Project W-320, 241-C-106 sluicing: Piping calculations. Volume 4
International Nuclear Information System (INIS)
Bailey, J.W.
1998-01-01
This supporting document has been prepared to make the FDNW calculations for Project W-320 readily retrievable. The objective of this calculation is to perform the structural analysis of the Pipe Supports designed for Slurry and Supernate transfer pipe lines in order to meet the requirements of applicable ASME codes. The pipe support design loads are obtained from the piping stress calculations W320-27-I-4 and W320-27-I-5. These loads are the total summation of the gravity, pressure, thermal and seismic loads. Since standard typical designs are used for each type of pipe support such as Y-Stop, Guide and Anchors, each type of support is evaluated for the maximum loads to which this type of supports are subjected. These loads are obtained from the AutoPipe analysis and used to check the structural adequacy of these supports
Project W-320, 241-C-106 sluicing: Piping calculations. Volume 4
Energy Technology Data Exchange (ETDEWEB)
Bailey, J.W.
1998-07-24
This supporting document has been prepared to make the FDNW calculations for Project W-320 readily retrievable. The objective of this calculation is to perform the structural analysis of the Pipe Supports designed for Slurry and Supernate transfer pipe lines in order to meet the requirements of applicable ASME codes. The pipe support design loads are obtained from the piping stress calculations W320-27-I-4 and W320-27-I-5. These loads are the total summation of the gravity, pressure, thermal and seismic loads. Since standard typical designs are used for each type of pipe support such as Y-Stop, Guide and Anchors, each type of support is evaluated for the maximum loads to which this type of supports are subjected. These loads are obtained from the AutoPipe analysis and used to check the structural adequacy of these supports.
Calculation of dynamic hydraulic forces in nuclear plant piping systems
International Nuclear Information System (INIS)
Choi, D.K.
1982-01-01
A computer code was developed as one of the tools needed for analysis of piping dynamic loading on nuclear power plant high energy piping systems, including reactor safety and relief value upstream and discharge piping systems. The code calculates the transient hydraulic data and dynamic forces within the one-dimensional system, caused by a pipe rupture or sudden value motion, using a fixed space and varying time grid-method of characteristics. Subcooled, superheated, homogeneous two-phase and transition flow regimes are considered. A non-equilibrium effect is also considered in computing the fluid specific volume and fluid local sonic velocity in the two-phase mixture. Various hydraulic components such as a spring loaded or power operated value, enlarger, orifice, pressurized tank, multiple pipe junction (tee), etc. are considered as boundary conditions. Comparisons of calculated results with available experimental data shows a good agreement. (Author)
International Nuclear Information System (INIS)
Gomez T, A. M.; Xolocostli M, V.; Lopez M, R.; Filio L, C.; Royl, P.
2014-10-01
In 2012 was modeled of primary and secondary container of the nuclear power plant of Laguna Verde (NPP-L V) for the CFD Gas-Flow code. These models were used to calculate hydrogen volume concentrations run release the reactor building in case of a severe accident. The results showed that the venting would produce detonation conditions in the venting level (level 33) and flammability at ground level of reload. One of the solutions to avoid reaching critical concentrations (flammable or detonable) inside the reactor building and thus safeguard the contentions is to make a rigid venting. The rigid vent is a pipe connected to the primary container could go to the level 33 of the secondary container and style fireplace climb to the top of the reactor building. The analysis of hydrogen transport inside the vent pipe can be influenced by various environmental criteria and factors vent, so a logical consequence of the 2012 analysis is the analysis of the gases transport within said pipe to define vent ideal conditions. For these evaluations the vent pipe was modeled with a fine mesh of 32 radial interior nodes and a coarse mesh of 4 radial interior nodes. With three-dimensional models were realized calculations that allow observing the influence of heat transfer in the long term, i.e. a complete analysis of exhaust (approx. 700 seconds). However, the most interesting results focus on the first milliseconds, when the H 2 coming from the atmosphere of the primary container faces the air in the vent pipe. These first milliseconds besides allowing evaluating the detonation criteria in great detail in the different tubular sections similarly allow evaluating the pressure wave that occurs in the pipe and that at some point slows to the fluid on the last tubular section and could produce a detonation inside the pipe. Results are presented for venting fixed conditions, showing possible detonations into the pipe. (Author)
Microcomputer generated pipe support calculations
International Nuclear Information System (INIS)
Hankinson, R.F.; Czarnowski, P.; Roemer, R.E.
1991-01-01
The cost and complexity of pipe support design has been a continuing challenge to the construction and modification of commercial nuclear facilities. Typically, pipe support design or qualification projects have required large numbers of engineers centrally located with access to mainframe computer facilities. Much engineering time has been spent repetitively performing a sequence of tasks to address complex design criteria and consolidating the results of calculations into documentation packages in accordance with strict quality requirements. The continuing challenges of cost and quality, the need for support engineering services at operating plant sites, and the substantial recent advances in microcomputer systems suggested that a stand-alone microcomputer pipe support calculation generator was feasible and had become a necessity for providing cost-effective and high quality pipe support engineering services to the industry. This paper outlines the preparation for, and the development of, an integrated pipe support design/evaluation software system which maintains all computer programs in the same environment, minimizes manual performance of standard or repetitive tasks, and generates a high quality calculation which is consistent and easily followed
Research program plan: piping. Volume 3
International Nuclear Information System (INIS)
Vagins, M.; Strosnider, J.
1985-07-01
Regulatory issues related to piping can be divided into the three areas of pipe cracking, postulated design basis pipe breaks, and design of piping for seismic and other dynamic loads. The first two of these issues are in the domain of the Materials Engineering Branch (MEBR), while the last of the three issues is the responsibility of the Mechanical/Structural Engineering Branch. This volume of the MEBR Research Plan defines the critical aspects of the pipe cracking and postulated design basis pipe break issues and identifies those research efforts and results necessary for their resolution. In general, the objectives of the MERB Piping Research Program are to provide experimentally validated analytic techniques and appropriate material properties characterization methods and data to support regulatory activities related to evaluating and ensuring piping integrity
Calculational study on reactivity effect of pipe intersections
International Nuclear Information System (INIS)
Okuno, Hiroshi; Naito, Yoshitaka; Kaneko, Toshiyuki.
1995-03-01
A simple formulation was proposed for evaluating the increment of reactivity due to the attachment of pipes to a vessel filled with fuel solution, and its validity was checked by numerical calculations. The formulation was based on the neutron balance equation which had been applied to the criticality safety analysis code MUTUAL for multi-unit systems, and the current formulation considered further the deviation of the representative neutron source point from the center of each pipe. The formulation was validated for models of 2- and 3-dimensional fuel systems by comparison with the precise calculations using the Monte Carlo code KENO-IV. For systems of pipes attached perpendicularly to the side of a cylindrical vessel, the size and number of negligible pipes were shown that corresponded to a very small increment (e.g. 0.3% Δk/k) of the neutron multiplication factor. (author)
Calculation of forces on reactor containment fan cooler piping
International Nuclear Information System (INIS)
Miller, J.S.; Ramsden, K.
2004-01-01
The purpose of this paper is to present the results of the Reactor Containment Fan Cooler (RCFC) system piping load calculations. These calculations are based on piping loads calculated using the EPRI methodology and RELAP5 to simulate the hydraulic behavior of the system. The RELAP5 generated loads were compared to loads calculated using the EPRI GL-96-06 methodology. This evaluation was based on a pressurized water reactor's RCFC coils thermal hydraulic behavior during a Loss of Offsite Power (LOOP) and a loss of coolant accident (LOCA). The RCFC consist of two banks of service water and chill water coils. There are 5 SX and 5 chill water coils per bank. Therefore, there are 4 RCFC units in the containment with 2 banks of coils per RCFC. Two Service water pumps provide coolant for the 4 RCFC units (8 banks total, 2 banks per RCFC unit and 2 RCFC units per pump). Following a LOOP/LOCA condition, the RCFC fans would coast down and upon being re-energized, would shift to low-speed operation. The fan coast down is anticipated to occur very rapidly due to the closure of the exhaust damper as a result of LOCA pressurization effects. The service water flow would also coast down and be restarted in approximately 43 seconds after the initiation of the event. The service water would drain from the RCFC coils during the pump shutdown and once the pumps restart, water is quickly forced into the RCFC coils causing hydraulic loading on the piping. Because of this scenario and the potential for over stressing the piping, an evaluation was performed by the utility using RELAP5 to assess the piping loads. Subsequent to the hydraulic loads being analyzed using RELAP5, EPRI through GL-96-06 provided another methodology to assess loads on the RCFC piping system. This paper presents the results of using the EPRI methodology and RELAP5 to perform thermal hydraulic load calculations. It is shown that both EPRI methodology and RELAP5 calculations can be used to generate hydraulic loads
Parametric calculations of fatigue-crack growth in piping
International Nuclear Information System (INIS)
Simonen, F.A.; Goodrich, C.W.
