state-of-art finite element tools, as well as costly and time consuming. Simple, but practical conservative techniques have not been established. In this study, a mechanics based solution is proposed in order to assess the magnitude of damage to a buried pipeline beneath a heavy surface impact considering the effect of a nearby pipe-structure interface. The proposed approach provides an easy to use tool in the early stage of evaluation before the decision of applying more costly technique can be made by owner of the nuclear facility.

  18. Report of the U.S. Nuclear Regulatory Commission Piping Review Committee. Summary and evaluation of historical strong-motion earthquake seismic response and damage to aboveground industrial piping

    International Nuclear Information System (INIS)

    1985-04-01

    The primary purpose of this report is to collect in one reference document the observation and experience that has been developed with regard to the seismic behavior of aboveground, building-supported, industrial-type process piping (similar to piping used in nuclear power plants) in strong-motion earthquakes. The report will also contain observations regarding the response of piping in strong-motion experimental tests and appropriate conclusions regarding the behavior of such piping in large earthquakes. Recommendations are included covering the future design of such piping to resist earthquake motion damage based on observed behavior in large earthquakes and simulated shake table testing. Since available detailed data on the behavior of aboveground (building-supported) piping are quite limited, this report will draw heavily on the observations and experiences of experts in the field. In Section 2 of this report, observed earthquake damage to aboveground piping in a number of large-motion earthquakes is summarized. In Section 3, the available experience from strong-motion testing of piping in experimental facilities is summarized. In Section 4 are presented some observations that attempt to explain the observed response of piping to strong-motion excitation from actual earthquakes and shake table testing. Section 5 contains the conclusions based on this study and recommendations regarding the future seismic design of piping based on the observed strong-motion behavior and material developed for the NPC Piping Review Committee. Finally, in Section 6 the references used in this study are presented. It should be understood that the use of the term piping in this report, in general, is limited to piping supported by building structures. It does not include behavior of piping buried in soil media. It is believed that the seismic behavior of buried piping is governed primarily by the deformation of the surrounding soil media and is not dependent on the inertial response

  19. Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe.

    Science.gov (United States)

    Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

    2015-01-01

    On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes.

  20. Heat Transfer Modeling of an Annular On-Line Spray Water Cooling Process for Electric-Resistance-Welded Steel Pipe

    Science.gov (United States)

    Chen, Zejun; Han, Huiquan; Ren, Wei; Huang, Guangjie

    2015-01-01

    On-line spray water cooling (OSWC) of electric-resistance-welded (ERW) steel pipes can replace the conventional off-line heat treatment process and become an important and critical procedure. The OSWC process improves production efficiency, decreases costs, and enhances the mechanical properties of ERW steel pipe, especially the impact properties of the weld joint. In this paper, an annular OSWC process is investigated based on an experimental simulation platform that can obtain precise real-time measurements of the temperature of the pipe, the water pressure and flux, etc. The effects of the modes of annular spray water cooling and related cooling parameters on the mechanical properties of the pipe are investigated. The temperature evolutions of the inner and outer walls of the pipe are measured during the spray water cooling process, and the uniformity of mechanical properties along the circumferential and longitudinal directions is investigated. A heat transfer coefficient model of spray water cooling is developed based on measured temperature data in conjunction with simulation using the finite element method. Industrial tests prove the validity of the heat transfer model of a steel pipe undergoing spray water cooling. The research results can provide a basis for the industrial application of the OSWC process in the production of ERW steel pipes. PMID:26201073

  1. Recent evaluations of crack-opening-area in circumferentially cracked pipes

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, S.; Brust, F.; Ghadiali, N.; Wilkowski, G.; Miura, N.

    1997-04-01

    Leak-before-break (LBB) analyses for circumferentially cracked pipes are currently being conducted in the nuclear industry to justify elimination of pipe whip restraints and jet shields which are present because of the expected dynamic effects from pipe rupture. The application of the LBB methodology frequently requires calculation of leak rates. The leak rates depend on the crack-opening area of the through-wall crack in the pipe. In addition to LBB analyses which assume a hypothetical flaw size, there is also interest in the integrity of actual leaking cracks corresponding to current leakage detection requirements in NRC Regulatory Guide 1.45, or for assessing temporary repair of Class 2 and 3 pipes that have leaks as are being evaluated in ASME Section XI. The objectives of this study were to review, evaluate, and refine current predictive models for performing crack-opening-area analyses of circumferentially cracked pipes. The results from twenty-five full-scale pipe fracture experiments, conducted in the Degraded Piping Program, the International Piping Integrity Research Group Program, and the Short Cracks in Piping and Piping Welds Program, were used to verify the analytical models. Standard statistical analyses were performed to assess used to verify the analytical models. Standard statistical analyses were performed to assess quantitatively the accuracy of the predictive models. The evaluation also involved finite element analyses for determining the crack-opening profile often needed to perform leak-rate calculations.

  2. An analysis of a pipe bend subjected to in-plane loads

    International Nuclear Information System (INIS)

    Hellen, T.K.

    1979-01-01

    This report describes a set of finite element analyses conducted on a pipe bend subjected to in-plane loads. The pipe is thin-walled, and two types of finite element, shells and solid bricks, are compared elastically. An alternative semi-analytical technique has also been used and experimental results are available, all of which show good correlative agreement. The use of suitable mesh refinement and order of numerical integration is examined. Finally, the solid elements are used to follow a loading sequence incorporating elasto-plastic behaviour as conducted by experiment. This work is an updated version of that used for the CEC benchmark calculations for the Fast Reactor Codes and Standards Working Group, Activity No 2, on Structural Analysis. (author)

  3. Mechanized ultrasonic inspection of austenitic pipe systems

    International Nuclear Information System (INIS)

    Dressler, K.; Luecking, J.; Medenbach, S.

    1999-01-01

    The contribution explains the system of standard testing methods elaborated by ABB ZAQ GmbH for inspection of austenitic plant components. The inspection tasks explained in greater detail are basic materials testing (straight pipes, bends, and pipe specials), and inspection of welds and dissimilar welds. The techniques discussed in detail are those for detection and sizing of defects. (orig./CB) [de

  4. Plastic influence functions for calculating J-integral of complex-cracks in pipe

    International Nuclear Information System (INIS)

    Jeong, Jae-Uk; Choi, Jae-Boong; Kim, Moon-Ki; Huh, Nam-Su; Kim, Yun-Jae

    2016-01-01

    In this study, the plastic influence functions, h_1, for estimates of J-integral of a pipe with a complex crack were newly proposed based on the systematic 3-dimensional (3-D) elastic-plastic finite element (FE) analyses by using Ramberg-Osgood (R-O) relation, in which global bending moment, axial tension and internal pressure were considered as loading conditions. Based on the present plastic influence functions, the GE/EPRI-type J-estimation scheme for complex-cracked pipes was suggested, and the results from the proposed J-estimation were compared with the FE results using both R-O fit parameters and actual tensile data of SA376 TP304 stainless steel. The comparison results demonstrate that although the proposed scheme provided sensitive J estimations according to fitting ranges of R-O parameters, it showed overall good agreements with the FE results using R-O relation. Thus, the proposed engineering J prediction method can be utilized to assess instability of a complex crack in pipes for R-O material. - Highlights: • New h_1values of GE/EPRI method for complex-cracked pipes are proposed. • The plastic limit loads of complex-cracked pipes using Mises yield criterion are provided. • The new J estimates of complex-cracked pipes are proposed based on GE/EPRI concept. • The proposed estimates for J are validated against 3-D finite element results.

  5. LABORATORY EVALUATION ON PERFORMANCE OF GLASS FIBER REINFORCED PLASTIC MORTAR PIPE CULVERTS

    OpenAIRE

    Huawang Shi; Lianyu Wei

    2018-01-01

    This paper investigated the performance and behaviour of glass fiber reinforced plastic mortar (FRPM) pipes under different loading conditions. FRPM pipes with inner diameter of 1500 mm were prefabricated in factory. Mechanics performance testing (ring and axial compressive strength and elastic modulus), stiffness and fatigue test were carried out in laboratory. Ring stiffness test provided pipe stiffness (PS) which is a function of geometry and material type of pipe through parallel plate lo...

  6. Pipe support

    International Nuclear Information System (INIS)

    Pollono, L.P.

    1979-01-01

    A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

  7. Abrasion behavior of graphite pebble in lifting pipe of pebble-bed HTR

    Energy Technology Data Exchange (ETDEWEB)

    Shen, Ke; Su, Jiageng [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Zhou, Hongbo [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Chinergy Co., LTD., Beijing 100193 (China); Peng, Wei; Liu, Bing [Institute of Nuclear and New Energy Technology, Advanced Nuclear Energy Technology Cooperation Innovation Center, The Key Laboratory of Advanced Nuclear Engineering and Safety, Ministry of Education, Beijing 10084 (China); Yu, Suyun, E-mail: suyuan@tsinghua.edu.cn [Center for Combustion Energy, The Key Laboratory for Thermal Science and Power Engineering, Ministry of Educations, Tsinghua University, Beijing 10084 (China)

    2015-11-15

    Highlights: • Quantitative determination of abrasion rate of graphite pebbles in different lifting velocities. • Abrasion behavior of graphite pebble in helium, air and nitrogen. • In helium, intensive collisions caused by oscillatory motion result in more graphite dust production. - Abstract: A pebble-bed high-temperature gas-cooled reactor (pebble-bed HTR) uses a helium coolant, graphite core structure, and spherical fuel elements. The pebble-bed design enables on-line refueling, avoiding refueling shutdowns. During circulation process, the pebbles are lifted pneumatically via a stainless steel lifting pipe and reinserted into the reactor. Inevitably, the movement of the fuel elements as they recirculate in the reactor produces graphite dust. Mechanical wear is the primary source of graphite dust production. Specifically, the sources are mechanisms of pebble–pebble contact, pebble–wall (structural graphite) contact, and fuel handling (pebble–metal abrasion). The key contribution to graphite dust production is from the fuel handling system, particularly from the lifting pipe. During pneumatic lift, graphite pebbles undergo multiple collisions with the stainless steel lifting pipe, thereby causing abrasion of the graphite pebbles and producing graphite dust. The present work explored the abrasion behavior of graphite pebble in the lifting pipe by measuring the abrasion rate at different lifting velocities. The abrasion rate of the graphite pebble in helium was found much higher than those in air and nitrogen. This gas environment effect could be explained by either tribology behavior or dynamic behavior. Friction testing excluded the possibility of tribology reason. The dynamic behavior of the graphite pebble was captured by analysis of the audio waveforms during pneumatic lift. The analysis results revealed unique dynamic behavior of the graphite pebble in helium. Oscillation and consequently intensive collisions occur during pneumatic lift, causing

  8. LABORATORY EVALUATION ON PERFORMANCE OF GLASS FIBER REINFORCED PLASTIC MORTAR PIPE CULVERTS

    Directory of Open Access Journals (Sweden)

    Huawang Shi

    2018-04-01

    Full Text Available This paper investigated the performance and behaviour of glass fiber reinforced plastic mortar (FRPM pipes under different loading conditions. FRPM pipes with inner diameter of 1500 mm were prefabricated in factory. Mechanics performance testing (ring and axial compressive strength and elastic modulus, stiffness and fatigue test were carried out in laboratory. Ring stiffness test provided pipe stiffness (PS which is a function of geometry and material type of pipe through parallel plate loading test (PPLT. The fatigue test and micro-structure measure method were used to evaluate the durability effects of FRPM under repeated compression load. Results indicated that FRPM pipes had better mechanic performances as the road culverts under soils. It may be helpful for the design and construction of FRPM culverts.

  9. Pipe Penetrating Radar: a New Tool for the Assessment of Critical Infrastructure

    Science.gov (United States)

    Ekes, C.; Neducz, B.

    2012-04-01

    This paper describes the development of Pipe Penetrating Radar (PPR), the underground in-pipe application of GPR, a non-destructive testing method that can detect defects and cavities within and outside mainline diameter (>18 in / 450mm) non-metallic (concrete, PVC, HDPE, etc.) underground pipes. The method uses two or more high frequency GPR antennae carried by a robot into underground pipes. The radar data is transmitted to the surface via fibre optic cable and is recorded together with the output from CCTV (and optionally sonar and laser). Proprietary software analyzes the data and pinpoints defects or cavities within and outside the pipe. Thus the testing can identify existing pipe and pipe bedding symptoms that can be addressed to prevent catastrophic failure due to sinkhole development and can provide useful information about the remaining service life of the pipe. The key innovative aspect is the unique ability to map pipe wall thickness and deterioration including cracks and voids outside the pipe, enabling accurate predictability of needed intervention or the timing of replacement. This reliable non-destructive testing method significantly impacts subsurface infrastructure condition based asset management by supplying previously unattainable measurable conditions. Keywords: pipe penetrating radar (PPR), ground penetrating radar (GPR), pipe inspection, concrete deterioration, municipal engineering

  10. Seismic ratchet-fatigue failure of piping systems

    International Nuclear Information System (INIS)

    Severud, L.K.; Anderson, M.J.; Lindquist, M.R.; Weiner, E.O.