1983-06-01
This study presents calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and number of stress cycles, the piping material, and environmental factors. The results indicate that the present flaw acceptance standards of ASME Section XI provide a relatively consistent set of allowable flaw sizes because the predicted life of flawed piping is relatively insensitive to pipe wall thickness, flaw aspect ratio, and piping material (ferritic versus austenitic). On the other hand, the results show that flaws that are acceptable under ASME Section XI can grow at unacceptable rates if the cyclic stresses are at the maximum level permitted by the design rules of ASME Section III. However, a review of the conservatisms inherent to the ASME code rules is presented to explain the low occurrence of piping fatigue failures in service. It is concluded that decreases in the allowable flaw sizes are not justified
International Nuclear Information System (INIS)
1985-04-01
This document summarizes a comprehensive review of NRC requirements for Nuclear Piping by the US NRC Piping Review Committee. Four topical areas, addressed in greater detail in Volumes 1 through 4 of this report, are included: (1) Stress Corrosion Cracking in Piping of Boiling Water Reactor Plants; (2) Evaluation of Seismic Design; (3) Evaluation of Potential for Pipe Breaks; and (4) Evaluation of Other Dynamic Loads and Load Combinations. This volume summarizes the major issues, reviews the interfaces, and presents the Committee's conclusions and recommendations for updating NRC requirements on these issues. This report also suggests research or other work that may be required to respond to issues not amenable to resolution at this time
Calculation of piping loads due to filling procedures
International Nuclear Information System (INIS)
Swidersky, Harald; Thiele, Thomas
2012-01-01
Filling procedures in piping systems are usually not load cases that are studied by fluid dynamic and structure dynamic analyses with respect to the integrity of pipes and supports. Although, their frequency is higher than that of postulated accidental transients, therefore they have to be considered for fatigue analyses. The piping and support loads due to filling procedures are caused by the density differences if the transported fluids, for instance in flows with the transport of gas bubbles. The impact duration of the momentum forces is defined by the flow velocity and the length of discontinuities in the piping segments. Filling procedures end very often with a shock pressure, caused by the impact and decelerating of the fluid front at smaller cross sections. The suitability of the thermally hydraulics program RELAP/MOD3.3 for the calculation of realistic loads from filling procedures was studied, the results compared with experimental data. It is shown that dependent on the discretization level the loads are partial significantly underestimated.
Calculation of loading on pipes during filling processes
International Nuclear Information System (INIS)
Thiele, Thomas; Swidersky, Harald
2013-01-01
Filling processes in pipe systems do normally not belong to load design cases for which the integrity of pipelines and their mountings are verified with fluid- and structure-dynamic analysis. However, their frequency of occurrence is several times higher than those of the postulated incident-induced transients. That is why they have to be taken into consideration within fatigue analysis. The loading on pipes or rather on their mountings during filling processes originates from differences in the density of the transported fluids, e.g. at transport of gas slugs within water flow. The exposure time of the flow momentum force is fixed by the height of the flow velocity and by the length of discontinuities in the pipeline sections. Filling procedures frequently end with a pressure surge which was caused by the impingement and decelaration of the water plug at orifices in pipe systems. The calculation of such processes with 1D fluid-dynamic or rather thermal-hydraulic programs requires an idealization of the real form of the two phase flow or respectively of the two phase interface. In the past, several two phase flow regime maps were developed and implemented in codes for this. In this paper, the applicability of the thermo-hydraulic program RELAP5/MOD3.3 which is established in nuclear engineering is examined in order to calculate realistic loads from plug flows during the filling processes. For this, post-test calculations of experiments have been performed and the results have been compared with the experimental results as well as with the classical analytical approach according to Joukowsky. The comparison shows that, dependent on the discretization, the calculated loads are indeed partly underestimated, though the calculation results according to the Joukowsky-approach lie above the measurements. (orig.)
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
Aksenov, Andrey; Malysheva, Anna
2018-03-01
An exact calculation of the heat exchange of evaporative surfaces is possible only if the physical processes of hydrodynamics of two-phase flows are considered in detail. Especially this task is relevant for the design of refrigeration supply systems for high-rise buildings, where powerful refrigeration equipment and branched networks of refrigerants are used. On the basis of experimental studies and developed mathematical model of asymmetric dispersed-annular flow of steam-water flow in horizontal steam-generating pipes, a calculation formula has been obtained for determining the boundaries of the zone of improved heat transfer and the critical value of the heat flux density. A new theoretical approach to the solution of the problem of the flow structure of a two-phase flow is proposed. The applied method of dissipative characteristics of a two-phase flow in pipes and the principle of a minimum rate of entropy increase in stabilized flows made it possible to obtain formulas that directly reflect the influence of the viscous characteristics of the gas and liquid media on their distribution in the flow. The study showed a significant effect of gravitational forces on the nature of the phase distribution in the cross section of the evaporative tubes. At a mass velocity of a two-phase flow less than 700 kg / m2s, the volume content of the liquid phase near the upper outer generating lines of the tube is almost an order of magnitude lower than the lower one. The calculation of the heat transfer crisis in horizontal evaporative tubes is obtained. The calculated dependence is in good agreement with the experimental data of the author and a number of foreign researchers. The formula generalizes the experimental data for pipes with the diameter of 6-40 mm in the pressure of 2-7 MPa.
Energy Technology Data Exchange (ETDEWEB)
Gomez T, A. M.; Xolocostli M, V. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Lopez M, R.; Filio L, C. [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico); Royl, P., E-mail: armando.gomez@inin.gob.mx [Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz I, D-76344 Eggenstein-Leopoldshafen (Germany)
2014-10-15
In 2012 was modeled of primary and secondary container of the nuclear power plant of Laguna Verde (NPP-L V) for the CFD Gas-Flow code. These models were used to calculate hydrogen volume concentrations run release the reactor building in case of a severe accident. The results showed that the venting would produce detonation conditions in the venting level (level 33) and flammability at ground level of reload. One of the solutions to avoid reaching critical concentrations (flammable or detonable) inside the reactor building and thus safeguard the contentions is to make a rigid venting. The rigid vent is a pipe connected to the primary container could go to the level 33 of the secondary container and style fireplace climb to the top of the reactor building. The analysis of hydrogen transport inside the vent pipe can be influenced by various environmental criteria and factors vent, so a logical consequence of the 2012 analysis is the analysis of the gases transport within said pipe to define vent ideal conditions. For these evaluations the vent pipe was modeled with a fine mesh of 32 radial interior nodes and a coarse mesh of 4 radial interior nodes. With three-dimensional models were realized calculations that allow observing the influence of heat transfer in the long term, i.e. a complete analysis of exhaust (approx. 700 seconds). However, the most interesting results focus on the first milliseconds, when the H{sub 2} coming from the atmosphere of the primary container faces the air in the vent pipe. These first milliseconds besides allowing evaluating the detonation criteria in great detail in the different tubular sections similarly allow evaluating the pressure wave that occurs in the pipe and that at some point slows to the fluid on the last tubular section and could produce a detonation inside the pipe. Results are presented for venting fixed conditions, showing possible detonations into the pipe. (Author)
Piping and pipeline calculations manual construction, design fabrication and examination
Ellenberger, Philip
2010-01-01
The lack of commentary, or historical perspective, regarding the codes and standards requirements for piping design and construction is an obstacle to the designer, manufacturer, fabricator, supplier, erector, examiner, inspector, and owner who want to provide a safe and economical piping system. An intensive manual, this book will utilize hundreds of calculation and examples based on of 40 years of personal experiences of the author as both an engineer and instructor. Each example demonstrates how the code and standard has been correctly and incorrectly applied. This book is a ?no non
W-320 waste retrieval sluicing system transfer line flushing volume and frequency calculation
International Nuclear Information System (INIS)
Bailey, J.W.