    1987-01-01

    Failures of piping systems during earthquakes have been rare. Those that have failed were either made of brittle material such as cast iron, were rigid systems between major components where component relative seismic motions tore the pipe out of the component, or were high pressure systems where a ratchet-fatigue fracture followed a local bulging of the pipe diameter. Tests to failure of an unpressurized 3-inch and a pressurized 6-inch diameter carbon steel nuclear pipe systems subjected to high-level shaking have been accomplished. The high-level shaking loads needed to cause failure were much higher than ASME Code rules would permit with present design limits. Failure analyses of these tests are presented and correlated to the test results. It was found that failure of the unpressurized system could be correlated well with standard ASME type fatigue analysis predictions. Moreover, the pressurized system failure occured in significantly less load cycles than predicted by standard fatigue analysis. However, a ratchet-fatigue and ductility exhaustion analysis of the pressurized system did correlate reasonably well. These findings indicate modifications to design analysis methods and the present ASME Code piping design rules to reduce unneeded conservatisms and to cover the ratchet-fatigue failure mode may be appropriate

  11. Effect of piping systems on surge in centrifugal compressors

    International Nuclear Information System (INIS)

    Tamaki, Hideaki

    2008-01-01

    There is a possibility that the exchange of the piping system may change the surge characteristic of a compressor. The piping system of a plant is not always the same as that of a test site. Then it is important to evaluate the effect of piping systems on surge characteristics in centrifugal compressors. Several turbochargers combined with different piping systems were tested. The lumped parameter model which was simplified to be solved easily was applied for the prediction of surge point. Surge lines were calculated with the linearlized lumped parameter model. The difference between the test and calculated results was within 10 %. Trajectory of surge cycle was also examined by solving the lumped parameter model. Mild surge and deep surge were successfully predicted. This study confirmed that the lumped parameter model was a very useful tool to predict the effect of piping systems on surge characteristics in centrifugal compressors, even though that was a simple model

  12. Study of elasticity and limit analysis of joints and branch pipe tee connections

    International Nuclear Information System (INIS)

    Plancq, David

    1997-01-01

    The industrial context of this study is the behaviour and sizing the pipe joints in PWR and fast neutron reactors. Two aspects have been approached in this framework. The first issue is the elastic behaviour of the pipe joining with a plane or spherical surface or with another pipe in order to get a better understanding of this components usually modelled in classical calculations in a very simplified way. We focused our search on the bending of an intersecting pipe. In the case of the intersection with a plane surface we have conducted our study on the basis of literature results. In the case of intersection on a spherical surface we have also solved entirely the problem by using a sphere shell description different from that usually utilized. Finally, we give an approach to obtain a simple result for the bending of branch pipe tee joints allowing the formulation of a specific finite element. The second issue approached is the limit analysis which allows characterising the plastic failure of this structures and defining reference constraints. This constraints are used in numerous applications. We mention here the rules of pipe sizing and analyzing under primary load, the mechanics of cracks and the definition of global plasticity criteria. To solve this problem we concentrated our studies on the development of a new calculation techniques for the limit load called elastic compensation method (ECM). We have tested it on a large number of classical structures and on the branch pipe tee connections. We propose also a very simple result regarding the lower limit of the bending of a tee junction

  13. Effects of induction heating parameters on controlling residual stress in intermediate size pipes

    International Nuclear Information System (INIS)

    Rybicki, E.F.; McGuire, P.A.

    1981-01-01

    Induction heating for stress improvement (IHSI) is a method for reducing the tensile weld induced stresses on the inner surfaces of the girth welded pipes. The process entails inductively heating the outside of a welded pipe while cooling the inner surface with flowing water. A 10-inch schedule 80 Type 304 stainless steel pipe was selected for this study. Residual stresses due to welding were first determined using a finite element computational model. 26 refs

  14. Seismic Capacity Estimation of Steel Piping Elbow under Low-cycle Fatigue Loading

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Bub Gyu; Kim, Sung Wan; Choi, Hyoung Suk; Kim, Nam Sik [Pusan National University, Busan (Korea, Republic of); Hahm, Dae Gi [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In some cases, this large relative displacement can increase seismic risk of the isolated facility. Especially, a inelastic behavior of crossover piping system to connect base isolated building and fixed base building can caused by a large relative displacement. Therefore, seismic capacity estimation for isolated piping system is needed to increase safety of nuclear power plant under seismic condition. Dynamic behavior analysis of piping system under seismic condition using shake table tests was performed by Touboul et al in 1995. In accordance with their study, plastic behavior could be occurred at pipe elbow under seismic condition. Experimental researches for dynamic behavior of typical piping system in nuclear power plant have been performed for several years by JNES(Japan Nuclear Energy Safety Organization) and NUPEC(Nuclear Power Engineering Corporation). A low cycle ratcheting fatigue test was performed with scaled model of elbow which is a weakest component in piping system by Mizuno et al. In-plane cyclic loading tests under internal pressure condition were performed to evaluate the seismic capacity of the steel piping elbow. Leakage phenomenon occurred on and near the crown in piping elbow. Those cracks grew up in axial direction. The fatigue curve was estimated from test results. In the fatigue curve, loading amplitude exponentially decreased as the number of cycles increased. A FEM model of piping elbow was modified with test results. The relationships between displacement and force from tests and numerical analysis was well matched.

  15. Pressure piping systems examination. 2. ed

    Energy Technology Data Exchange (ETDEWEB)

    1993-03-01

    This Code is Part 13 of the IP Model Code of Safe Practice in the Petroleum Industry. Its purpose is to provide a guide to safe practices in the in-service examination and test of piping systems used in the petroleum and chemical industries. The Code gives general requirements regarding the provision and maintenance of adequate documentation, in-service examination, the control of modifications and repairs, examination frequency, protective devices and testing of piping systems. (author)

  16. Presentation of accessibility equipment for primary pipings, IHX, pumps and appertaining manipulator tests

    International Nuclear Information System (INIS)

    Hahn, G.; Hoeft, E.

    1980-01-01

    Accessibility and inservice procedure of SNR-300 components are described. Due to the high radiation level in the primary system it was necessary to develop special equipment to permit access to the testing components. The pertinent examination methods for surveying welding seams are acoustic (ultrasonic) and optical procedures (TV cameras, surface crack tests). This can be done by remote-controlled manipulators and special devices, which can transport the inspection system by rails to the testing position. Presently, relatively limited experience exists for such remote-controlled handling in nuclear power plants. Thus model experiments were carried out on a model pipe section at INTERATOM. The performed test shows that the concept planned to perform inservice by using remote-controlled manipulators can be realized successfully. (author)

  17. Remote controlled in-pipe manipulators for dye-penetrant inspection and grinding of weld roots inside of pipes

    International Nuclear Information System (INIS)

    Seeberger, E.K.

    2000-01-01

    Technical plants which have to satisfy stringent safety criteria must be continuously kept in line with the state of art. This applies in particular to nuclear power plants. The quality of piping in nuclear power plants has been improved quite considerably in recent years. By virtue of the very high quality requirements fulfilled in the manufacture of medium-carrying and pressure-retaining piping, one of the focal aspects of in-service inspections is the medium wetted inside of the piping. A remote controlled pipe crawler has been developed to allow to perform dye penetrant testing of weld roots inside piping (ID ≥ 150 mm). The light crawler has been designed such that it can be inserted into the piping via valves (gate valves, check valves,...) with their internals removed. Once in the piping, all crawler movements are remotely controlled (horizontal and vertical pipes incl. the elbows). If indications are found these discontinuities are ground according to a qualified procedure using a special grinding head attached to the crawler with complete extraction of all grinding residues. The in-pipe grinding is a special qualified three (3) step performance that ensures no residual tensile stress (less than 50 N/mm 2 ) in the finish machined austenitic material surface. The in-pipe inspection system, qualified according to both the specifications of the German Nuclear Safety Standards Commission (KTA) and the American Society of Mechanical Engineers (ASME), has already been used successfully in nuclear power plants on many occasions. (author)

  18. ANALYSIS OF MATERIALS IN AN EXPERIMENTAL TESTING PIPE SYSTEM FOR AN INHIBITOR OF MUSSEL KILL

    Energy Technology Data Exchange (ETDEWEB)

    Daniel P. Molloy

    2003-06-04

    A comprehensive series of 16 laboratory experiments demonstrated that the presence of vinyl tubing within a recirculating pipe system was responsible for lowering zebra mussel kill following treatment with the bacterium Pseudomonas fluorescens. All vinyl tubing was replaced in all testing units with silicone tubing, and high mussel kill (>95%) was then obtained.

  19. Transient heat pipe investigations for space power systems

    International Nuclear Information System (INIS)

    Merrigan, M.A.; Keddy, E.S.; Sena, J.T.

    1985-01-01

    A 4-meter long, high temperature, high power, molybdenum-lithium heat pipe has been fabricated and tested in transient and steady state operation at temperatures to 1500 K. Maximum power throughput during the tests was approximately 37 kW/cm 2 for the 1.4 cm diameter vapor space of the annular wick heat pipe. The evaporator flux density for the tests was 150.0 W/cm 2 over a length of 40 cm. Condenser length was approximately 3.0 m with radiant heat rejection from the condenser to a coaxial, water cooled radiation calorimeter. A variable radiation shield, controllable from the outside of the vacuum enclosure, was used to vary the load on the heat pipe during the tests. 1 ref., 9 figs

  20. Store for burnt-up fuel elements of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1981-01-01

    Burnt-up fuel elements of nuclear reactors have to be cooled during storage. For this reason the boxes which surround the fuel elements can have cooling air flowing round them in natural flow. This air is taken through the walls of a storage building through zones of parallel pipes, whose diameter and spacing are in the ratio of 1 : 0.5 to 1 : 2. The pipes have dust filters. Prefilters with fan drive are situated in parallel with the inlet pipe zones. (orig.) [de

  1. An elevator for locked drilling pipes

    Energy Technology Data Exchange (ETDEWEB)

    Gurbanov, R.S.; Abbasov, E.M.; Ismailov, A.A.; Mamedov, Yu.S.; Safarov, A.A.

    1983-01-01

    An elevator is proposed, which includes a body with a door. To reduce the probability of gas shows in a well with high speed lowering and lifting of the column of locked drilling pipes through providing the possibility of feeding a drilling mud in this case into the mine, the elevator is equipped with a pneumatic cylinder with a two way hollow rod, on one face of which a sealing element is mounted for sealing the drilling pipe and on the other, an adapter for feeding the drilling mud. The rod is linked with the sleeve of the pneumatic cylinder, which is rigidly linked with the body with the capability of axial movement without rotation.

  2. Ice plugging of pipes using liquid nitrogen

    International Nuclear Information System (INIS)

    Twigg, R.J.

    1987-03-01

    This report presents a study on the ice plugging of pipe using liquid nitrogen, and is based on a literature review and on discussions with individuals who use the technique. Emphasis is placed on ferritic alloys, primarily carbon steels, in pipe sized up to 60 cm in diameter and on austenitic stainless steels in pipe sizes up to 30 cm in diameter. This technique is frequently used for leak testing in nuclear facilities

  3. Development of VHTR high temperature piping in KHI

    International Nuclear Information System (INIS)

    Suzuki, Nobuhiro; Takano, Shiro

    1981-01-01

    The high temperature pipings used for multi-purpose high temperature gas-cooled reactors are the internally insulated pipings for transporting high temperature, high pressure helium at 1000 deg C and 40 kgf/cm 2 , and the influences exerted by their performance as well as safety to the plants are very large. Kawasaki Heavy Industries, Ltd., has engaged in the development of the high temperature pipings for VHTRs for years. In this report, the progress of the development, the test carried out recently and the problems for future are described. KHI manufactured and is constructing a heater and internally insulated helium pipings for the large, high temperature structure testing loop constructed by Japan Atomic Energy Research Institute. The design concept for the high temperature pipings is to separate the temperature boundary and the pressure boundary, therefore, the double walled construction with internal heat insulation was adopted. The requirements for the high temperature pipings are to prevent natural convection, to prevent bypass flow, to minimize radiation heat transfer and to reduce heat leak through insulator supporters. The heat insulator is composed of two layers, metal laminate insulator and fiber insulator of alumina-silica. The present state of development of the high temperature pipings for VHTRs is reported. (Kako, I.)

  4. Vibration test and endurance test for HANARO 36-element fuel assembly

    International Nuclear Information System (INIS)

    Ryu, Jeong Soo; Kim, Heon ll; Chung, Heung June

    1998-06-01

    Vibration test and endurance test for HANARO DU (depleted uranium) 36-element fuel assembly which was fabricated by KAERI were carried out based on the HANARO operation conditions. The endurance test of 22 days was added to the previous 18 days test. The vibration test was performed at various flow rates. Vibration frequency for the 36-element fuel assembly is between 11 to 14.5 Hz. And the maximum vibration displacement is less than 100 μm. From the endurance test result, it can be concluded that the appreciable fretting wear for the 36-element fuel assembly and the hexagonal flow tube was not observed. (author). 4 refs., 5 tabs., 29 figs

  5. Investigation of Circular Woven Composite Preforms for Composite Pipes

    Directory of Open Access Journals (Sweden)

    Amid Hooman

    2016-06-01

    Full Text Available The main traditional technique for commercial manufacturing of composite pipes is filament winding in which the winding angle and the discontinuity of the structure (caused by starting and ending points of the winding process are two important matters of concern. In the present study, circular woven fabric with its orthogonal net-shaped continuous structure was produced from polyester yarns. Fabric was wet with epoxy and hand lay-up was used to manufacture the composite pipes. Composite pipes were subjected to internal hydrostatic pressure and their burst strength was recorded. In addition, tensile strength of flat laminas was assessed in the warp and weft directions. We estimated and analysed the failure strength of composite pipes using Tresca’s failure criterion and Finite Element (FE modeling. The experimental burst strength was almost 23% more than the FE model and 77% more than the theoretical estimate.

  6. Nuclear piping system damping data studies

    International Nuclear Information System (INIS)

    Ware, A.G.; Arendts, J.G.

    1985-01-01

    A programm has been conducted at the Idaho National Engineering Laboratory to study structural damping data for nuclear piping systems and to evaluate if changes in allowable damping values for structural seismic analyses are justified. The existing pipe damping data base was examined, from which a conclusion was made that there were several sets of data to support higher allowable values. The parameters which most influence pipe damping were identified and an analytical investigation demonstrated that increased damping would reduce the required number of seismic supports. A series of tests on several laboratory piping systems was used to determine the effect of various parameters such as types of supports, amplitude of vibration, frequency, insulation, and pressure on damping. A multiple regression analysis was used to statistically assess the influence of the various parameters on damping, and an international pipe damping data bank has been formed. (orig.)

  7. Effect of Cr content on the FAC of pipe material at 150 .deg. C

    International Nuclear Information System (INIS)

    Park, Tae Jun; Kim, Hong Pyo

    2013-01-01

    Flow accelerated corrosion (FAC) of the carbon steel piping in nuclear power plants (NPPs) has been major issue in nuclear industry. During the FAC, a protective oxide layer on carbon steel dissolves into flowing water leading to a thinning of the oxide layer and accelerating corrosion of base material. As a result, severe failures may occur in the piping and equipment of NPPs. Effect of alloying elements on FAC of pipe materials was studied with rotating cylinder FAC test facility at 150 .deg. C and at flow velocity of 4m/s. The facility is equipped with on line monitoring of pH, conductivity, dissolved oxygen(DO) and temperature. Test solution was the demineralized water, and DO concentration was less than 1 ppb. Surface appearance of A 106 Gr. B which is used widely in secondary pipe in NPPs showed orange peel appearance, typical appearance of FAC. The materials with Cr content higher than 0.17wt.% showed pit. The pit is thought to early degradation mode of FAC. The corrosion product within the pit was enriched with Cr, Mo, Cu, Ni and S. But S was not detected in SA336 F22V with 2.25wt.% Cr. The enrichment of Cr and Mo seemed to be related with low, solubility of Cr and Mo compared to Fe. Measured FAC rate was compared with Ducreaux's relationship and showed slightly lower FAC rate than Ducreaux's relationship

  8. Multi-mode vibration control of piping system

    International Nuclear Information System (INIS)

    Minowa, Takeshi; Seto, Kazuto; Iiyama, Fumiya; Sodeyama, Hiroshi

    1999-01-01

    In this paper, dual dynamic absorbers are applied to the piping system in order to control the multiple vibration modes. ANSYS, which is one of the software based on FEM(finite element method), is used for the design of dual dynamic absorbers as well as for the determination of their optimum installing positions. The dual dynamic absorbers designed optimally for controlling the first three vibration modes perform just like a houde damper in higher frequency and have an effect on controlling higher modes. To use this advantage, three dual dynamic absorbers are installed in positions where they influence higher modes, and not only the first three modes of the piping system but also the extensive modes are controlled. Practical experimental study has also been carried out and it is shown that a dual dynamic absorber is suitable for controlling the vibration of the piping system. (author)

  9. Vibration monitoring of the primary piping system during the hot functional tests of the Muelheim-Kaerlich PWR

    International Nuclear Information System (INIS)

    Bauernfeind, V.; Bloem, T.; Pache, W.; Diederich, H.J.

    1992-01-01

    During the hot functional tests of the Muelheim-Kaerlich plant, which was the first plant of its type, vibration measurements were made on the reactor pressure vessel and its internal parts and on the primary piping system and the main coolant pumps. This paper contains the results of the measurements taken on the pipes and the pumps with an interpretation of these measurements based on an analytical model of the primary system. The main aim of the measurement programs is to confirm that the components, which are of new structural design, are adequately dimensioned for the operational vibration loads during the service life of the reactor. In addition, the vibrational modes of the hot lines, the steam generators and the pumps with the adjacent cold lines were determined. These values were compared with the analytically calculated resonance frequencies and eigenforms. A good correspondence was found. In the course of these comparisons, information about the modelling of the supporting structures and the efficiency of the damping elements during normal operation was obtained. The vibration of the main coolant pumps was also continuously monitored. The pump surveillance system for each pump includes two non-contacting displacement sensors for measuring the kinetic shaft orbit, as well as velocity sensors for recording the vibrational velocity of the pump motor housing. During the continuous monitoring, it was checked whether the signal amplitudes remained within the allowable limits. In addition the frequency content of the signals was determined periodically. In this way deviations could be detected immediately and be explained by means of subsequent correlation analysis. Thus amplitude changes resulting from resonance effects were identified. (orig.)

  10. Support structure for reactor core constituent element

    International Nuclear Information System (INIS)

    Aida, Yasuhiko.

    1993-01-01

    A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

  11. Study on dinamic behavior and least burying depth of underground protective pipe

    Energy Technology Data Exchange (ETDEWEB)

    Kataoka, Tetsuyuki; Kokusyo, Goji; Tanaka, Yukihisa; Kobayashi, Seiichi

    1988-03-30

    Effect of unit load per travel wheel on the protective pipe was studied when electricity cable distribution lines were buried in the depth less than the present standard for electric equipment, and logical burying depth was investigated. Test items were material test of the protective pipe, indoor load test, and field test at loamy ground. Impact resistance hard PVC pipe was used as the protective pipe, and its strength and elastic modulus were measured. Along with these tests, it was confirmed that there was no problem of cracking by repeated flattening or breakage by fatigue. By indoor test, it was observed that, in case of shallow burying, creap deformation was small, stress concentrate occured at the middle of axial direction, and that flattening ratio was seriously affected by the method of backfilling. Field test was conducted by applying the static load of a 20 ton dump truck, and the deformation, stress, and subsidence of the protective pipe, were measured. As the conclusion of those experiments, it was found that burying of protective pipe in the depth of not less than 30 cm is allowable, as long as sufficient bakfilling is made. (14 figs, 3 tabs, 3 refs)

  12. Specialist meeting on leak before break in reactor piping and vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bartholome, G.; Bazant, E.; Wellein, R. [Siemens KWU, Stuttgart (Germany)] [and others

    1997-04-01

    A series of research projects sponsored by the Federal Minister for Education, Science, Research and Technology, Bonn are summarized and compared to utility, manufacturer, and vendor tests. The purpose of the evaluation was to experimentally verify Leak-before-Break behavior, confirm the postulation of fracture preclusion for piping (straight pipe, bends and branches), and quantify the safety margin against massive failure. The results are applicable to safety assessment of ferritic and austenitic piping in primary and secondary nuclear power plant circuits. Moreover, because of the wide range of the test parameters, they are also important for the design and assessment of piping in other technical plant. The test results provide justification for ruling out catastrophic fractures, even on pipes of dimensions corresponding to those of a main coolant pipe of a pressurized water reactor plant on the basis of a mechanical deterministic safety analysis in correspondence with the Basis Safety Concept (Principle of Fracture Exclusion).

  13. A p-version embedded model for simulation of concrete temperature fields with cooling pipes

    Directory of Open Access Journals (Sweden)

    Sheng Qiang

    2015-07-01

    Full Text Available Pipe cooling is an effective method of mass concrete temperature control, but its accurate and convenient numerical simulation is still a cumbersome problem. An improved embedded model, considering the water temperature variation along the pipe, was proposed for simulating the temperature field of early-age concrete structures containing cooling pipes. The improved model was verified with an engineering example. Then, the p-version self-adaption algorithm for the improved embedded model was deduced, and the initial values and boundary conditions were examined. Comparison of some numerical samples shows that the proposed model can provide satisfying precision and a higher efficiency. The analysis efficiency can be doubled at the same precision, even for a large-scale element. The p-version algorithm can fit grids of different sizes for the temperature field simulation. The convenience of the proposed algorithm lies in the possibility of locating more pipe segments in one element without the need of so regular a shape as in the explicit model.

  14. Experimental and analytical studies on creep failure of reactor coolant piping

    Energy Technology Data Exchange (ETDEWEB)

    Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakamura, N.

    1999-07-01

    Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

  15. Experimental and analytical studies on creep failure of reactor coolant piping

    International Nuclear Information System (INIS)

    Maeda, Akio; Maruyama, Yu; Hashimoto, Kazuichiro; Harada, Yuhei; Shibazaki, Hiroaki; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun; Nakamura, N.

    1999-01-01

    Thermal and structural responses of reactor coolant piping under and elevated internal pressure and temperature are being investigated in WIND project at JAERI. In a recent failure test in which a nuclear grade type 316 stainless steel pipe with an outer diameter of 114.3 mm and a wall thickness of 13.5 mm was used and an internal pressure was kept at approximately 15 MPa. A failure of the piping was observed when the temperature was sustained at 970degC for one hour. In parallel with conducting the tests, post-test analyses were performed. The objective of the analyses is to assess analytical models for the creep deformation and failure of the piping at elevated internal pressure and temperature simulating thermal-hydraulic conditions during a severe accident. The major material properties needed for the analysis were measured at elevated temperatures. Coefficients of a creep constitutive equation including the tertiary stage were determined with the measured creep data and incorporated into ABAQUS code. The analysis reasonably reproduced the time history of the enlargement of the piping diameter, and the wall thickness and the diameter of the piping at the failure. It was also found that the piping failure timing obtained from the analysis agreed well with the test result. (author)

  16. Test on large-scale seismic isolation elements, 2

    International Nuclear Information System (INIS)

    Mazda, T.; Moteki, M.; Ishida, K.; Shiojiri, H.; Fujita, T.

    1991-01-01

    Seismic isolation test program of Central Research Inst. of Electric Power Industry (CRIEPI) to apply seismic isolation to Fast Breeder Reactor (FBR) plant was started in 1987. In this test program, demonstration test of seismic isolation elements was considered as one of the most important research items. Facilities for testing seismic isolation elements were built in Abiko Research Laboratory of CRIEPI. Various tests of large-scale seismic isolation elements were conducted up to this day. Many important test data to develop design technical guidelines was obtained. (author)

  17. Gravity-assist heat pipes for thermal control systems

    International Nuclear Information System (INIS)

    Deverall, J.E.; Keddy, E.S.; Kemme, J.E.; Phillips, J.R.

    1975-06-01

    Sodium heat pipes, operating in the gravity-assist mode, have been incorporated into irradiation capsules to provide a means for establishing and controlling a desired specimen temperature. Investigations were made of new wick structures for potassium heat pipes to operate at lower temperatures and higher heat transfer rates, and a helical trough wick structure was developed with an improved heat transfer capability in the temperature range of interest. Test results of these heat pipes led to the study of a new heat pipe limit which had not previously been considered. (12 references) (U.S.)

  18. Conceptual design of pipe whip restraints using interactive computer analysis

    International Nuclear Information System (INIS)

    Rigamonti, G.; Dainora, J.

    1975-01-01

    Protection against pipe break effects necessitates a complex interaction between failure mode analysis, piping layout, and structural design. Many iterations are required to finalize structural designs and equipment arrangements. The magnitude of the pipe break loads transmitted by the pipe whip restraints to structural embedments precludes the application of conservative design margins. A simplified analytical formulation of the nonlinear dynamic problems associated with pipe whip has been developed and applied using interactive computer analysis techniques. In the dynamic analysis, the restraint and the associated portion of the piping system, are modeled using the finite element lumped mass approach to properly reflect the dynamic characteristics of the piping/restraint system. The analysis is performed as a series of piecewise linear increments. Each of these linear increments is terminated by either formation of plastic conditions or closing/opening of gaps. The stiffness matrix is modified to reflect the changed stiffness characteristics of the system and re-started using the previous boundary conditions. The formation of yield hinges are related to the plastic moment of the section and unloading paths are automatically considered. The conceptual design of the piping/restraint system is performed using interactive computer analysis. The application of the simplified analytical approach with interactive computer analysis results in an order of magnitude reduction in engineering time and computer cost. (Auth.)

  19. Nuclear piping and pipe support design and operability relating to loadings and small bore piping

    International Nuclear Information System (INIS)

    Stout, D.H.; Tubbs, J.M.; Callaway, W.O.; Tang, H.T.; Van Duyne, D.A.

    1994-01-01

    The present nuclear piping system design practices for loadings, multiple support design and small bore piping evaluation are overly conservative. The paper discusses the results developed for realistic definitions of loadings and loading combinations with methodology for combining loads under various conditions for supports and multiple support design. The paper also discusses a simplified method developed for performing deadweight and thermal evaluations of small bore piping systems. Although the simplified method is oriented towards the qualification of piping in older plants, this approach is applicable to plants designed to any edition of the ASME Section III or B31.1 piping codes

  20. Propagation of ovalization along straight pipes and elbows

    International Nuclear Information System (INIS)

    Millard, A.; Roche, R.L.

    1981-01-01

    The aim of this paper is to present analytical solutions for the propagation of evalization and the variation of the flexibility factor along pipe bends terminated by straight pipes or flanges, under in-plane bending, assuming an elastic material behaviour. The influence of the various strains in analysed in the simple case of a straight pipe, subjected to an elliptical cross-section shape deformation at one end. The results enlighten the very important part played by the distorsion in the propagation. They have been compared with finite elements solutions and with simple experiments. The solution is developed for an elbow terminated by a straight pipe or a flange, following the Von Karman's approach: local displacements are expanded in Fourier series, the coefficients of which vary along the curvilinear abscissa, like the rotation of the cross-section as a whole; the differential equations as well as the boundary conditions are found by minimization of the total potential energy of the assembly. The solutions are compared to existing and experimental results. (orig./HP)

  1. Static analysis of a piping system with elbows

    International Nuclear Information System (INIS)

    Bryan, B.J.

    1994-01-01

    Vibration tests of elbows to failure were performed in Japan in the early 1970s. The piping system included two elbows and an eccentric mass. Tests were run both pressurized and unpressurized. This report documents a static analysis of the piping system in which the elbows are subjected to out of plane bending. The effects of internal pressure and material plasticity are investigated

  2. Environmental Assisted Fatigue Evaluation of Direct Vessel Injection Piping Considering Thermal Stratification

    International Nuclear Information System (INIS)

    Kim, Taesoon; Lee, Dohwan

    2016-01-01

    As the environmentally assisted fatigue (EAF) due to the primary water conditions is to be a critical issue, the fatigue evaluation for the components and pipes exposed to light water reactor coolant conditions has become increasingly important. Therefore, many studies to evaluate the fatigue life of the components and pipes in LWR coolant environments on fatigue life of materials have been conducted. Among many components and pipes of nuclear power plants, the direct vessel injection piping is known to one of the most vulnerable pipe systems because of thermal stratification occurred in that systems. Thermal stratification occurs because the density of water changes significantly with temperature. In this study, fatigue analysis for DVI piping using finite element analysis has been conducted and those results showed that the results met design conditions related with the environmental fatigue evaluation of safety class 1 pipes in nuclear power plants. Structural and fatigue integrity for the DVI piping system that thermal stratification occurred during the plant operation has conducted. First of all, thermal distribution of the piping system is calculated by computational fluid dynamic analysis to analyze the structural integrity of that piping system. And the fatigue life evaluation considering environmental effects was carried out. Our results showed that the DVI piping system had enough structural integrity and fatigue life during the design lifetime of 60 years

  3. Performance predictions and measurements for space-power-system heat pipes

    International Nuclear Information System (INIS)

    Prenger, F.C. Jr.

    1981-01-01

    High temperature liquid metal heat pipes designed for space power systems have been analyzed and tested. Three wick designs are discussed and a design rationale for the heat pipe is provided. Test results on a molybdenum, annular wick heat pipe are presented. Performance limitations due to boiling and capillary limits are presented. There is evidence that the vapor flow in the adiabatic section is turbulent and that the transition Reynolds number is 4000

  4. Vacuum pipe for e+e- interactions

    International Nuclear Information System (INIS)

    Hoard, C.T.