1997-01-01
The calculations contained in this analysis document establish the technical basis for the volume, frequency, and flushing fluid to be utilized for routine Waste Retrieval Sluicing System (WRSS) process line flushes. The WRSS was installed by Project W-320, Tank 241-C-106 Sluicing. The double contained pipelines being flushed have 4 inch stainless steel primary pipes. The flushes are intended to prevent hydrogen buildup in the transfer lines and to provide ALARA conditions for maintenance personnel
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
A DYNA3D calculation for impact on a pipe target
International Nuclear Information System (INIS)
Neilson, A.J.
1983-11-01
This report describes experimental studies to examine the response of pipework, typical of that used in nuclear power plants, to the impact of missiles representing fragments of disintegrating machinery. The finite element code DYNA3D has been used to make a calculation for one experiment in which an instrumented target pipe was impacted by a cylindrical steel billet. Transient displacement of the missile and target as well as permanent deformations of the target pipe were well-predicted by the code. The code reproduced the main features of the experimental transient strain measurements with the timings of the various straining phases being calculated very closely. Detailed quantitative comparisons cannot be made because of the lack of appropriate facilities in the GRAPE post-processing code. (U.K.)
Numerical calculation of flashing from long pipes using a two-field model
International Nuclear Information System (INIS)
Rivard, W.C.; Torrey, M.D.
1976-05-01
A two-field model for two-phase flows, in which the vapor and liquid phases have different densities, velocities, and temperatures, has been used to calculate the flashing of water from long pipes. The IMF (Implicit Multifield) technique is used to numerically solve the transient equations that govern the dynamics of each phase. The flow physics is described with finite rate phase transitions, interfacial friction, heat transfer, pipe wall friction, and appropriate state equations. The results of the calculations are compared with measured histories of pressure, temperature, and void fraction. A parameter study indicates the relative sensitivity of the results to the various physical models that are used
Numerical calculation of flashing from long pipes using a two-field model
International Nuclear Information System (INIS)
Rivard, W.C.; Torrey, M.D.
1975-11-01
A two-field model for two-phase flows, in which the vapor and liquid phases have different densities, velocities, and temperatures, has been used to calculate the flashing of water from long pipes. The IMF (Implicit Multifield) technique is used to numerically solve the transient equations that govern the dynamics of each phase. The flow physics is described with finite rate phase transitions, interfacial friction, heat transfer, pipe wall friction, and appropriate state equations. The results of the calculations are compared with measured histories of pressure, temperature, and void fraction. A parameter study indicates the relative sensitivity of the results to the various physical models that are used
Fatigue and fracture mechanics in pressure vessels and piping. PVP-Volume 304
International Nuclear Information System (INIS)
Mehta, H.S.; Wilkowski, G.; Takezono, S.; Bloom, J.; Yoon, K.; Aoki, S.; Rahman, S.; Nakamura, T.; Brust, F.; Yoshimura, S.
1995-01-01
Fracture mechanics and fatigue evaluations are an important part of the structural integrity analyses to assure safe operation of pressure vessels and piping components during their service life. The paper presented in this volume illustrate the application of fatigue and fracture mechanics techniques to assess the structural integrity of a wide variety of Pressure Vessels and Piping components. The papers are organized in six sections: (1) fatigue and fracture--vessels; (2) fatigue and fracture--piping; (3) fatigue and fracture--material property evaluations; (4) constraint effects in fracture mechanics; (5) probabilistic fracture mechanics analyses; and (6) user's experience with failure assessment diagrams. Separate abstracts were prepared for most of the papers in this book
Prediction of gas volume fraction in fully-developed gas-liquid flow in a vertical pipe
International Nuclear Information System (INIS)
Islam, A.S.M.A.; Adoo, N.A.; Bergstrom, D.J.; Wang, D.F.
2015-01-01
An Eulerian-Eulerian two-fluid model has been implemented for the prediction of the gas volume fraction profile in turbulent upward gas-liquid flow in a vertical pipe. The two-fluid transport equations are discretized using the finite volume method and a low Reynolds number κ-ε turbulence model is used to predict the turbulence field for the liquid phase. The contribution to the effective turbulence by the gas phase is modeled by a bubble induced turbulent viscosity. For the fully-developed flow being considered, the gas volume fraction profile is calculated using the radial momentum balance for the bubble phase. The model potentially includes the effect of bubble size on the interphase forces and turbulence model. The results obtained are in good agreement with experimental data from the literature. The one-dimensional formulation being developed allows for the efficient assessment and further development of both turbulence and two-fluid models for multiphase flow applications in the nuclear industry. (author)
Prediction of gas volume fraction in fully-developed gas-liquid flow in a vertical pipe
Energy Technology Data Exchange (ETDEWEB)
Islam, A.S.M.A.; Adoo, N.A.; Bergstrom, D.J., E-mail: nana.adoo@usask.ca [University of Saskatchewan, Department of Mechanical Engineering, Saskatoon, SK (Canada); Wang, D.F. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)
2015-07-01
An Eulerian-Eulerian two-fluid model has been implemented for the prediction of the gas volume fraction profile in turbulent upward gas-liquid flow in a vertical pipe. The two-fluid transport equations are discretized using the finite volume method and a low Reynolds number κ-ε turbulence model is used to predict the turbulence field for the liquid phase. The contribution to the effective turbulence by the gas phase is modeled by a bubble induced turbulent viscosity. For the fully-developed flow being considered, the gas volume fraction profile is calculated using the radial momentum balance for the bubble phase. The model potentially includes the effect of bubble size on the interphase forces and turbulence model. The results obtained are in good agreement with experimental data from the literature. The one-dimensional formulation being developed allows for the efficient assessment and further development of both turbulence and two-fluid models for multiphase flow applications in the nuclear industry. (author)
Computer program TMOC for calculating of pressure transients in fluid filled piping networks
International Nuclear Information System (INIS)
Siikonen, T.
1978-01-01
The propagation of a pressure wave in fluid filles tubes is significantly affected by the pipe wall motion and vice versa. A computer code TMOC (Transients by the Method of Characteristics) is being developed for the analysis of the coupled fluid and pipe wall transients. Because of the structural feedback, the pressure can be calculated more accurately than in the programs commonly used. (author)
Energy Technology Data Exchange (ETDEWEB)
Niedermayer, U., E-mail: u.niedermayer@gsi.de [Technische Universitaet Darmstadt, Institut fuer Theorie Elektromagnetischer Felder, Schlossgartenstrasse 8, 64289 Darmstadt (Germany); Boine-Frankenheim, O. [Technische Universitaet Darmstadt, Institut fuer Theorie Elektromagnetischer Felder, Schlossgartenstrasse 8, 64289 Darmstadt (Germany)
2012-09-21
The resistive wall impedance is one of the main sources for beam instabilities in synchrotrons and storage rings. The fast ramped SIS18 synchrotron at GSI and the projected SIS100 synchrotron for FAIR both employ thin (0.3 mm) stainless steel beam pipes in order to reduce eddy current effects. The lowest betatron sidebands are at about 100 kHz, which demands accurate impedance predictions in the low frequency (LF) range where the beam pipe and possibly also the structures behind the pipe are the dominating impedance sources. The longitudinal and transverse resistive wall impedances of a circular multi-layer pipe are calculated analytically using the field matching technique. We compare the impedances obtained from a radial wave model, which corresponds to the setup used in bench measurements, with the axial wave model, which corresponds to an actual beam moving with relativistic velocity. For thin beam pipes the induced wall current and the corresponding shielding properties of the pipe are important. In both models the wall current is obtained analytically. The characteristic frequencies for the onset of the wall current are calculated from equivalent lumped element circuits corresponding to the radial model. For more complex structures, like the SIS100 beam pipe, we use a numerical method, in which the impedance is obtained from the total power loss. The method is validated by the analytic expressions for circular beam pipes.
International Nuclear Information System (INIS)
Chan, A.L.; Curtis, D.J.; Rybicki, E.F.; Lu, S.C.
1981-08-01
This volume describes the analyses used to evaluate stresses due to loads other than seismic excitations in the primary coolant loop piping of a selected four-loop pressurized water reactor nuclear power station. The results of the analyses are used as input to a simulation procedure for predicting the probability of pipe fracture in the primary coolant system. Sources of stresses considered in the analyses are pressure, dead weight, thermal expansion, thermal gradients through the pipe wall, residual welding, and mechanical vibrations. Pressure and thermal transients arising from plant operations are best estimates and are based on actual plant operation records supplemented by specified plant design conditions. Stresses due to dead weight and thermal expansion are computed from a three-dimensional finite element model that uses a combination of pipe, truss, and beam elements to represent the reactor coolant loop piping, reactor pressure vessel, reactor coolant pumps, steam generators, and the pressurizer. Stresses due to pressure and thermal gradients are obtained by closed-form solutions. Calculations of residual stresses account for the actual heat impact, welding speed, weld preparation geometry, and pre- and post-heat treatments. Vibrational stresses due to pump operation are estimated by a dynamic analysis using existing measurements of pump vibrations
Subprogram Calculating The Distance Between Pipe And Plane For Automatic Piping System Design
International Nuclear Information System (INIS)
Satmoko, Ari
2001-01-01
DISTLNPL subprogram was created using Auto LISP software. This subprogram is planned to complete CAPD (Computer Aided Piping Design) software being developed. The CAPD works under the following method: suggesting piping system line and evaluating whether any obstacle allows the proposed line to be constructed. DISTLNPL is able to compute the distance between pipe and any equipment having plane dimension such as wall, platform, floors, and so on. The pipe is modeled by using a line representing its axis, and the equipment is modeled using a plane limited by some lines. The obtained distance between line and plane gives information whether the pipe crosses the equipment. In the case of crashing, the subprogram will suggest an alternative point to be passed by piping system. So far, DISTLNPL has not been able to be accessed by CAPD yet. However, this subprogram promises good prospect in modeling wall, platform, and floors
Calculation and evaluation methodology of the flawed pipe and the compute program development
International Nuclear Information System (INIS)
Liu Chang; Qian Hao; Yao Weida; Liang Xingyun
2013-01-01
Background: The crack will grow gradually under alternating load for a pressurized pipe, whereas the load is less than the fatigue strength limit. Purpose: Both calculation and evaluation methodology for a flawed pipe that have been detected during in-service inspection is elaborated here base on the Elastic Plastic Fracture Mechanics (EPFM) criteria. Methods: In the compute, the depth and length interaction of a flaw has been considered and a compute program is developed per Visual C++. Results: The fluctuating load of the Reactor Coolant System transients, the initial flaw shape, the initial flaw orientation are all accounted here. Conclusions: The calculation and evaluation methodology here is an important basis for continue working or not. (authors)
Reliability of piping system components. Volume 4: The pipe failure event database
Energy Technology Data Exchange (ETDEWEB)
Nyman, R; Erixon, S [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Tomic, B [ENCONET Consulting GmbH, Vienna (Austria); Lydell, B [RSA Technologies, Visat, CA (United States)
1996-07-01
Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A `data driven and systems oriented` analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs.
Reliability of piping system components. Volume 4: The pipe failure event database
International Nuclear Information System (INIS)
Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.
1996-07-01
Available public and proprietary databases on piping system failures were searched for relevant information. Using a relational database to identify groupings of piping failure modes and failure mechanisms, together with insights from published PSAs, the project team determined why, how and where piping systems fail. This report represents a compendium of technical issues important to the analysis of pipe failure events, and statistical estimation of failure rates. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A 'data driven and systems oriented' analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failure. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 42 refs, 25 figs
Simplified method of calculating residual stress in circumferential welding of piping
International Nuclear Information System (INIS)
Umemoto, Tadahiro
1984-01-01
Many circumferential joints of piping are used in as-welded state, but in these welded joints, the residual stress as high as the yield stress of materials arises, and causes to accelerate stress corrosion cracking and corrosion fatigue. The experiment or the finite element method to clarify welding residual stress requires much time and labor, and is expensive, therefore, the author proposed the simplified method of calculation. The heating and cooling process of welding is very complex, and cannot be modeled as it is, therefore, it was assumed that in multiple layer welding, the welding condition of the last layer determines the residual stress, that material constants are invariable regardless of temperature, that the temperature distribution and residual stress are axisymmetric, and that there is repeated stress-strain relation in the vicinity of welded parts. The temperature distribution at the time of welding, thermal stress and welding residual stress are analyzed, and the material constants used for the calculation of residual stress are given. As the example of calculation, the effect of welding heat input and materials is shown. The extension of the method to a thick-walled pipe is discussed. (Kako, I.)
International Nuclear Information System (INIS)
Grebner, H.; Hoefler, A.; Hunger, H.
1991-01-01
In this paper calculations to experiments on leak opening and leak rates of piping components are presented. The experiments are performed at the HDR-facility at Karlstein/Germany and up to now straight pipes and pipe branches were considered. Numerical and experimental results are compared. (author)
Energy Technology Data Exchange (ETDEWEB)
Swidersky, Harald; Thiele, Thomas [TUeV Sued Industrie Service GmbH, Muenchen (Germany)
2012-11-01
Filling procedures in piping systems are usually not load cases that are studied by fluid dynamic and structure dynamic analyses with respect to the integrity of pipes and supports. Although, their frequency is higher than that of postulated accidental transients, therefore they have to be considered for fatigue analyses. The piping and support loads due to filling procedures are caused by the density differences if the transported fluids, for instance in flows with the transport of gas bubbles. The impact duration of the momentum forces is defined by the flow velocity and the length of discontinuities in the piping segments. Filling procedures end very often with a shock pressure, caused by the impact and decelerating of the fluid front at smaller cross sections. The suitability of the thermally hydraulics program RELAP/MOD3.3 for the calculation of realistic loads from filling procedures was studied, the results compared with experimental data. It is shown that dependent on the discretization level the loads are partial significantly underestimated.
Development of Automatic Visceral Fat Volume Calculation Software for CT Volume Data
Directory of Open Access Journals (Sweden)
Mitsutaka Nemoto
2014-01-01
Full Text Available Objective. To develop automatic visceral fat volume calculation software for computed tomography (CT volume data and to evaluate its feasibility. Methods. A total of 24 sets of whole-body CT volume data and anthropometric measurements were obtained, with three sets for each of four BMI categories (under 20, 20 to 25, 25 to 30, and over 30 in both sexes. True visceral fat volumes were defined on the basis of manual segmentation of the whole-body CT volume data by an experienced radiologist. Software to automatically calculate visceral fat volumes was developed using a region segmentation technique based on morphological analysis with CT value threshold. Automatically calculated visceral fat volumes were evaluated in terms of the correlation coefficient with the true volumes and the error relative to the true volume. Results. Automatic visceral fat volume calculation results of all 24 data sets were obtained successfully and the average calculation time was 252.7 seconds/case. The correlation coefficients between the true visceral fat volume and the automatically calculated visceral fat volume were over 0.999. Conclusions. The newly developed software is feasible for calculating visceral fat volumes in a reasonable time and was proved to have high accuracy.