    1982-10-01

    The design, fabrication and testing of the beryllium vacuum chamber within the Mark II detector at SLAC is described. The Be chamber encloses one interaction point of the PEP circulating ring and is a part of its beam pipe. The Be chamber is captured within the Secondary Vertex Detector (SVD), a drift chamber, which is in turn centered in the Mark II drift chamber. Both ends of the beryllium pipe are brazed to aluminum/stainless transitions for connection to stainless steel bellows. A concentric radiation-screen liner of titanium foil runs the full length of the beryllium pipe

  5. An evaluation of detection ability of ultrasonic testing with a large aperture transducer for axial cracks in cast stainless steel pipe welds

    International Nuclear Information System (INIS)

    Nishikawa, Yoshito; Ishida, Hitoshi; Kurozumi, Yasuo

    2013-01-01

    Ultrasonic testing is difficult to apply to cast stainless steel which is the material of the main coolant pipes in pressurized water reactors, because of the large attenuation and scattering of ultrasonic waves caused by its macro structure. In this study, ultrasonic testing for progression of axial fatigue cracks of a welded area in the test piece of cast stainless steel pipe was performed using double big-size ultrasonic probes which were formerly developed in INSS. It was found that detection of defects that were over 6% of the target depth for the specimen thickness of 69mm is possible, and detection of defects with over 10% of the target depth is possible for all test conditions. (author)

  6. EXPERIMENTAL AND NUMERICAL INVESTIGATION OF FLEXIBLE BURIED PIPE DEFORMATION BEHAVIOR UNDER VARIOUS BACKFILL CONDITIONS

    Directory of Open Access Journals (Sweden)

    Niyazi Uğur TERZİ

    2009-01-01

    Full Text Available Deformation characteristics of polyethylene based flexible pipes are different than rigid pipes such as concrete and iron pipes. Deflection patterns and stress-strain behaviors of flexible pipes have strict relation between the engineering properties of backfill and its settlement method. In this study, deformation behavior of a 100 mm HDPE flexible pipe under vertical loads is investigated in laboratory conditions. Steel test box, pressurized membrane, raining system, linear position transducers and strain gauge rosettes are used in the laboratory tests. In order to analyze the buried pipe performance; Masada Derivation Formula which is mostly used by designers is employed. According to the test and mathematical studies, it is understood that relative density of backfill and its settlement method is a considerable effect on buried pipe performance and Masada Derivation method is very efficient for predicting the pipe performance.

  7. Geochemical soil sampling for deeply-buried mineralized breccia pipes, northwestern Arizona

    Science.gov (United States)

    Wenrich, K.J.; Aumente-Modreski, R. M.

    1994-01-01

    Thousands of solution-collapse breccia pipes crop out in the canyons and on the plateaus of northwestern Arizona; some host high-grade uranium deposits. The mineralized pipes are enriched in Ag, As, Ba, Co, Cu, Mo, Ni, Pb, Sb, Se, V and Zn. These breccia pipes formed as sedimentary strata collapsed into solution caverns within the underlying Mississippian Redwall Limestone. A typical pipe is approximately 100 m (300 ft) in diameter and extends upward from the Redwall Limestone as much as 1000 m (3000 ft). Unmineralized gypsum and limestone collapses rooted in the Lower Permian Kaibab Limestone or Toroweap Formation also occur throughout this area. Hence, development of geochemical tools that can distinguish these unmineralized collapse structures, as well as unmineralized breccia pipes, from mineralized breccia pipes could significantly reduce drilling costs for these orebodies commonly buried 300-360 m (1000-1200 ft) below the plateau surface. Design and interpretation of soil sampling surveys over breccia pipes are plagued with several complications. (1) The plateau-capping Kaibab Limestone and Moenkopi Formation are made up of diverse lithologies. Thus, because different breccia pipes are capped by different lithologies, each pipe needs to be treated as a separate geochemical survey with its own background samples. (2) Ascertaining true background is difficult because of uncertainties in locations of poorly-exposed collapse cones and ring fracture zones that surround the pipes. Soil geochemical surveys were completed on 50 collapse structures, three of which are known mineralized breccia pipes. Each collapse structure was treated as an independent geochemical survey. Geochemical data from each collapse feature were plotted on single-element geochemical maps and processed by multivariate factor analysis. To contrast the results between geochemical surveys (collapse structures), a means of quantifying the anomalousness of elements at each site was developed. This

  8. Vibration analysis of the piping system using the modal analysis method, 1

    International Nuclear Information System (INIS)

    Fujikawa, Takeshi; Kurohashi, Michiya; Inoue, Yoshio

    1975-01-01

    Modal analysis method was developed for the vibration analysis of piping system in nuclear or chemical plants, with finite element theory, and verified by sinusoidal vibration method. The natural vibration equation for pipings was derived with stiffness, attenuation and mass matrices, and eigenvalues are obtained with usual method, then the forced vibration equation for pipings was derived with the same manner, and the special solutions are given by modal method from the eigenvalues of the natural vibration equation. Three simple piping models (one, two and three dimensional) were made, and the natural vibration frequency was measured with forced input from an electrical dynamic shaker and a sound speaker. The experimental values of natural vibration frequency showed good agreement with the results by the analytical method. Therefore the theoretical approach for piping system vibration was proved to be valid. (Iwase, T.)

  9. Experimental Investigation on Corrosion of Cast Iron Pipes

    Directory of Open Access Journals (Sweden)

    H. Mohebbi

    2011-01-01

    Full Text Available It is well known that corrosion is the predominant mechanism for the deterioration of cast iron pipes, leading to the reduction of pipe capacity and ultimate collapse of the pipes. In order to assess the remaining service life of corroded cast iron pipes, it is imperative to understand the mechanisms of corrosion over a long term and to develop models for pipe deterioration. Although many studies have been carried out to determine the corrosion behavior of cast iron, little research has been undertaken to understand how cast iron pipes behave over a longer time scale than hours, days, or weeks. The present paper intends to fill the gap regarding the long-term corrosion behaviour of cast iron pipes in the absence of historical data. In this paper, a comprehensive experimental program is presented in which the corrosion behaviour of three exservice pipes was thoroughly examined in three simulated service environments. It has been found in the paper that localised corrosion is the primary form of corrosion of cast iron water pipes. It has also been found that the microstructure of cast irons is a key factor that affects the corrosion behaviour of cast iron pipes. The paper concludes that long-term tests on corrosion behaviour of cast iron pipes can help develop models for corrosion-induced deterioration of the pipes for use in predicting the remaining service life of the pipes.

  10. Assessment of water pipes durability under pressure surge

    Science.gov (United States)

    Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

    2017-10-01

    Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

  11. Influence of the pipe diameter on the structure of the gas-liquid interface in a vertical two-phase pipe flow

    International Nuclear Information System (INIS)

    Prasser, H. M.; Beyer, M.; Boettger, A.; Carl, H.; Lucas, D.; Schaffrath, A.; Schutz, P.; Weiss, F. P.; Zschau, J.

    2003-01-01

    Two-phase flow tests in a 194.1 mm diameter vertical pipe (DN200) with an air-water mixture are reported. Close to the upper end of a 9 m tall test section a wire-mesh sensor is installed that delivers instantaneous void fraction distributions over the entire cross section with time resolution of 2500 frames per second. The sensor disposes of 64 x 64 measuring points, which corresponds to a spatial resolution of 3 mm. Beside an fast flow visualisations, void-fraction profiles and bubble size distributions were obtained. Earlier, similar experiments were carried out in a pipe of 51.2 mm inner diameter (DN50). A comparison of the data from the two different facilities allows to study the scaling effects on void fraction profiles, bubbles size distributions and the flow patterns. In the small pipe, the increase of the air flow rate leads to a transition from bubbly via slug to churn turbulent flow. The transition to slug flow is accompanied by the appearance of a second peak in the bubble size distribution that corresponds to the class of large Taylor bubbles. A similar qualitative behaviour was found in the large pipe, though the large bubble fraction has a significantly bigger mean diameter at identical superficial velocities, the peak is less tall but wider. Bubbles move more freely than in the small pipe, since the confining action of the pipe walls to the flow is less pronounced, while the large Taylor bubbles occupy almost the entire cross section in case of the small pipe. Furthermore, the bubbles show much more deformations in the large pipe. Shapes of such large bubbles were characterised in three dimensions for the first time. They can rather be complicated and far from the shape of ideal Taylor bubbles. Also the small bubble fraction tends to bigger sizes in the large pipe

  12. Finite element analysis of the collapse and post-collapse behavior of steel pipes applications to the oil industry

    CERN Document Server

    Dvorkin, Eduardo N

    2013-01-01

    This book presents a detailed discussion of the models that were developed to simulate the collapse and post-collapse behavior of steel pipes. The finite element method offers to engineers the possibility of developing models to simulate the collapse behavior of casings inside oil wells and the collapse behavior of deepwater pipelines. However, if technological decisions are going to be reached from these model results, with implications for the economic success of industrial operations, for the occupational safety and health and for the environment, the engineering models need to be highly reliable. Using these models engineers can quantify the effect of manufacturing tolerances, wear, corrosion, etc. This book describes in great details the experimental programs that are developed to validate the numerical results.

  13. Leakage Detection and Estimation Algorithm for Loss Reduction in Water Piping Networks

    Directory of Open Access Journals (Sweden)

    Kazeem B. Adedeji

    2017-10-01

    Full Text Available Water loss through leaking pipes constitutes a major challenge to the operational service of water utilities. In recent years, increasing concern about the financial loss and environmental pollution caused by leaking pipes has been driving the development of efficient algorithms for detecting leakage in water piping networks. Water distribution networks (WDNs are disperse in nature with numerous number of nodes and branches. Consequently, identifying the segment(s of the network and the exact leaking pipelines connected to this segment(s where higher background leakage outflow occurs is a challenging task. Background leakage concerns the outflow from small cracks or deteriorated joints. In addition, because they are diffuse flow, they are not characterised by quick pressure drop and are not detectable by measuring instruments. Consequently, they go unreported for a long period of time posing a threat to water loss volume. Most of the existing research focuses on the detection and localisation of burst type leakages which are characterised by a sudden pressure drop. In this work, an algorithm for detecting and estimating background leakage in water distribution networks is presented. The algorithm integrates a leakage model into a classical WDN hydraulic model for solving the network leakage flows. The applicability of the developed algorithm is demonstrated on two different water networks. The results of the tested networks are discussed and the solutions obtained show the benefits of the proposed algorithm. A noteworthy evidence is that the algorithm permits the detection of critical segments or pipes of the network experiencing higher leakage outflow and indicates the probable pipes of the network where pressure control can be performed. However, the possible position of pressure control elements along such critical pipes will be addressed in future work.

  14. Performance Measurements of a 7 mm-Diameter Hydrogen Heat Pipe

    International Nuclear Information System (INIS)

    Abdel-Bary, M.; Abdel-Samad, S.; Kiliana, K.; Ritman, J.; Abdel-Bary, M.; Abdel-Samad, S.

    2008-01-01

    A gravity assisted heat pipe with 7-mm diameter has been developed and tested to cool a liquid hydrogen target for extracted beam experiments at COSY. The liquid flowing down from the condenser surface is separated from the vapor flowing up by a thin wall 3 mm diameter plastic tube located concentrically inside the heat pipe. The heat pipe was tested at different inclination angles with respect to the horizontal plane. The heat pipe showed good operating characteristics because of the low radiation heat load from the surroundings, low heat capacity due to the small mass, higher sensitivity to heat loads (to overcome the heat load before the complete vaporization of the liquid in the target cell) due to the higher vapor speed inside the heat pipe which transfers the heat load to the condenser

  15. Fracture assessment of Savannah River Reactor carbon steel piping

    International Nuclear Information System (INIS)

    Mertz, G.E.; Stoner, K.J.; Caskey, G.R.; Begley, J.A.

    1991-01-01

    The Savannah River Site (SRS) production reactors have been in operation since the mid-1950's. One postulated failure mechanism for the reactor piping is brittle fracture of the original A285 and A53 carbon steel piping. Material testing of archival piping determined (1) the static and dynamic tensile properties; (2) Charpy impact toughness; and (3) the static and dynamic compact tension fracture toughness properties. The nil-ductility transition temperature (NDTT), determined by Charpy impact test, is above the minimum operating temperature for some of the piping materials. A fracture assessment was performed to demonstrate that potential flaws are stable under upset loading conditions and minimum operating temperatures. A review of potential degradation mechanisms and plant operating history identified weld defects as the most likely crack initiation site for brittle fracture. Piping weld defects, as characterized by radiographic and metallographic examination, and low fracture toughness material properties were postulated at high stress locations in the piping. Normal operating loads, upset loads, and residual stresses were assumed to act on the postulated flaws. Calculated allowable flaw lengths exceed the size of observed weld defects, indicating adequate margins of safety against brittle fracture. Thus, a detailed fracture assessment was able to demonstrate that the piping systems will not fail by brittle fracture, even though the NDTT for some of the piping is above the minimum system operating temperature

  16. Evaluation of seismic margins for an in-plant piping system

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    Earthquake experience as well as experiments indicate that, in general, piping systems are quite rugged in resisting seismic loadings. Therefore there is a basis to hold that the seismic margin against pipe failure is very high for systems designed according to current practice. However, there is very little data, either from tests or from earthquake experience, on the actual margin or excess capacity (against failure from seismic loading) of in-plant piping systems. Design of nuclear power plant piping systems in the US is governed by the criteria given in the ASME Boiler and Pressure Vessel (B ampersand PV) Code, which assure that pipe stresses are within specified allowable limits. Generally linear elastic analytical methods are used to determine the stresses in the pipe and forces in pipe supports. The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. Note that in the present context, seismic margin refers to the deterministic excess capacities of piping or supports compared to their design capacities. The excess seismic capacities or margins of a prototypical in-plant piping system and its components are evaluated by comparing measured inputs and responses from high-level simulated seismic experiments with design loads and allowables. Large excess capacities are clearly demonstrated against pipe and overall system failure with the lower bound being about four. For snubbers the lower bound margin is estimated at two and for rigid strut supports at five. 4 refs., 2 figs., 2 tabs

  17. Pressure and Temperature of the Room 1 for the Pipe Break Accidents of the 3-Pin Fuel Test Loop

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. K.; Chi, D. Y.; Sim, B. S.; Park, K. N.; Ahn, S. H.; Lee, J. M.; Lee, C. Y.; Kim, H. R

    2005-08-15

    This report deals with the prediction of the pressure and temperature of the room 1 for the pipe break accidents of the 3-pin fuel test loop. The 3-pin fuel test loop is an experimental facility for nuclear fuel tests at the operation conditions similar to those of PWR and CANDU power plants. Because the most processing systems of the 3-pin fuel test loop are placed in the room 1. The structural integrity of the room 1 should be evaluated for the postulated accident conditions. Therefore the pressures and temperatures of the room 1 needed for the structural integrity evaluation have been calculated by using MARS code. The pressures and temperatures of the room 1 have been calculated in various conditions such as the thermal hydraulic operation parameters, the locations of pipe break, and the thermal properties of the room 1 wall. It is assumed that the pipe break accident occurs in the letdown operation without regeneration, because the mass and energy release to the room 1 is expected to be the largest. As a result of the calculations the maximum pressure and temperature are predicted to be 208kPa and 369.2K(96.0 .deg. C) in case the heat transfer is considered in the room 1 wall. However the pressure and temperature are asymptotically 243kPa and 378.1K(104.9 .deg. C) assuming that the heat transfer does not occur in the room 1 wall.