Study on pressure pulsation and piping vibration of complex piping of reciprocating compressor
International Nuclear Information System (INIS)
Xu Bin; Feng Quanke; Yu Xiaoling
2008-01-01
This paper presents a preliminary research on the piping vibration and pressure pulsation of reciprocating compressor piping system. On the basis of plane wave theory, the calculation of gas column natural frequency and pressure pulsation in complex pipelines is done by using the transfer matrix method and stiffness matrix method, respectively. With the discretization method of FEM, a mathematical model for calculating the piping vibration and stress of reciprocating compressor piping system is established, and proper boundary conditions are proposed. Then the structural modal and stress of the piping system are calculated with CAESAR II. The comparison of measured and calculated values found that the one dimensional wave equation can accurately calculate the natural frequency and pressure pulsation in gas column of piping system for reciprocating compressor. (authors)
International Nuclear Information System (INIS)
Anon.
1993-01-01
This WRC Bulletin concentrates on the major failure mechanisms observed in nuclear power plant piping during the past three decades and on corrective actions taken to minimize or eliminate such failures. These corrective actions are applicable to both replacement piping and the next generation of light-water reactors. This WRC Bulletin was written with the objective of meeting a need for piping criteria in Advanced Light-Water Reactors, but there is application well beyond the LWR industry. This Volume, in particular, is equally applicable to current nuclear power plants, fossil-fueled power plants, and chemical plants including petrochemical. Implementation of the recommendations for mitigation of specific problems should minimize severe failures or cracking and provide substantial economic benefit. This volume uses a case history approach to high-light various failure mechanisms and the corrective actions used to resolve such failures. Particular attention is given to those mechanisms leading to severe piping failures, where severe denotes complete severance, large ''fishmouth'' failures, or long throughwall cracks releasing a minimum of 50 gpm. The major failure mechanisms causing severe failure are erosion-corrosion and vibrational fatigue. Stress corrosion cracking also has been a common problem in nuclear piping systems. In addition thermal fatigue due to mixing-tee and to thermal stratification also is discussed as is microbiologically-induced corrosion. Finally, water hammer, which represents the ultimate in internally-generated dynamic high-energy loads, is discussed
Non-destructive technique to verify clearance of pipes
Directory of Open Access Journals (Sweden)
Savidou Anastasia
2010-01-01
Full Text Available A semi-empirical, non-destructive technique to evaluate the activity of gamma ray emitters in contaminated pipes is discussed. The technique is based on in-situ measurements by a portable NaI gamma ray spectrometer. The efficiency of the detector for the pipe and detector configuration was evaluated by Monte Carlo calculations performed using the MCNP code. Gamma ray detector full-energy peak efficiency was predicted assuming a homogeneous activity distribution over the internal surface of the pipe for 344 keV, 614 keV, 662 keV, and 1332 keV photons, representing Eu-152, Ag-118m, Cs-137, and Co-60 contamination, respectively. The effect of inhomogeneity on the accuracy of the technique was also examined. The model was validated against experimental measurements performed using a Cs-137 volume calibration source representing a contaminated pipe and good agreement was found between the calculated and experimental results. The technique represents a sensitive and cost-effective technology for calibrating portable gamma ray spectrometry systems and can be applied in a range of radiation protection and waste management applications.
Pipe failure probability - the Thomas paper revisited
International Nuclear Information System (INIS)
Lydell, B.O.Y.
2000-01-01
Almost twenty years ago, in Volume 2 of Reliability Engineering (the predecessor of Reliability Engineering and System Safety), a paper by H. M. Thomas of Rolls Royce and Associates Ltd. presented a generalized approach to the estimation of piping and vessel failure probability. The 'Thomas-approach' used insights from actual failure statistics to calculate the probability of leakage and conditional probability of rupture given leakage. It was intended for practitioners without access to data on the service experience with piping and piping system components. This article revisits the Thomas paper by drawing on insights from development of a new database on piping failures in commercial nuclear power plants worldwide (SKI-PIPE). Partially sponsored by the Swedish Nuclear Power Inspectorate (SKI), the R and D leading up to this note was performed during 1994-1999. Motivated by data requirements of reliability analysis and probabilistic safety assessment (PSA), the new database supports statistical analysis of piping failure data. Against the background of this database development program, the article reviews the applicability of the 'Thomas approach' in applied risk and reliability analysis. It addresses the question whether a new and expanded database on the service experience with piping systems would alter the original piping reliability correlation as suggested by H. M. Thomas
Energy Technology Data Exchange (ETDEWEB)
Hinson-Rider, G.
1977-10-04
A fluid carrying pipe is described having an integral transparent portion formed into a longitudinally extending cylindrical lens that focuses solar heat rays to a focal axis within the volume of the pipe. The pipe on the side opposite the lens has a heat ray absorbent coating for absorbing heat from light rays that pass through the focal axis.
International Nuclear Information System (INIS)
Stevenson, J.D.
1995-11-01
Volume 2 of the ''Survey of Strong Motion Earthquake Effects on Thermal Power Plants in California with Emphasis on Piping Systems'' contains Appendices which detail the detail design and seismic response of several power plants subjected to strong motion earthquakes. The particular plants considered include the Ormond Beach, Long Beach and Seal Beach, Burbank, El Centro, Glendale, Humboldt Bay, Kem Valley, Pasadena and Valley power plants. Included is a typical power plant piping specification and photographs of typical power plant piping specification and photographs of typical piping and support installations for the plants surveyed. Detailed piping support spacing data are also included
An example demonstrating the conservatism of pipe calculations using KTA safety standard 3201.2
International Nuclear Information System (INIS)
Zeitner, W.
1991-01-01
The conservatism of the code calculation is demonstrated by using an example of a highly stressed pipe subject to internal pressure and a dynamic bending moment. For this reason the allowable code loadings are compared with the load carrying capacity, which is derived by realistic analysis (plastic strains) and experiment. The latter analysis is based on measured stress-strain curves of materials and stresses at which crack initiation occurs. The experiment shows that the pipe is capable of withstanding considerably higher loads than the code permits. The realistic analysis explains this discrepancy. (orig.)
International Nuclear Information System (INIS)
Park, Byoung Yoon; Hansen, Francis D.
2004-01-01
The regulatory compliance determination for the Waste Isolation Pilot Plant includes the consideration of room closure. Elements of the geomechanical processes include salt creep, gas generation and mechanical deformation of the waste residing in the rooms. The WIPP was certified as complying with regulatory requirements based in part on the implementation of room closure and material models for the waste. Since the WIPP began receiving waste in 1999, waste packages have been identified that are appreciably more robust than the 55-gallon drums characterized for the initial calculations. The pipe overpack comprises one such waste package. This report develops material model parameters for the pipe overpack containers by using axisymmetrical finite element models. Known material properties and structural dimensions allow well constrained models to be completed for uniaxial, triaxial, and hydrostatic compression of the pipe overpack waste package. These analyses show that the pipe overpack waste package is far more rigid than the originally certified drum. The model parameters developed in this report are used subsequently to evaluate the implications to performance assessment calculations
International Nuclear Information System (INIS)
Zareei, A.; Nabavi, S.M.
2016-01-01
In this paper, stress intensity factors are calculated at the deepest point of an internal circumferential semi-elliptical crack in a pipe subjected to any arbitrary load. Based on the three dimensional finite element analysis, a weight function is proposed for high aspect ratio semi-elliptical cracks in pipes. An effective expression is developed analytically to evaluate the stress intensity factor using the weight function method. For several crack face stress fields and welding residual stress distributions, the weight function is validated against finite element data and those in the literature. Based on the comparison results, it can be concluded that the solution proposed in this paper is effective in engineering applications. - Highlights: • Analysis of internal circumferential semi-elliptical cracks with high aspect ratio in pipes. • A weight function is proposed for the calculation of the stress intensity factors for the deepest point of the crack. • An effective closed form expression is proposed to evaluate the stress intensity factors. • Prediction of stress intensity factors for any applied stress gradients through the wall thickness without any limitations. • A three-dimensional finite element modeling employs to calculate the stress intensity factors for different geometries.
Energy Technology Data Exchange (ETDEWEB)
Sigg, K. C.; Coffield, R. D.
2002-09-01
High Reynolds number test data has recently been reported for both single and multiple piping elbow design configurations at earlier ASME Fluid Engineering Division conferences. The data of these studies ranged up to a Reynolds number of 42 x 10[sup]6 which is significantly greater than that used to establish design correlations before the data was available. Many of the accepted design correlations, based on the lower Reynolds number data, date back as much as fifty years. The new data shows that these earlier correlations are extremely conservative for high Reynolds number applications. Based on the recent high Reynolds number information a new recommended method has been developed for calculating irrecoverable pressure loses in piping systems for design considerations such as establishing pump sizing requirements. This paper describes the recommended design approach and additional testing that has been performed as part of the qualification of the method. This qualification testing determined the irrecoverable pressure loss of a piping configuration that would typify a limiting piping section in a complicated piping network, i.e., multiple, tightly coupled, out-of-plane elbows in series under high Reynolds number flow conditions. The overall pressure loss measurements were then compared to predictions, which used the new methodology to assure that conservative estimates for the pressure loss (of the type used for pump sizing) were obtained. The recommended design methodology, the qualification testing and the comparison between the predictions and the test data are presented. A major conclusion of this study is that the recommended method for calculating irrecoverable pressure loss in piping systems is conservative yet significantly lower than predicted by early design correlations that were based on the extrapolation of low Reynolds number test data.
Hydraulic simulation of the systems of a nuclear power plant for charges calculation in piping
International Nuclear Information System (INIS)
Masriera, N.
1990-01-01
This work presents a general description of the methodology used by the ENACE S.A. Fluids Working Group for hydraulics simulation of a nuclear power plant system for the calculation charges in piping. (Author) [es
International Nuclear Information System (INIS)
Fan Chunli; Sun Fengrui; Yang Li
2008-01-01
In the paper, the irregular configuration of the inner pipe boundary is identified based on the estimation of the circumferential distribution of the effective thermal conductivity of pipe wall. In order to simulate the true temperature measurement in the numerical examples, the finite element method is used to calculate the temperature distribution at the outer pipe surface based on the irregular shaped inner pipe boundary to be determined. Then based on this simulated temperature distribution the inverse identification work is conducted by employing the modified one-dimensional correction method, along with the finite volume method, to estimate the circumferential distribution of the effective thermal conductivity of the pipe wall. Thereafter, the inner pipe boundary shape is calculated based on the conductivity estimation result. A series of numerical experiments with different temperature measurement errors and different thermal conductivities of pipe wall have certified the effectiveness of the method. It is proved that the method is a simple, fast and accurate one for this inverse heat conduction problem.
Assessment of RELAP5/MOD3.2.2γ against flooding database in horizontal-to-inclined pipes
International Nuclear Information System (INIS)
Kim, Hyoung Tae; No, Hee Cheon
2001-01-01
A total of 356 experimental data for the onset of flooding are compiled for the data bank and used for the assessment of RELAP5/MOD3.2.2γ predictions of Counter-Current Flow Limitation (CCFL) in horizontal-to-inclined pipes simulating a PWR hot leg. The predictions of the flooding gas velocity in the database are known to be largely dependent on the horizontal pipe length-to-diameter ratio (L/D). RELAP5 calculations are compared with the experimental data where L/D is varied within the range of database. The present input model used for the simulation of CCFL is validated to reasonably calculate the gradient of water level in the horizontal pipes connected with the inclined volumes. RELAP5 calculations show that the RELAP5 predicts the flooding points qualitatively well but higher gas flow rate is required to initiate the flooding compared with the experimental data if the L/D is as low that of the hot legs of typical PWRs. Standard RELAP5 code is modified to apply the user specified CCFL curve not only to veritical volumes but also to the horizontal volumes. The calculation value by the modified version lies well on the applied CCFL curve even if flooding occurs at lower gas velocity thatn predicted by the CCFL curve in standard RELAP5
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
Project W-320, 241-C-106 sluicing: Piping calculations. Volume 8
International Nuclear Information System (INIS)
Bailey, J.W.
1998-01-01
This supporting document has been prepared to make the FDNW calculations for Project W-320 readily retrievable. The objective of this calculation is to perform the hydraulic analysis on the slurry line and the supernate line for W-320. This calculation will use the As-Built conditions of the slurry line and the supernate line. Booster Pump Curves vs System Curves shall be generated for the supernate system and the slurry system
Project W-320, 241-C-106 sluicing HVAC calculations, Volume 4
Energy Technology Data Exchange (ETDEWEB)
Bailey, J.W.
1998-07-30
This supporting document has been prepared to make the FDNW calculations for Project W-320, readily retrievable. The report contains the following design calculations: Cooling load in pump pit 241-AY-102; Pressure relief seal loop design; Process building piping stress analysis; Exhaust skid maximum allowable leakage criteria; and Recirculation heat, N509 duct requirements.
Theoretical and experimental investigation of the performance of solar thermosyphon heat pipe
International Nuclear Information System (INIS)
Hamidi, A.A.; Khalji Asadi, M.; Yousefi, L.; Moeini, G.
2001-01-01
Thermosyphon is a kind of heat pipe consisting of a tube which after through degassing has been filled with the required working fluid under vacuum, the pipe is equipped with wide fines on both sides in order to absorb solar radiation effectively. In order to eliminate conduction and convection heat transfer phenomena the tube is situated inside an evacuated glass bulb. In order to increase the efficiency and improve the design and working conditions of various types of heat pipes, a fundamental knowledge of the variation of operating parameters inside the heat pipes is necessary. In this paper, effective operating parameters of a thermosyphon heat pipe in uniform and steady condition are studied. These parameters include saturation temperature of the fluid inside the pipe, the variation of liquid and vapor flow rates inside the pipe and finally the pressure drop of liquid and vapor along the length of the pipe. The modeling is first started by writing an energy balance for the control volume of the pipe so that a first approximation for the above mentioned parameters is obtained. In this balance, depending on the type of fluid next to the condenser section and the type of heat transfer phenomena (free or forced convection) and also with due regards to the experimental correlations available, first the Nusselt number and then the heat transfer coefficient is calculated. From the latter, a first estimate of the required values for the liquid and vapor flow rates are found to be 0.222 and 0.0001126 Kg/s, respectively. The thickness of the film was determined to be 0.2 mm. In order to calculate the variations of the above mentioned parameters along the length of the tube, mass heat and momentum balances were written in next step for the control volumes on the liquid film, vapor phase and the system as a whole. Diagrams of these variations were obtained. The results were compared with both the data available in the literature and the experimental findings of a heat
Calculated Atomic Volumes of the Actinide Metals
DEFF Research Database (Denmark)
Skriver, H.; Andersen, O. K.; Johansson, B.
1979-01-01
The equilibrium atomic volume is calculated for the actinide metals. It is possible to account for the localization of the 5f electrons taking place in americium.......The equilibrium atomic volume is calculated for the actinide metals. It is possible to account for the localization of the 5f electrons taking place in americium....
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.
International Nuclear Information System (INIS)
Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang
2000-01-01
Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning
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.)
International Nuclear Information System (INIS)
Short, W.E.; Zamrik, S.Y.
1985-01-01
State-of-the-art engineering practices in pressure vessel and piping technology are the result of continual efforts in the evaluation of problems which have been experienced and the development of appropriate design and analysis methods for those applications. The resulting advances in technology benefit industry with properly engineered, safe, cost-effective pressure vessels and piping systems. To this end, advanced study continues in specialized areas of mechanical engineering such as fracture mechanics, experimental stress analysis, high pressure applications and related material considerations, as well as advanced techniques for evaluation of commonly encountered design problems. This volume is comprised of current technical papers on various aspects of fracture, fatigue and advanced mechanics as related to the design and analysis of pressure vessels and piping
Energy Technology Data Exchange (ETDEWEB)
Millman, D. L. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States); Griesheimer, D. P.; Nease, B. R. [Bechtel Marine Propulsion Corporation, Bertis Atomic Power Laboratory (United States); Snoeyink, J. [Dept. of Computer Science, Univ. of North Carolina at Chapel Hill (United States)
2012-07-01
In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)
International Nuclear Information System (INIS)
Millman, D. L.; Griesheimer, D. P.; Nease, B. R.; Snoeyink, J.