  18. Finite element evaluation of erosion/corrosion affected reducing elbow

    International Nuclear Information System (INIS)

    Basavaraju, C.

    1996-01-01

    Erosion/corrosion is a primary source for wall thinning or degradation of carbon steel piping systems in service. A number of piping failures in the power industry have been attributed to erosion/corrosion. Piping elbow is one of such susceptible components for erosion/corrosion because of increased flow turbulence due to its geometry. In this paper, the acceptability of a 12 in. x 8 in. reducing elbow in RHR service water pump discharge piping, which experienced significant degradation due to wall thinning in localized areas, was evaluated using finite element analysis methodology. Since the simplified methods showed very small margin and recommended replacement of the elbow, a detailed 3-D finite element model was built using shell elements and analyzed for internal pressure and moment loadings. The finite element analysis incorporated the U.T. measured wall thickness data at various spots that experienced wall thinning. The results showed that the elbow is acceptable as-is until the next fuel cycle. FEA, though cumbersome, and time consuming is a valuable analytical tool in making critical decisions with regard to component replacement of border line situation cases, eliminating some conservatism while not compromising the safety

  19. Novel developments in linear modal description of piping system dynamic behavior

    International Nuclear Information System (INIS)

    Revesz, Z.

    1989-01-01

    Novel developments in dynamic analysis of piping systems are described. The ASME BPV Codes, 1986 describes methods that are considered as adequate to analyze piping systems under dynamic loading, and also states that the method described in the codes are not the only acceptable ones. With straightforward application of the principles and methods laid down in the code novel numerical techniques can be developed. These techniques allow to obtain correct, conservative estimates of the piping system response and to reduce the computed stresses the same time. Beyond that, the particular algorithm which is presented is also suitable to analyze systems which include non-linear (viscous) damping elements

  20. The Cryogenic Test Bed experiments: Cryogenic heat pipe flight experiment CRYOHP (STS-53). Cryogenic two phase flight experiment CRYOTP (STS-62). Cryogenic flexible diode flight experiment CRYOFD

    Science.gov (United States)

    Thienel, Lee; Stouffer, Chuck

    1995-09-01

    This paper presents an overview of the Cryogenic Test Bed (CTB) experiments including experiment results, integration techniques used, and lessons learned during integration, test and flight phases of the Cryogenic Heat Pipe Flight Experiment (STS-53) and the Cryogenic Two Phase Flight Experiment (OAST-2, STS-62). We will also discuss the Cryogenic Flexible Diode Heat Pipe (CRYOFD) experiment which will fly in the 1996/97 time frame and the fourth flight of the CTB which will fly in the 1997/98 time frame. The two missions tested two oxygen axially grooved heat pipes, a nitrogen fibrous wick heat pipe and a 2-methylpentane phase change material thermal storage unit. Techniques were found for solving problems with vibration from the cryo-collers transmitted through the compressors and the cold heads, and mounting the heat pipe without introducing parasitic heat leaks. A thermally conductive interface material was selected that would meet the requirements and perform over the temperature range of 55 to 300 K. Problems are discussed with the bi-metallic thermostats used for heater circuit protection and the S-Glass suspension straps originally used to secure the BETSU PCM in the CRYOTP mission. Flight results will be compared to 1-g test results and differences will be discussed.

  1. Application of ultrasonic testing technique to detect gas accumulation in important pipings for pressurized water reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Fushimi, Yasuyuki [Inst. of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2002-09-01

    Since 1988, the USNRC has pointed out that gas-binding events might occur at high head safety injection (HHSI) pumps of pressurized water reactors (PWRs). In Japanese PWR plants, corrective actions were taken in response to gas-binding events that occurred on HHSI pumps in the USA, so no gas accumulation event has been reported so far. However, when venting frequency is prolonged with operating cycle extension, the probability of gas accumulation in pipings may increase as in the USA. The purpose of this study was to establish a technique to identify gas accumulation and to measure the gas volume accurately. Taking dominant causes of the gas-binding events in the USA into consideration, we pointed out the following sections in the Japanese PWRs where gas srtipping and/or gas accumulation might occur: residual heat removal system pipings and charging/safety injection pump minimum flow line. Then an ultrasonic testing technique, adopted to identify gas accumulation in the USA, was applied to those sections of the typical Japanese PWR. Consequently, no gas accumulation was found in those pipings. (author)

  2. Margins for an in-plant piping system under dynamic loading

    International Nuclear Information System (INIS)

    Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

    1991-01-01

    The objective of this study is to verify that piping designed according to current practice does indeed have a large margin against failure and to quantify the excess capacity for piping and dynamic pipe supports on the basis of data obtained in a series of high-level seismic experiments (designated SHAM) on an in-plant piping system at the HDR (Heissdampfreaktor) Test Facility in Germany. 4 refs., 6 tabs

  3. Evaluation methods of vibration stress of small bore piping

    Energy Technology Data Exchange (ETDEWEB)

    Hiramatsu, Miki; Sasaki, Toru [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    Fatigue fracture by vibration stress is one of the main causes of troubles which occur at small bore piping in nuclear power plants. Therefore at the plants they manage small bore piping using a method in which their vibration accelerations are measured and the vibration stresses are calculated. In this work, vibration tests for two sets of mock-ups simulating actual piping in the plants by sinusoidal oscillation and by that obtained at an actual plant were carried out, and then an evaluation method was developed to obtain proper value of vibration stress from the measured data by the vibration tests. In comparison of the vibration stress obtained from the measured acceleration with that directly measured using strain gauges, it is confirmed that accurate vibration stress can be evaluated by a formula in which the real center of gravity of small bore piping and the acceleration of main (system) piping are considered. (author)

  4. Detecting pipe bursts by monitoring water demand

    NARCIS (Netherlands)

    Bakker, M.; Vreeburg, J.H.G.; Van der Roer, M.; Sperber, V.

    2012-01-01

    An algorithm which compares measured and predicted water demands to detect pipe bursts was developed and tested on three data sets of water demand and reported pipe bursts of three years. The algorithm proved to be able to detect bursts where the water loss exceeds 30% of the average water demand in

  5. A contribution for stress analysis in bend acessories of piping systems

    International Nuclear Information System (INIS)

    Melo, F.J.M.Q. de; Castro, P.M.S.T. de

    1986-01-01

    Analytical and numerical studies of the linear elastic behavior of bend pipes, with tangent pipes or flanged ends, such as used in nuclear power plants are presented. Two analytical techniques were developed; one is based on the integration of Euler equation and the other one is based on a Fourier analysis. The results obtained using these approaches are compared with results obtained by a finite element code for 'semiloof shells. (Author) [pt

  6. Using a heat pipe (TPTC for dissipating energy generated by an electronic circuit

    Directory of Open Access Journals (Sweden)

    Rodrigo Correa

    2010-01-01

    Full Text Available This paper presents an experimental investigation aimed at estimating the thermal efficiency of a heat pipe compared to the most common elements for removing heat from a circuit (i.e., an electric fan and a fin - extended surface. The input voltage frequency for a standard power circuit was changed for the experiments, whilst all the other parameters were kept constant. An experimental statistical design was used as an analytical tool. Unexpectedly, the heat pipe showed the lowest thermal efficiency for all the experiments, although it had the advantage of being a passive element having low volume and no mobile parts.

  7. Nonlinear piping damping and response predictions

    International Nuclear Information System (INIS)

    Severud, L.K.; Weiner, E.O.; Lindquist, M.R.; Anderson, M.J.; Wagner, S.E.

    1986-10-01

    The high level dynamic testing of four prototypic piping systems, used to provide benchmarks for analytical prediction comparisons, is overviewed. The size of pipe tested ranged from one-inch to six-inches in diameter and consisted of carbon steel or stainless steel material. Failure of the tested systems included progressive gross deformation or some combination of ratchetting-fatigue. Pretest failure predictions and post test comparisons using simplified elastic and elasto-plastic methods are presented. Detailed non-linear inelastic analyses are also shown, along with a typical ratchet-fatigue failure calculation. A simplified method for calculating modal equivalent viscous damping for snubbers and plastic hinges is also described. Conclusions are made regarding the applicability of the various analytical failure predictive methods and recommendations are made for future analytic and test efforts

  8. HPFRCC - Extruded Pipes

    DEFF Research Database (Denmark)

    Stang, Henrik; Pedersen, Carsten

    1996-01-01

    The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique development......-w$ relationship is presented. Structural development involved definition of a new type of semi-flexiblecement based pipe, i.e. a cement based pipe characterized by the fact that the soil-pipe interaction related to pipe deformation is an importantcontribution to the in-situ load carrying capacity of the pipe...

  9. Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki [and others

    1997-04-01

    The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program.

  10. Effect of dynamic monotonic and cyclic loading on fracture behavior for Japanese carbon steel pipe STS410

    International Nuclear Information System (INIS)

    Kinoshita, Kanji; Murayama, Kouichi; Ogata, Hiroyuki

    1997-01-01

    The fracture behavior for Japanese carbon steel pipe STS410 was examined under dynamic monotonic and cyclic loading through a research program of International Piping Integrity Research Group (EPIRG-2), in order to evaluate the strength of pipe during the seismic event The tensile test and the fracture toughness test were conducted for base metal and TIG weld metal. Three base metal pipe specimens, 1,500mm in length and 6-inch diameter sch.120, were employed for a quasi-static monotonic, a dynamic monotonic and a dynamic cyclic loading pipe fracture tests. One weld joint pipe specimen was also employed for a dynamic cyclic loading test In the dynamic cyclic loading test, the displacement was controlled as applying the fully reversed load (R=-1). The pipe specimens with a circumferential through-wall crack were subjected four point bending load at 300C in air. Japanese STS410 carbon steel pipe material was found to have high toughness under dynamic loading condition through the CT fracture toughness test. As the results of pipe fracture tests, the maximum moment to pipe fracture under dynamic monotonic and cyclic loading condition, could be estimated by plastic collapse criterion and the effect of dynamic monotonic loading and cyclic loading was a little on the maximum moment to pipe fracture of the STS410 carbon steel pipe. The STS410 carbon steel pipe seemed to be less sensitive to dynamic and cyclic loading effects than the A106Gr.B carbon steel pipe evaluated in IPIRG-1 program

  11. Main steam system piping response under safety/relief valve opening events

    International Nuclear Information System (INIS)

    Swain, E.O.; Esswein, G.A.; Hwang, H.L.; Nieh, C.T.

    1980-01-01

    The stresses in the main steam branch pipe of a Boiling Water Reactor due to safety/relief valve blowdown has been measured from an in situ piping system test. The test results were compared with analytical results. The predicted stresses using the current state of art analytical methods used for BWR SRV discharge transient piping response loads were found to be conservative when compared to the measured stress values. 3 refs

  12. Flexible heat pipes with integrated bioinspired design

    Directory of Open Access Journals (Sweden)

    Chao Yang

    2015-02-01

    Full Text Available In this work we report the facile fabrication and performance evaluation of flexible heat pipes that have integrated bioinspired wick structures and flexible polyurethane polymer connector design between the copper condenser and evaporator. Inside the heat pipe, a bioinspired superhydrophilic strong-base-oxidized copper mesh with multi-scale micro/nano-structures was used as the wicking material and deionized water was selected as working fluid. Thermal resistances of the fabricated flexible heat pipes charged with different filling ratios were measured under thermal power inputs ranging from 2 W to 12 W while the device was bent at different angles. The fabricated heat pipes with a 30% filling ratio demonstrated a low thermal resistance less than 0.01 K/W. Compared with the vertically oriented straight heat pipes, bending from 30° up to 120° has negligible influence on the heat-transfer performance. Furthermore, repeated heating tests indicated that the fabricated flexible heat pipes have consistent and reliable heat-transfer performance, thus would have important applications for advanced thermal management in three dimensional and flexible electronic devices.

  13. Seismic evaluation of piping systems using screening criteria

    International Nuclear Information System (INIS)

    Campbell, R.D.; Landers, D.F.; Minichiello, J.C.; Slagis, G.C.; Antaki, G.A.

    1994-01-01

    This document may be used by a qualified review team to identify potential sources of seismically induced failure in a piping system. Failure refers to the inability of a piping system to perform its expected function following an earthquake, as defined in Table 1. The screens may be used alone or with the Seismic Qualification Utility Group -- Generic Implementation Procedure (SQUG-GIP), depending on the piping system's required function, listed in Table 1. Features of a piping system which do not the screening criteria are called outliers. Outliers must either be resolved through further evaluations, or be considered a potential source of seismically induced failure. Outlier evaluations, which do not necessarily require the qualification of a complete piping system by stress analysis, may be based on one or more of the following: simple calculations of pipe spans, search of the test or experience data, vendor data, industry practice, etc

  14. Investigation of transient cavitating flow in viscoelastic pipes

    International Nuclear Information System (INIS)

    Keramat, A; Tijsseling, A S; Ahmadi, A

    2010-01-01

    A study on water hammer in viscoelastic pipes when the fluid pressure drops to liquid vapour pressure is performed. Two important concepts including column separation and the effects of retarded strains in the pipe wall on the fluid response have been investigated separately in recent works, but there is some curiosity as to how the results for pressure and discharge are when column separation occurs in a viscoelastic pipe. For pipes made of plastic such as polyethylene (PE) and polyvinyl chloride (PVC), viscoelasticity is a crucial mechanical property which changes the hydraulic and structural transient responses. Based on previous developments in the analysis of water hammer, a model which is capable of analysing column separation in viscoelastic pipes is presented and used for solving the selected case studies. For the column-separation modelling the Discrete Vapour Cavity Model (DVCM) is utilised and the viscoelasticity property of the pipe wall is modelled by Kelvin-Voigt elements. The effects of viscoelasticity play an important role in the column separation phenomenon because it changes the water hammer fundamental frequency and so affects the time of opening or collapse of the cavities. Verification of the implemented computer code is performed for the effects of viscoelasticity and column separation - separately and simultaneously - using experimental results from the literature. In the provided examples the focus is placed on the simultaneous effect of viscoelasticity and column separation on the hydraulic transient response. The final conclusions drawn are that if rectangular grids are utilised the DVCM gives acceptable predictions of the phenomenon and that the pipe wall material's retarded behaviour strongly dampens the pressure spikes caused by column separation.

  15. Investigation of transient cavitating flow in viscoelastic pipes

    Science.gov (United States)

    Keramat, A.; Tijsseling, A. S.; Ahmadi, A.

    2010-08-01

    A study on water hammer in viscoelastic pipes when the fluid pressure drops to liquid vapour pressure is performed. Two important concepts including column separation and the effects of retarded strains in the pipe wall on the fluid response have been investigated separately in recent works, but there is some curiosity as to how the results for pressure and discharge are when column separation occurs in a viscoelastic pipe. For pipes made of plastic such as polyethylene (PE) and polyvinyl chloride (PVC), viscoelasticity is a crucial mechanical property which changes the hydraulic and structural transient responses. Based on previous developments in the analysis of water hammer, a model which is capable of analysing column separation in viscoelastic pipes is presented and used for solving the selected case studies. For the column-separation modelling the Discrete Vapour Cavity Model (DVCM) is utilised and the viscoelasticity property of the pipe wall is modelled by Kelvin-Voigt elements. The effects of viscoelasticity play an important role in the column separation phenomenon because it changes the water hammer fundamental frequency and so affects the time of opening or collapse of the cavities. Verification of the implemented computer code is performed for the effects of viscoelasticity and column separation - separately and simultaneously - using experimental results from the literature. In the provided examples the focus is placed on the simultaneous effect of viscoelasticity and column separation on the hydraulic transient response. The final conclusions drawn are that if rectangular grids are utilised the DVCM gives acceptable predictions of the phenomenon and that the pipe wall material's retarded behaviour strongly dampens the pressure spikes caused by column separation.

  16. Experimental Investigation of Heat Pipe Startup Under Reflux Mode

    Science.gov (United States)

    Ku, Jentung

    2018-01-01

    In the absence of body forces such as gravity, a heat pipe will start as soon as its evaporator temperature reaches the saturation temperature. If the heat pipe operates under a reflux mode in ground testing, the liquid puddle will fill the entire cross sectional area of the evaporator. Under this condition, the heat pipe may not start when the evaporator temperature reaches the saturation temperature. Instead, a superheat is required in order for the liquid to vaporize through nucleate boiling. The amount of superheat depends on several factors such as the roughness of the heat pipe internal surface and the gravity head. This paper describes an experimental investigation of the effect of gravity pressure head on the startup of a heat pipe under reflux mode. In this study, a heat pipe with internal axial grooves was placed in a vertical position with different tilt angles relative to the horizontal plane. Heat was applied to the evaporator at the bottom and cooling was provided to the condenser at the top. The liquid-flooded evaporator was divided into seven segments along the axial direction, and an electrical heater was attached to each evaporator segment. Heat was applied to individual heaters in various combinations and sequences. Other test variables included the condenser sink temperature and tilt angle. Test results show that as long as an individual evaporator segment was flooded with liquid initially, a superheat was required to vaporize the liquid in that segment. The amount of superheat required for liquid vaporization was a function of gravity pressure head imposed on that evaporator segment and the initial temperature of the heat pipe. The most efficient and effective way to start the heat pipe was to apply a heat load with a high heat flux to the lowest segment of the evaporator.

  17. Ductile failure of pipes with defects under combined pressure and bending

    International Nuclear Information System (INIS)

    Darlaston, B.J.L.; Harrison, R.P.

    1977-01-01

    The main part of the experimental programme was carried out on 3.5'' diam. pipes with a wall thickness of 0.064''. Various lengths of defect were assessed but only two depths, 0.044'' and 0.060''. Some full penetration defect tests were carried out under bending loading. The defects were 0.012'' wide and nominally flat bottomed. The tensile properties of the pipes were determined by taking specimens from each of the tubes. The pipes were exposed to pressure only test, bending only test and combined bending and pressure test. The results are given in tables. The observations led to the postulation of a design rule relating to the effect of defect in pipes under combined internal pressure and bending. It applies only to ductile situations in which the mode of failure is by a collapse mechanism: If the failure of a pipe containing an axial defect occurs by plastic collapse then provided the bending moment does not exceed half that for collapse due to bending alone, it will have a negligible effect on the failure pressure. (J.B.)

  18. Mechanical Properties of Post Irradiation Primary Cooling Piping of Bandung Research Reactor

    International Nuclear Information System (INIS)

    Histori; Renaningsih S; Sri Nitiswati; Ari Triyadi

    2003-01-01

    Testing on primary coolant piping of research reactor Bandung have been done. Primary coolant piping were made from Al 6061-T6. The goal of this activity is to investigate the mechanical properties changes caused by aging process after 33 years in irradiated. Type of testing i.e visual examination, thickness measurement, tensile and hardness test were done. The test data shown that there was a deposit at the inside surface of pipe, thickness decreased about 0.2 mm, tensile strength is 293 MPa, yield strength is 262 MPa, while the hardness is about 83 HRE (mean value). The test data than compared with ASTM standard. As the conclusion tensile and yield strength of pipe still fulfill the ASTM requirements, except the hardness is unsignificantly less/decreased. (author)

  19. Inelastic analysis of piping systems. A beam-type method for creep and plasticity

    International Nuclear Information System (INIS)

    Roche, R.L.; Hoffmann, A.; Millard, A.

    1979-01-01

    Since many years, piping systems are designed and calculated under elasticity assumptions, using a beam-type method. Thus, the analysis of large systems may be performed at a relatively low cost, using a finite element program. However such a method can not account for inelastic phenomena like plastic deformations or creep. The application of refined three-dimensional shell type method is possible for local components such as curved sections but leads to prohibitive costs for complete piping systems. Therefore simplified methods have been developed, based on a 'global plasticity or creep model'. Following the conventional elastic approach, the pipe element is characterized by variables associated with the center line in the following way: generalized stresses are obtained by integration of local stresses giving way to hoop and tension stresses and to bending and torsional moments; the conjugated strains are identified with uniform hoop and longitudinal strains and variations in neutral axis curvatuves. For plasticity problems, the yield surface is defined by a diagonal quadratic function in terms of the generalized stresses and work hardening parameters. By addition of the Hill's principle and a hardening rule, the formulation is similar to the one commonly used in finite element method. Geometric non linearity due to important deformations of the cross section (often termed 'ovalization') may be treated simultaneously with material non linearity. For this purpose the displacement normal to the pipe surface is represented by trigonometric series expansion, the coefficients of which are determined by minimizing the strain energy over the cross section. The method presented is believed to be a simple economical and accurate tool, for dimensioning computations of large piping systems

  20. Structural and stress analysis of nuclear piping systems

    International Nuclear Information System (INIS)

    Hata, Hiromichi

    1982-01-01

    The design of the strength of piping system is important in plant design, and its outline on the example of PWRs is reported. The standards and guides concerning the design of the strength of piping system are shown. The design condition for the strength of piping system is determined by considering the requirements in the normal operation of plants and for the safety design of plants, and the loads in normal operation, testing, credible accident and natural environment are explained. The methods of analysis for piping system are related to the transient phenomena of fluid, piping structure and local heat conduction, and linear static analysis, linear time response analysis, nonlinear time response analysis, thermal stress analysis and fluid transient phenomenon analysis are carried out. In the aseismatic design of piping system, it is desirable to avoid the vibration together with a building supporting it, and as a rule, to make it into rigid structure. The piping system is classified into high temperature and low temperature pipings. The formulas for calculating stress and the allowable condition, the points to which attention must be paid in the design of piping strength and the matters to be investigated hereafter are described. (Kako, I.)

  1. Comparison and evaluation of flexible and stiff piping systems

    International Nuclear Information System (INIS)

    Hahn, W.; Tang, H.T.; Tang, Y.K.

    1983-01-01

    An experimental and numerical study was performed on a piping system, with various support configurations, to assess the difference in piping response for flexible and stiff piping systems. Questions have arisen concerning a basic design philosophy employed in present day piping designs. One basic question is, the reliability of a flexible piping system greater than that of a stiff piping system by virtue of the fact that a flexible system has fewer snubber supports. With fewer snubbers, the pipe is less susceptible to inadvertent thermal stresses introduced by snubber malfunction during normal operation. In addition to the technical issue, the matter of cost savings in flexible piping system design is a significant one. The costs associated with construction, in-service inspection and maintenance are all significantly reduced by reducing the number of snubber supports. The evaluation study, sponsored by the Electric Power Research Institute, was performed on a boiler feedwater line at Consolidated Edison's Indian Point Unit 1. In this study, the boiler feedwater line was tested and analyzed with two fundamentally different support systems. The first system was very flexible, employing rod and spring hangers, and represented the 'old' design philosophy. The pipe system was very flexible with this support system, due to the long pipe span lengths between supports and the fact that there was only one lateral support. This support did not provide much restraint since it was near an anchor. The second system employed strut and snubber supports and represented the 'modern' design philosophy. The pipe system was relatively stiff with this support system, primarily due to the increased number of supports, including lateral supports, thereby reducing the pipe span lengths between supports. The second support system was designed with removable supports to facilitate interchange of the supports with different support types (i.e., struts, mechanical snubbers and hydraulic

  2. Radiation-resistance of polyurethane pipes for cooling liquid in BES III

    International Nuclear Information System (INIS)

    Li Xunfeng; Zheng Lifang; Ji Quan; Wu Ping; Wang Li

    2009-01-01

    Gamma ray radiation and neutron radiation are predominant in the working conditions of BES III, and the radiation-resistance aging of polyurethane pipes is very important in this condition, as the pipes of cooling liquid for beam pipe and SCQ (superconducting quadrupole) vacuum pipe in BESIII. Polyester polyurethane pipes and polyether polyurethane pipes were irradiated by gamma ray and neutron. The radiation doses were as much as ten years' doses in BES. Pressure test, FTIR and thermal analysis were used to study the radiation-resistance of these two kinds of polyurethane pipes. The results show that the radiation-resistance and thermal stability of polyester polyurethane pipes are prior to those of polyether polyurethane pipes, and the pressure resistance of polyester polyurethane pipes is almost maintained after the irradiation by gamma ray and neutron, but the polyether polyurethane pipes can be aged and ruptured after the irradiation by neutron. (authors)

  3. Modeling of residual stress mitigation in austenitic stainless steel pipe girth weldment

    International Nuclear Information System (INIS)

    Li, M.; Atteridge, D.G.; Anderson, W.E.; West, S.L.

    1994-01-01

    This study provides numerical procedures to model 40-cm-diameter, schedule 40, Type 304L stainless steel pipe girth welding and a newly proposed post-weld treatment. The treatment can be used to accomplish the goal of imparting compressive residual stresses at the inner surface of a pipe girth weldment to prevent/retard the intergranular stress corrosion cracking (IGSCC) of the piping system in nuclear reactors. This new post-weld treatment for mitigating residual stresses is cooling stress improvement (CSI). The concept of CSI is to establish and maintain a certain temperature gradient across the pipe wall thickness to change the final stress state. Thus, this process involves sub-zero low temperature cooling of the inner pipe surface of a completed girth weldment, while simultaneously keeping the outer pipe surface at a slightly elevated temperature with the help of a certain heating method. Analyses to obtain quantitative results on pipe girth welding and CSI by using a thermo-elastic-plastic finite element model are described in this paper. Results demonstrate the potential effectiveness of CSI for introducing compressive residual stresses to prevent/retard IGSCC. Because of the symmetric nature of CSI, it shows great potential for industrial application

  4. An assessment of seismic margins in nuclear plant piping

    International Nuclear Information System (INIS)

    Chen, W.P.; Jaquay, K.R.; Chokshi, N.C.; Terao, D.

    1995-01-01

    Interim results of an ongoing program to assist the U.S. Nuclear Regulatory Commission (NRC) in developing regulatory positions on the seismic analyses of piping and overall safety margins of piping systems are reported. Results of reviews of previous seismic testing, primarily the Electric Power Research Institute (EPRI)/NRC Piping and Fitting Dynamic Reliability Program, and assessments of the ASME Code, Section III, piping seismic design criteria as revised by the 1994 Addenda are reported. Major issues are identified herein only. Technical details are to be provided elsewhere. (author). 4 refs., 2 figs

  5. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M. Kothari; Gerard T. Pittard

    2005-04-01

    Utilities in the U.S. operate over 75,000 km (47,000 miles) of old cast-iron pipes for gas distribution. The bell-and-spigot joints that connect pipe sections together tend to leak as these pipes age. Current repair practices are costly and highly disruptive. The objective of this program is to design, test and commercialize a robotic system capable of sealing multiple cast-iron bell and spigot joints from a single pipe entry point. The proposed system will perform repairs while the pipe remains in service by traveling through the pipe, cleaning each joint surface, and installing a stainless-steel sleeve lined with an epoxy-impregnated felt across the joint. This approach will save considerable time and labor, avoid traffic disruption, and eliminate any requirement to interrupt service to customers (which would result in enormous expense to utilities). Technical challenges include: (1) repair sleeves must compensate for diametric variation and eccentricity of old cast-iron pipes; (2) the assembly must travel long distances through pipes containing debris; (3) the pipe wall must be effectively cleaned in the immediate area of the joint to assure good bonding of the sleeve; and (4) an innovative bolt-on entry fitting is required to conduct repair operations on live mains. The development effort is divided into eleven tasks. Task 1 (Program Management) and Task 2 (Establishment of Detailed Design Specifications) were completed previously. Task 3 (Design and Fabricate Ratcheting Stainless-Steel Repair Sleeves) has progressed to installing prototype sleeves in test cast-iron pipe segments. Efforts in the current quarter continued to be focused on Tasks 4-8. Highly valuable lessons were learned from field tests of the 4-inch gas pipe repair robot in cast-iron pipe at Public Service Electric & Gas. (These field tests were conducted and reported previously.) Several design issues were identified which need to be implemented in both the small- and large-diameter repair

  6. LHC Experimental Beam Pipe Upgrade during LS1

    CERN Document Server

    Lanza, G; Baglin, V; Chiggiato, P

    2014-01-01

    The LHC experimental beam pipes are being improved during the ongoing Long Shutdown 1 (LS1). Several vacuum chambers have been tested and validated before their installation inside the detectors. The validation tests include: leak tightness, ultimate vacuum pressure, material outgassing rate, and residual gas composition. NEG coatings are assessed by sticking probability measurement with the help of Monte Carlo simulations. In this paper the motivation for the beam pipe upgrade, the validation tests of the components and the results are presented and discussed.