2012-01-01
In this paper we consider a new generalized algorithm for the efficient calculation of component object volumes given their equivalent constructive solid geometry (CSG) definition. The new method relies on domain decomposition to recursively subdivide the original component into smaller pieces with volumes that can be computed analytically or stochastically, if needed. Unlike simpler brute-force approaches, the proposed decomposition scheme is guaranteed to be robust and accurate to within a user-defined tolerance. The new algorithm is also fully general and can handle any valid CSG component definition, without the need for additional input from the user. The new technique has been specifically optimized to calculate volumes of component definitions commonly found in models used for Monte Carlo particle transport simulations for criticality safety and reactor analysis applications. However, the algorithm can be easily extended to any application which uses CSG representations for component objects. The paper provides a complete description of the novel volume calculation algorithm, along with a discussion of the conjectured error bounds on volumes calculated within the method. In addition, numerical results comparing the new algorithm with a standard stochastic volume calculation algorithm are presented for a series of problems spanning a range of representative component sizes and complexities. (authors)
Energy Technology Data Exchange (ETDEWEB)
Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P. [Battelle, Columbus, OH (United States)
1995-04-01
This is the seventh progress report of the U.S. Nuclear Regulatory Commission`s research program entitled {open_quotes}Short Cracks in Piping and Piping Welds{close_quotes}. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports.
International Nuclear Information System (INIS)
Wilkowski, G.M.; Ghadiali, N.; Rudland, D.; Krishnaswamy, P.; Rahman, S.; Scott, P.
1995-04-01
This is the seventh progress report of the U.S. Nuclear Regulatory Commission's research program entitled open-quotes Short Cracks in Piping and Piping Weldsclose quotes. The program objective is to verify and improve fracture analyses for circumferentially cracked large-diameter nuclear piping with crack sizes typically used in leak-before-break (LBB) analyses and in-service flaw evaluations. All work in the eight technical tasks have been completed. Ten topical reports are scheduled to be published. Progress only during the reporting period, March 1993 - December 1994, not covered in the topical reports is presented in this report. Details about the following efforts are covered in this report: (1) Improvements to the two computer programs NRCPIPE and NRCPIPES to assess the failure behavior of circumferential through-wall and surface-cracked pipe, respectively; (2) Pipe material property database PIFRAC; (3) Circumferentially cracked pipe database CIRCUMCK.WKI; (4) An assessment of the proposed ASME Section III design stress rule changes on pipe flaw tolerance; and (5) A pipe fracture experiment on a section of pipe removed from service degraded by microbiologically induced corrosion (MIC) which contained a girth weld crack. Progress in the other tasks is not repeated here as it has been covered in great detail in the topical reports
Energy Technology Data Exchange (ETDEWEB)
Nyman, R; Erixon, S; Tomic, B; Lydell, B
1995-12-01
SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today`s PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs.
International Nuclear Information System (INIS)
Nyman, R.; Erixon, S.; Tomic, B.; Lydell, B.
1995-12-01
SKI has undertaken a multi-year research project to establish a comprehensive passive component failure database, validate failure rate parameter estimates and establish a model framework for integrating passive component failures in existing PSAs. Phase 1 of the project produced a relational database on worldwide piping system failure events in the nuclear and chemical industries. This phase 2 report gives a graphical presentation of piping system operating experience, and compares key failure mechanisms in commercial nuclear power plants and chemical process industry. Inadequacies of traditional PSA methodology are addressed, with directions for PSA methodology enhancements. A data-driven-and-systems-oriented analysis approach is proposed to enable assignment of unique identities to risk-significant piping system component failures. Sufficient operating experience does exist to generate quality data on piping failures. Passive component failures should be addressed by today's PSAs to allow for aging analysis and effective, on-line risk management. 111 refs, 36 figs, 20 tabs
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)
PIPE STRESS and VERPIP codes for stress analysis and verifications of PEC reactor piping
International Nuclear Information System (INIS)
Cesari, F.; Ferranti, P.; Gasparrini, M.; Labanti, L.
1975-01-01
To design LMFBR piping systems following ASME Sct. III requirements unusual flexibility computer codes are to be adopted to consider piping and its guard-tube. For this purpose PIPE STRESS code previously prepared by Southern-Service, has been modified. Some subroutine for detailed stress analysis and principal stress calculations on all the sections of piping have been written and fitted in the code. Plotter can also be used. VERPIP code for automatic verifications of piping as class 1 Sct. III prescriptions has been also prepared. The results of PIPE STRESS and VERPIP codes application to PEC piping are in section III of this report
International Nuclear Information System (INIS)
Schoenfelder, C.; Kellner, A.
1985-01-01
An approximated representative part of a PWR-feed-water-line was modelled and used to calculate the displacements of the piping system and the loads on it, caused by pressure pulse due to pump failure and subsequent check valve closure. The computation was performed with the code SAPHYR which contains the fluid code ROLAST and the structure code SAPIENS, calculating simultaneously and interactively. The results were compared with an uncoupled calculation without fluid/structure interaction. It was shown that neglecting the fluid/structure interaction can lead to considerable overestimations - in some cases up to a factor of 3 - of the loads on the structures. (orig.)
Gas-Induced Water-hammer Loads Calculation for Safety Related Systems
Energy Technology Data Exchange (ETDEWEB)
Lee, Seungchan; Yoon, Dukjoo [Korea Hydro and Nuclear Power Co., LTd, Daejeon (Korea, Republic of); Lee, Dooyong [Seoul National Univ., Seoul (Korea, Republic of)
2013-05-15
Of particular interest, gas accumulation can result in system pressure transient in pump discharge piping following a pump start. Consequently, this evolves into a gas-water, a water-hammer event and the accompanying force imbalances on the piping segments can be sufficient to challenge the piping supports and restraint. This paper describes an method performing to the water-hammer loads to determine the maximum loading that would occur in the piping system following the safety injection signal and to evaluate its integrity. For a given gas void volumes in the discharge piping, the result of the calculation shows the maximum loads of 18,894.2psi, which is smaller than the allowable criteria. Also, the maximum peak axial force imbalances acting on the support is 1,720lbf as above.
Gas-Induced Water-hammer Loads Calculation for Safety Related Systems
International Nuclear Information System (INIS)
Lee, Seungchan; Yoon, Dukjoo; Lee, Dooyong
2013-01-01
Of particular interest, gas accumulation can result in system pressure transient in pump discharge piping following a pump start. Consequently, this evolves into a gas-water, a water-hammer event and the accompanying force imbalances on the piping segments can be sufficient to challenge the piping supports and restraint. This paper describes an method performing to the water-hammer loads to determine the maximum loading that would occur in the piping system following the safety injection signal and to evaluate its integrity. For a given gas void volumes in the discharge piping, the result of the calculation shows the maximum loads of 18,894.2psi, which is smaller than the allowable criteria. Also, the maximum peak axial force imbalances acting on the support is 1,720lbf as above
Fatigue evaluation of socket welded piping in nuclear power plant
International Nuclear Information System (INIS)
Vecchio, R.S.