  7. International piping integrity research group (IPIRG) program final report

    International Nuclear Information System (INIS)

    Schmidt, R.; Wilkowski, G.; Scott, P.; Olsen, R.; Marschall, C.; Vieth, P.; Paul, D.

    1992-04-01

    This is the final report of the International Piping Integrity Research Group (IPIRG) Programme. The IPIRG Programme was an international group programme managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United states. The objective of the programme was to develop data needed to verify engineering methods for assessing the integrity of nuclear power plant piping that contains circumferential defects. The primary focus was an experimental task that investigated the behaviour of circumferentially flawed piping and piping systems to high-rate loading typical of seismic events. To accomplish these objectives a unique pipe loop test facility was designed and constructed. The pipe system was an expansion loop with over 30 m of 406-mm diameter pipe and five long radius elbows. Five experiments on flawed piping were conducted to failure in this facility with dynamic excitation. The report: provides background information on leak-before-break and flaw evaluation procedures in piping; summarizes the technical results of the programme; gives a relatively detailed assessment of the results from the various pipe fracture experiments and complementary analyses; and, summarizes the advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG Program

  8. Experimental studies of PWR primary piping under loca

    International Nuclear Information System (INIS)

    Caumette, Pierre; Garcia, J.L.

    1980-07-01

    The experimental program performed on AQUITAINE II facility is directed to study the mechanical behavior of primary PWR pipes and the forces exerted on the neighbouring structures as a consequence of a breach opening. It has been developed in the form of a quadripartite agreement between the Commissariat a l'Energie Atomique, Framatome, Electricite de France and Westinghouse. Some forty tests have been carried out with different pipe configurations (straight tube, elbow, S- or U-shaped tube) and different break types (single or double guillotine). The following aspects are investigated: - the dynamic behavior of the pipe and in particular the formation of a plastic hinge at the restraint; - the impact function of a pipe or an energy-absorbing bumper; - the lateral stability of both ends of a pipe, after a double-guillotine break [fr

  9. Limit loads for pipe bends under combined pressure and in-plane bending based on finite element limit analysis

    International Nuclear Information System (INIS)

    Oh, Chang Sik; Kim, Yun Jae

    2006-01-01

    In the present paper, approximate plastic limit load solutions for pipe bends under combined internal pressure and bending are obtained from detailed three-dimensional (3-D) FE limit analyses based on elastic-perfectly plastic materials with the small geometry change option. The present FE results show that existing limit load solutions for pipe bends are lower bounds but can be very different from the present FE results in some cases, particularly for bending. Accordingly closed-form approximations are proposed for pipe bends under combined pressure and in-plane bending based on the present FE results. The proposed limit load solutions would be a basis of defective pipe bends and be useful to estimate non-linear fracture mechanics parameters based on the reference stress approach

  10. SEALING LARGE-DIAMETER CAST-IRON PIPE JOINTS UNDER LIVE CONDITIONS

    Energy Technology Data Exchange (ETDEWEB)

    Kiran M. Kothari; Gerard T. Pittard

    2005-01-01

    Utilities in the U.S. operate over 75,000 km (47,000 miles) of old cast-iron pipes for gas distribution. The bell-and-spigot joints that connect pipe sections together tend to leak as these pipes age. Current repair practices are costly and highly disruptive. The objective of this program is to design, test and commercialize a robotic system capable of sealing multiple cast-iron bell and spigot joints from a single pipe entry point. The proposed system will perform repairs while the pipe remains in service by traveling through the pipe, cleaning each joint surface, and installing a stainless-steel sleeve lined with an epoxy-impregnated felt across the joint. This approach will save considerable time and labor, avoid traffic disruption, and eliminate any requirement to interrupt service to customers (which would result in enormous expense to utilities). Technical challenges include: (1) repair sleeves must compensate for diametric variation and eccentricity of cast-iron pipes; (2) the assembly must travel long distances through pipes containing debris; (3) the pipe wall must be effectively cleaned in the immediate area of the joint to assure good bonding of the sleeve; and (4) an innovative bolt-on entry fitting is required to conduct repair operations on live mains. The development effort is divided into eleven tasks. Task 1 (Program Management) and Task 2 (Establishment of Detailed Design Specifications) were completed in prior quarters while Task 3 (Design and Fabricate Ratcheting Stainless-Steel Repair Sleeves) has progressed to installing prototype sleeves in cast-iron test pipe segments. Efforts in the current quarter continued to focus on Tasks 4-8. Highly valuable lessons were learned from field tests of the 4-inch gas pipe repair robot in cast-iron pipe at Public Service Electric & Gas. (These field tests were conducted and reported last quarter.) These tests identified several design issues which need to be implemented in both the small- and large

  11. Pipe-flange detection with GPR

    International Nuclear Information System (INIS)

    Bonomo, Néstor; De la Vega, Matías; Martinelli, Patricia; Osella, Ana

    2011-01-01

    This paper describes an application of the ground penetrating radar (GPR) method for detecting pipe flanges. A case history is described in which GPR was successfully used to locate pipe flanges along an 8 km metal pipeline, using a fixed-offset methodology, from the ground surface. Summaries of numerical simulations and in situ tests, performed before the definitive prospecting to evaluate the feasibility of detection, are included. Typical GPR signals are analysed and several examples shown. Constant-time sections of data volumes and migration are evaluated with the goal of distinguishing flange signals from rock signals in unclear situations. The applied methodology was effective for detecting the pipe flanges in relatively short times, with accuracies below 10 cm in the horizontal direction and 20 cm in the vertical direction

  12. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik [School of Materials Science and Engineering, Andong National University, Andong (Korea, Republic of); Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae [Power Engineering Research Institute, KEPCO Engineering and Construction Company, Seongnam (Korea, Republic of)

    2015-02-15

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  13. Real-time corrosion control system for cathodic protection of buried pipes for nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Ki Tae; Kim, Hae Woong; Kim, Young Sik; Chang, Hyun Young; Lim, Bu Taek; Park, Heung Bae

    2015-01-01

    Since the operation period of nuclear power plants has increased, the degradation of buried pipes gradually increases and recently it seems to be one of the emerging issues. Maintenance on buried pipes needs high quality of management system because outer surface of buried pipe contacts the various soils but inner surface reacts with various electrolytes of fluid. In the USA, USNRC and EPRI have tried to manage the degradation of buried pipes. However, there is little knowledge about the inspection procedure, test and manage program in the domestic nuclear power plants. This paper focuses on the development and build-up of real-time monitoring and control system of buried pipes. Pipes to be tested are tape-coated carbon steel pipe for primary component cooling water system, asphalt-coated cast iron pipe for fire protection system, and pre-stressed concrete cylinder pipe for sea water cooling system. A control system for cathodic protection was installed on each test pipe which has been monitored and controlled. For the calculation of protection range and optimization, computer simulation was performed using COMSOL Multiphysics (Altsoft co.)

  14. Interpretation of a seismic test of the IPIRG2 program

    International Nuclear Information System (INIS)

    Blay, N.; Gantenbein, F.

    1995-01-01

    In the framework of the linear and non linear analysis of PWR cracked pipes under seismic loading, the calculations of the 1.2 seismic test of the important IPIRG2 program (International Piping Integrity Research Group) was undertaken. This seismic test was performed on a pipe with a surface crack and loaded by an imposed displacement. A low level and a high level of excitation were applied to the pipe. The calculations are made with a global model including a through wall crack pipe finite element. The modal analysis made for the non-cracked pipe and the real geometrical characteristics gives a first frequency of the pipe with pressure and temperature in good agreement with the test. For the cracked pipe, the first frequency decrease is less than 0.5%. The low level response was calculated with a linear model by modal combination in order to study the importance of the both inertial and differential displacement responses in the total response. For both configurations, non-cracked and cracked, the inertial contribution to the moment at the crack location is approximately equal to 80% of the total moment. For the linear behaviour, the influence of the crack appears weak. The non linear calculations are performed with the equivalent crack previously defined up to penetration. To study the behaviour after penetration, various hypothesis for the crack size are taken. (authors). 3 refs., 6 figs., 4 tabs

  15. Crack shape developments and leak rates for circumferential complex-cracked pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brickstad, B.; Bergman, M. [SAQ Inspection Ltd., Stockholm (Sweden)

    1997-04-01

    A computerized procedure has been developed that predicts the growth of an initial circumferential surface crack through a pipe and further on to failure. The crack growth mechanism can either be fatigue or stress corrosion. Consideration is taken to complex crack shapes and for the through-wall cracks, crack opening areas and leak rates are also calculated. The procedure is based on a large number of three-dimensional finite element calculations of cracked pipes. The results from these calculations are stored in a database from which the PC-program, denoted LBBPIPE, reads all necessary information. In this paper, a sensitivity analysis is presented for cracked pipes subjected to both stress corrosion and vibration fatigue.

  16. Heat pipe thermal control of slender optics probes

    International Nuclear Information System (INIS)

    Prenger, F.C.

    1979-01-01

    The thermal design for a stereographic viewing system is presented. The design incorporates an annular heat pipe and thermal isolation techniques. Test results are compared with design predictions for a prototype configuration. Test data obtained during heat pipe startup showing temperature gradients along the evaporator wall are presented. Correlations relating maximum wall temperature differences to a liquid Reynolds number were obtained at low power levels. These results are compared with Nusselt's Falling Film theory

  17. An inspection of pipe by snake robot

    Directory of Open Access Journals (Sweden)

    František Trebuňa

    2016-10-01

    Full Text Available The article deals with development and application of snake robot for inspection pipes. The first step involves the introduction of a design of mechanical and electrical parts of the snake robot. Next, the analysis of the robot locomotion is introduced. For the curved pipe, potential field method is used. By this method, the system is able to generate path for the head and rear robot, linking the environment with obstacles, which are represented by the walls of the pipe. Subsequently, the solution of potential field method is used in inverse kinematic model, which respects tasks as obstacle avoidance, joint limit avoidance, and singularity avoidance. Mentioned approach is then tested on snake robot in provisional pipe with rectangular cross section. For this research, software Matlab (2013b is used as the control system in cooperation with the control system of robot, which is based on microcontrollers. By experiments, it is shown that designed robot is able to pass through straight and also curved pipe.

  18. Test on large-scale seismic isolation elements

    International Nuclear Information System (INIS)

    Mazda, T.; Shiojiri, H.; Oka, Y.; Fujita, T.; Seki, M.

    1989-01-01

    Demonstration test of seismic isolation elements is considered as one of the most important items in the application of seismic isolation system to fast breeder reactor (FBR) plant. Facilities for testing seismic isolation elements have been built. This paper reports on tests for fullscale laminated rubber bearing and reduced scale models are conducted. From the result of the tests, the laminated rubber bearings turn out to satisfy the specification. Their basic characteristics are confirmed from the tests with fullscale and reduced scale models. The ultimate capacity of the bearings under the condition of ordinary temperature are evaluated

  19. OTEC Cold Water Pipe-Platform Subsystem Dynamic Interaction Validation

    Energy Technology Data Exchange (ETDEWEB)

    Varley, Robert [Lockheed Martin Corporation, Manassas, VA (United States); Halkyard, John [John Halkyard and Associates, Houston, TX (United States); Johnson, Peter [BMT Scientific Marine Services, Inc., Houston, TX (United States); Shi, Shan [Houston Offshore Engineering, Houston, TX (United States); Marinho, Thiago [Federal Univ. of Rio de Janeiro (Brazil). LabOceano

    2014-05-09

    A commercial floating 100-megawatt (MW) ocean thermal energy conversion (OTEC) power plant will require a cold water pipe (CWP) with a diameter of 10-meter (m) and length of up to 1,000 m. The mass of the cold water pipe, including entrained water, can exceed the mass of the platform supporting it. The offshore industry uses software-modeling tools to develop platform and riser (pipe) designs to survive the offshore environment. These tools are typically validated by scale model tests in facilities able to replicate real at-sea meteorological and ocean (metocean) conditions to provide the understanding and confidence to proceed to final design and full-scale fabrication. However, today’s offshore platforms (similar to and usually larger than those needed for OTEC applications) incorporate risers (or pipes) with diameters well under one meter. Secondly, the preferred construction method for large diameter OTEC CWPs is the use of composite materials, primarily a form of fiber-reinforced plastic (FRP). The use of these material results in relatively low pipe stiffness and large strains compared to steel construction. These factors suggest the need for further validation of offshore industry software tools. The purpose of this project was to validate the ability to model numerically the dynamic interaction between a large cold water-filled fiberglass pipe and a floating OTEC platform excited by metocean weather conditions using measurements from a scale model tested in an ocean basin test facility.

  20. Branch-pipe-routing approach for ships using improved genetic algorithm

    Science.gov (United States)

    Sui, Haiteng; Niu, Wentie

    2016-09-01

    Branch-pipe routing plays fundamental and critical roles in ship-pipe design. The branch-pipe-routing problem is a complex combinatorial optimization problem and is thus difficult to solve when depending only on human experts. A modified genetic-algorithm-based approach is proposed in this paper to solve this problem. The simplified layout space is first divided into threedimensional (3D) grids to build its mathematical model. Branch pipes in layout space are regarded as a combination of several two-point pipes, and the pipe route between two connection points is generated using an improved maze algorithm. The coding of branch pipes is then defined, and the genetic operators are devised, especially the complete crossover strategy that greatly accelerates the convergence speed. Finally, simulation tests demonstrate the performance of proposed method.

  1. Space reactor fuel element testing in upgraded TREAT

    International Nuclear Information System (INIS)

    Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

  2. Space reactor fuel element testing in upgraded TREAT

    Science.gov (United States)

    Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

    1993-01-01

    The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

  3. Guide to good practices for equipment and piping labeling

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-12-01

    This Guide to Good Practices is written to enhance understanding of, and provide direction for, Equipment and Piping Labeling, Chapter XVIII of Department of Energy (DOE) Order 5480.19, Conduct of Operations Requirements for DOE Facilities. The practices in this guide should be considered when planning or reviewing labeling programs. Contractors are advised to adopt procedures that meet the intent of DOE Order 5480.19. Equipment and Piping Labeling is an element of an effective Conduct of Operations program. The complexity and array of activities performed in DOE facilities dictate the necessity for a coordinated labeling program to promote safe and efficient operations.

  4. Fluid structure interaction in piping systems

    Energy Technology Data Exchange (ETDEWEB)

    Svingen, Bjoernar

    1996-12-31

    The Dr. ing. thesis relates to an analysis of fluid structure interaction in piping systems in the frequency domain. The governing equations are the water hammer equations for the liquid, and the beam-equations for the structure. The fluid and structural equations are coupled through axial stresses and fluid continuity relations controlled by the contraction factor (Poisson coupling), and continuity and force relations at the boundaries (junction coupling). A computer program has been developed using the finite element method as a discretization technique both for the fluid and for the structure. This is made for permitting analyses of large systems including branches and loops, as well as including hydraulic piping components, and experiments are executed. Excitations are made in a frequency range from zero Hz and up to at least one thousand Hz. Frequency dependent friction is modelled as stiffness proportional Rayleigh damping both for the fluid and for the structure. With respect to the water hammer equations, stiffness proportional damping is seen as an artificial (bulk) viscosity term. A physical interpretation of this term in relation to transient/oscillating hydraulic pipe-friction is given. 77 refs., 72 figs., 4 tabs.

  5. PPOOLEX experiments with a modified blowdown pipe outlet

    International Nuclear Information System (INIS)

    Laine, J.; Puustinen, M.; Raesaenen, A.

    2009-08-01

    This report summarizes the results of the experiments with a modified blowdown pipe outlet carried out with the PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through a vertical DN200 blowdown pipe to the condensation pool. Four reference experiments with a straight pipe and ten with the Forsmark type collar were carried out. The main purpose of the experiment series was to study the effect of a blowdown pipe outlet collar design on loads caused by chugging phenomena (rapid condensation) while steam is discharged into the condensation pool. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. During the experiments the initial temperature level of the condensation pool water was either 20-25 or 50-55 deg. C. The steam flow rate varied from 400 to 1200 g/s and the temperature of incoming steam from 142 to 185 deg. C. In the experiments with 20-25 deg. C pool water, even 10 times higher pressure pulses were measured inside the blowdown pipe in the case of the straight pipe than with the collar. In this respect, the collar design worked as planned and removed the high pressure spikes from the blowdown pipe. Meanwhile, there seemed to be no suppressing effect on the loads due to the collar in the pool side in this temperature range. Registered loads in the pool were approximately in the same range (or even a little higher) with the collar as with the straight pipe. In the experiments with 50-55 deg. C pool water no high pressure pulses were measured inside the blowdown pipe either with the straight pipe or with the collar. In this case, more of the suppressing effect is probably due to the warmer pool water than due to the modified pipe outlet. It has been observed already in the earlier experiments with a straight pipe in the POOLEX and PPOOLEX facilities that warm pool water has a diminishing effect on

  6. PPOOLEX experiments with a modified blowdown pipe outlet

    Energy Technology Data Exchange (ETDEWEB)

    Laine, J.; Puustinen, M.; Raesaenen, A. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

    2009-08-15

    This report summarizes the results of the experiments with a modified blowdown pipe outlet carried out with the PPOOLEX test facility designed and constructed at Lappeenranta University of Technology. Steam was blown into the dry well compartment and from there through a vertical DN200 blowdown pipe to the condensation pool. Four reference experiments with a straight pipe and ten with the Forsmark type collar were carried out. The main purpose of the experiment series was to study the effect of a blowdown pipe outlet collar design on loads caused by chugging phenomena (rapid condensation) while steam is discharged into the condensation pool. The PPOOLEX test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. During the experiments the initial temperature level of the condensation pool water was either 20-25 or 50-55 deg. C. The steam flow rate varied from 400 to 1200 g/s and the temperature of incoming steam from 142 to 185 deg. C. In the experiments with 20-25 deg. C pool water, even 10 times higher pressure pulses were measured inside the blowdown pipe in the case of the straight pipe than with the collar. In this respect, the collar design worked as planned and removed the high pressure spikes from the blowdown pipe. Meanwhile, there seemed to be no suppressing effect on the loads due to the collar in the pool side in this temperature range. Registered loads in the pool were approximately in the same range (or even a little higher) with the collar as with the straight pipe. In the experiments with 50-55 deg. C pool water no high pressure pulses were measured inside the blowdown pipe either with the straight pipe or with the collar. In this case, more of the suppressing effect is probably due to the warmer pool water than due to the modified pipe outlet. It has been observed already in the earlier experiments with a straight pipe in the POOLEX and PPOOLEX facilities that warm pool water has a diminishing effect on

  7. Avoiding steam-bubble-collapse-induced water hammers in piping systems

    International Nuclear Information System (INIS)

    Chou, Y.; Griffith, P.

    1989-10-01

    In terms of the frequency of occurrence, steam bubble collapse in subcooled water is the dominant initiating mechanism for water hammer events in nuclear power plants. Water hammer due to steam bubble collapse occurs when water slug forms in stratified horizontal flow, or when steam bubble is trapped at the end of the pipe. These types of water hammer events have been studied experimentally and analytically in order to develop stability maps showing those combinations of filling velocities and liquid subcooling that cause water hammer and those which don't. In developing the stability maps, experiments with different piping orientations were performed in a low pressure laboratory apparatus. Details of these experiments are described, including piping arrangement, test procedures, and test results. Visual tests using a transparent Lexan pipe are also performed to study the flow regimes accompanying the water hammer events. All analytical models were tested by comparison with the corresponding experimental results. Based on these models, and step-by-step approach for each flow geometry is presented for plant designers and engineers to follow in avoiding water hammer induced by steam bubble collapse when admitting cold water into pipes filled with steam. 37 refs., 54 figs., 2 tabs

  8. Criteria for accepting piping thermal expansion movements during FFTF plant startup

    International Nuclear Information System (INIS)

    Clark, G.L.; Anderson, M.J.

    1981-03-01

    A deflection measurement program was conducted as a final step in the design qualification of the Fast Flux Test Facility liquid sodium piping. Measurements were obtained from the ambient empty position, through the 400 0 F (204 0 C) sodium fill, to an 800 0 F (427 0 C) maximum iso-thermal test condition. The program was designed to confirm that the pipe responded as predicted under both deadweight and thermal expansion loads. This paper describes the design of the test programs; the criteria used to select appropriate measurement locations from the approximately 4000 supports used on this pipe; and the criteria used to accept test results

  9. A study of inter linkage effects on Candu feeder piping

    International Nuclear Information System (INIS)

    Li, M.; Aggarwal, M.L.; Meysner, A.

    2005-01-01

    A CANDU (Canadian Deuterium Uranium) reactor core consists of a large number of fuel channels where heat is generated. Two feeder pipes are connected to each fuel channel to transport D 2 O coolant into and out of the reactor core. The feeder piping is designed to the requirements of Class 1 piping of Section III NB of the ASME Boiler and Pressure Vessel and CSA Codes. Feeder piping stress analysis is being performed to demonstrate the code compliance check and the fitness for service of feeders. In the past, stress analyses were conducted for each individual feeder without including interaction effects among connected feeders. Interaction effects occur as a result of linkages that exist between feeders to prevent fretting and impacting damage during normal, abnormal and accident conditions. In this paper, a 'combined' approach is adopted to include all feeders connected by inter linkages into one feeder piping model. MSC/NASTRAN finite element software was used in the stress simulation, which contains up to 127 feeder pipes. The ASME Class 1 piping analysis was conducted to investigate the effects of the linkages between feeders. Both seismic time history and broadened response spectra methods were used in the seismic stress calculation. The results show that the effect of linkages is significant in dynamic stresses for all feeder configurations, as well as in static stresses for certain feeder configurations. The single feeder analysis could either underestimate or overestimate feeder stresses depending on the pipe geometry and bend wall thickness. (authors)

  10. Development of integrated insulation joint for cooling pipe in tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Abe, Tetsuya; Kawamura, Masashi; Yamazaki, Seiichiro.

    1994-08-01

    In a tokamak fusion reactor, an electrically insulated part is needed for an in-vessel piping system in order to break an electric circuit loop. When a closed loop is formed in the piping system, large induced electromagnetic forces during a plasma disruption (rapid plasma current quench) could give damages on the piping system. Ceramic brazing joint is a conventional method for the electric circuit break, but an application to the fusion reactor is not feasible due to its brittleness. Here, a stainless steel/ceramics/stainless steel functionally gradient material (FGM) has been proposed and developed as an integrated insulation joint of the piping system. Both sides of the joint can be welded to the main pipes, and expected to be reliable even in the fusion reactor environment. When the FGM joint is manufactured by way of a sintering process, a residual thermal stress is the key issue. Through detailed computations of the residual thermal stress and several trial productions, tubular elements of FGM joints have been successfully manufactured. (author)

  11. Prediction of three-dimensional residual stresses at localised indentations in pipes

    International Nuclear Information System (INIS)

    Hyde, T.H.; Luo, R.; Becker, A.A.

    2012-01-01

    Residual stresses are investigated using Finite Element (FE) analyses at localised indentations in pipes with and without internal pressures due to reverse plasticity caused by springback of the surrounding material after removal of the indenter. The indentation loading is applied via rigid 3D short indenters. The effects of the residual indentation depth, internal pressure, indenter size and different material properties on the residual stresses for different pipes have been investigated by carrying out parametric sensitivity studies. In order to predict the residual stresses, empirical formulations have been developed, which show a good correlation with the FE for residual stresses for pipes with diameter to thickness ratios of 35–72. - Highlights: ► A comprehensive elastic–plastic FE analysis of residual stresses caused by localised pipe indentations is presented. ► The effects of residual indentation depth, internal pressure, indenter size and material properties have been studied. ► Empirical formulations have been developed, which show a good correlation with the FE for residual stresses for pipes with diameter to thickness ratios of 35–72.

  12. Analytical model of impedance in elliptical beam pipes

    CERN Document Server

    Pesah, Arthur Chalom

    2017-01-01

    Beam instabilities are among the main limitations in building higher intensity accelerators. Having a good impedance model for every accelerators is necessary in order to build components that minimize the probability of instabilities caused by the interaction beam-environment and to understand what piece to change in case of intensity increasing. Most of accelerator components have their impedance simulated with finite elements method (using softwares like CST Studio), but simple components such as circular or flat pipes are modeled analytically, with a decreasing computation time and an increasing precision compared to their simulated model. Elliptical beam pipes, while being a simple component present in some accelerators, still misses a good analytical model working for the hole range of velocities and frequencies. In this report, we present a general framework to study the impedance of elliptical pipes analytically. We developed a model for both longitudinal and transverse impedance, first in the case of...

  13. Study on finned pipe performance as a ground heat exchanger

    Science.gov (United States)

    Lin, Qinglong; Ma, Jinghui; Shi, Lei

    2017-08-01

    The GHEs (ground heat exchangers) is an important element that determines the thermal efficiency of the entire ground-source heat-pump system. The aim of the present study is to clarify thermal performance of a new type GHE pipe, which consists straight fins of uniform cross sectional area. In this paper, GHE model is introduced and an analytical model of new type GHE pipe is developed. The heat exchange rate of BHEs utilizing finned pips is 40.42 W/m, which is 16.3% higher than normal BHEs, based on simulation analyses.

  14. Experiments and calculations to leak openings and leak rates on typical piping components and systems

    International Nuclear Information System (INIS)

    Hoefler, A.; Grebner, H.

    1992-01-01

    Calculations of leak opening and leak rate for through cracks in piping components have been performed. The analyses are pre- or mostly post-calculations to experiments performed at the HDR facility under PWR operating conditions. Piping components under consideration are small diameter straight pipes with circumferential cracks, pipe bends with longitudinal or circumferential cracks and pipe branches with weldment cracks. The component are loaded by internal pressure and opening as well as closing bending moment. The finite element method and two-phase flow leak rate programs are used for the calculations. Results of the analyses are presented as J-integral values, crack opening displacements and areas and leak rates as well as comparisons to the experimental results. 6 refs., 16 figs., 2 tabs

  15. Review of application code and standards for mechanical and piping design of HANARO fuel test loop

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. Y.

    1998-02-01

    The design and installation of the irradiation test facility for verification test of the fuel performance are very important in connection with maximization of the utilization of HANARO. HANARO fuel test loop was designed in accordance with the same code and standards of nuclear power plant because HANARO FTL will be operated the high pressure and temperature same as nuclear power plant operation conditions. The objective of this study is to confirm the propriety of application code and standards for mechanical and piping of HANARO fuel test loop and to decide the technical specification of FTL systems. (author). 18 refs., 8 tabs., 6 figs.

  16. Flow patterns and hydraulic losses in quasi-coil pipes : The effects of configuration of bend cross section, curvature ratio and bend angle

    OpenAIRE

    Shimizu, Yukimaru; Sugino, Koichi; Yasui, Masaji; Hayakawa, Yukitaka; Kuzuhara, Sadao

    1985-01-01

    Pipes with bend combinations are much used in the heat exchangers, since the curved path in the bends promotes the mixing in flow for active heat transfer. In the present paper, one of the pipes with bend combinations, namely, quasi-coiled pipes composed of many bend elements are investigated, and the relationships between the hydraulic loss and the secondary flow are studied experimentally. The configurations of the cross sections, the bent angles and the curvature ratios of the bend element...

  17. Thermal performance of solar air collection-storage system with phase change material based on flat micro-heat pipe arrays