1996-01-01
Fatigue failures in piping systems occur, almost without exception, at the welded connections. In nuclear power plant systems, such failures occur predominantly at the socket welds of small diameter piping ad fillet attachment welds under high-cycle vibratory conditions. Nearly all socket weld fatigue failures are identified by leaks which, though not high in volume, generally are costly due to attendant radiological contamination. Such fatigue cracking was recently identified in the 3/4 in. diameter recirculation and relief piping socket welds from the reactor coolant system (RCS) charging pumps at a nuclear power plant. Consequently, a fatigue evaluation was performed to determine the cause of cracking and provide an acceptable repair. Socket weld fatigue life was evaluated using S-N type fatigue life curves for welded structures developed by AASHTO and the assessment of an effective cyclic stress range adjacent to each socket weld. Based on the calculated effective tress ranges and assignment of the socket weld details to the appropriate AASHTO S-N curves, the socket weld fatigue lives were calculated and found to be in excellent agreement with the accumulated cyclic life to-date
International Nuclear Information System (INIS)
1998-02-01
Pipe Crawler reg-sign is a pipe surveying system for performing radiological characterization and/or free release surveys of piping systems. The technology employs a family of manually advanced, wheeled platforms, or crawlers, fitted with one or more arrays of thin Geiger Mueller (GM) detectors operated from an external power supply and data processing unit. Survey readings are taken in a step-wise fashion. A video camera and tape recording system are used for video surveys of pipe interiors prior to and during radiological surveys. Pipe Crawler reg-sign has potential advantages over the baseline and other technologies in areas of cost, durability, waste minimization, and intrusiveness. Advantages include potentially reduced cost, potential reuse of the pipe system, reduced waste volume, and the ability to manage pipes in place with minimal disturbance to facility operations. Advantages over competing technologies include potentially reduced costs and the ability to perform beta-gamma surveys that are capable of passing regulatory scrutiny for free release of piping systems
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
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
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
International Nuclear Information System (INIS)
Bjoerndahl, Olof; Letzter, Adam; Marcinkiewicz, Jerzy; Segle, Peter
2007-03-01
Transient thermohydraulic events often control the design of piping systems in nuclear power plants. Water hammers due to valve closure, pressure transients caused by steam collapse and pipe break all result in structural loads that are characterised by a high frequency content. What also characterises these pressures/forces is the specific spatial and time dependence that is acting on the piping system and found in the wave propagation in the contained fluid. The aim with this project has been to develop recommendations for analysis of the stress response in piping systems subjected to thermohydraulic transients. Basis for this work is that the so called two-step-method is applied and that the structural response is calculated with modal superposition. Derived analysis criteria are based on the assumption that the associated volume strain energy in the wave propagation for the contained fluid may be well defined by a parameter, here called ε PN . The stress response in the piping system is assumed to be completely determined with certain accuracy for that part of the volume strain energy in the wave propagation associated with this parameter. A comprehensive work has been done to determine the accuracy in loadings calculated with RELAP5. Properties such as period elongation and associated spurious oscillations in the pressure wave transient have been investigated. Furthermore, has the characteristics of the artificial numerical damping in RELAP5 been identified. Based on desired accuracy of the thermohydraulic analysis together with knowledge about the duration of the thermohydraulic perturbation, the lowest upper frequency limit f Pipe , in the modal base that is required for the structure model is calculated. With perturbation is meant such as a valve closure. According to suggested criteria and with the upper frequency limit set, the essential parameters i) largest size of the elements in the structure model and ii) the largest applicable time step in the
Project W-320, 241-C-106 sluicing piping calculations, Volume 7
International Nuclear Information System (INIS)
Bailey, J.W.
1998-01-01
The object of this report is to calculate the hydraulic forces imposed at the sluicer nozzle. This is required by Project W-320 waste retrieval for tank 241-C-106. The method of analysis used is Bernoulli's momentum equation for stead flow
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
Ultrasound automated volume calculation in reproduction and in pregnancy.
Ata, Baris; Tulandi, Togas
2011-06-01
To review studies assessing the application of ultrasound automated volume calculation in reproductive medicine. We performed a literature search using the keywords "SonoAVC, sonography-based automated volume calculation, automated ultrasound, 3D ultrasound, antral follicle, follicle volume, follicle monitoring, follicle tracking, in vitro fertilization, controlled ovarian hyperstimulation, embryo volume, embryonic volume, gestational sac, and fetal volume" and conducted the search in PubMed, Medline, EMBASE, and the Cochrane Database of Systematic Reviews. Reference lists of identified reports were manually searched for other relevant publications. Automated volume measurements are in very good agreement with actual volumes of the assessed structures or with other validated measurement methods. The technique seems to provide reliable and highly reproducible results under a variety of conditions. Automated measurements take less time than manual measurements. Ultrasound automated volume calculation is a promising new technology which is already used in daily practice especially for assisted reproduction. Improvements to the technology will undoubtedly render it more effective and increase its use. Copyright © 2011 American Society for Reproductive Medicine. Published by Elsevier Inc. All rights reserved.
Energy Technology Data Exchange (ETDEWEB)
1984-11-01
The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations.
International Nuclear Information System (INIS)
1984-11-01
The Executive Director for Operations (EDO) in establishing the Piping Review Committee concurred in its overall scope that included an evaluation of the potential for pipe breaks. The Pipe Break Task Group has responded to this directive. This report summarizes a review of regulatory documents and contains the Task Group's recommendations for application of the leak-before-break (LBB) approach to the NRC licensing process. The LBB approach means the application of fracture mechanics technology to demonstrate that high energy fluid piping is very unlikely to experience double-ended ruptures or their equivalent as longitudinal or diagonal splits. The Task Group's reommendations and discussion are founded on current and ongoing NRC staff actions as presented in Section 3.0 of this report. Additional more detailed comments and discussion are presented in Section 5.0 and in Appendices A and B. The obvious issues are the reexamination of the large pipe break criteria and the implications of any changes in the criteria as they influence items such as jet loads and pipe whip. The issues have been considered and the Task Group makes the following recommendations
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)
Calculation of the major material parameters of heat carriers for cryogenic heat pipes
International Nuclear Information System (INIS)
Molt, W.
1976-07-01
In order to make predictions on the efficiency of cryogenic heat pipes, the material parameters of the heat carrier such as surface tension, viscosity, evaporation heat and density of the liquid should be known. The author therefore investigates suitable interpolation methods and equations which enable the calculation of the desired material parameter at a certain temperature from other known quantities or which require that 1 to 3 material parameters at different temperatures are known. The calculations are limited to the temperature between critical temperature and triple point, since this is the only temperature region in which the heat carrier is in its liquid phase. The applicability and exactness of the equations is tested using known experimental data on N 2 , O 2 , CH 4 and partly on CF 4 . (orig./TK) [de
International Nuclear Information System (INIS)
Aydin, A.; Durak, H.; Ucan, E.S.; Kaya, G.C.; Ceylan, E.; Kiter, G.
2004-01-01
Although extensive work has been done on cigarette smoking and its effects on pulmonary function, there are limited number of studies on water-pipe smoking. The effects of water-pipe smoking on health are not widely investigated. The aim of this study was to determine the effects of water-pipe smoking on pulmonary permeability. Technetium-99m DTPA inhalation scintigraphy was performed on 14 water-pipe smoker volunteers (all men, mean age 53.7±9.8) and 11 passive smoker volunteers (1 woman, 10 men, mean age 43.8±12). Clearance half-time (T 1/2) was calculated by placing a monoexponential fit on the time activity curves. Penetration index (PI) of the radioaerosol was also calculated. PI was 0.58±0.14 and 0.50±0.12 for water-pipe smokers (WPS) and passive smokers (PS) respectively. T 1/2 of peripheral lung was 57.3±12.7 and 64.6±13.2 min, central airways was 55.8±23.5 and 80.1±35.2 min for WPS and PS, respectively (p≤0.05). Forced expiratory volume in one second/forced vital capacity (FEV 1 /FVC)% was 82.1±8.5 (%) and 87.7±6.5 (%) for WPS and PS, respectively (0.025< p≤0.05). We suggest that water-pipe smoking effects pulmonary epithelial permeability more than passive smoking. Increased central mucociliary clearance in water-pipe smoking may be due to preserved humidity of the airway tracts. (author)
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.
Piping engineering for nuclear power plant
International Nuclear Information System (INIS)
Curto, N.; Schmidt, H.; Muller, R.
1988-01-01
In order to develop piping engineering, an adequate dimensioning and correct selection of materials must be secured. A correct selection of materials together with calculations and stress analysis must be carried out with a view to minimizing or avoiding possible failures or damages in piping assembling, which could be caused by internal pressure, weight, temperature, oscillation, etc. The piping project for a nuclear power plant is divided into the following three phases. Phase I: Basic piping design. Phase II: Final piping design. Phase III: Detail engineering. (Author)
International Nuclear Information System (INIS)
Stubbe, E.J.; VanHoenacker, L.; Otero, R.
1994-02-01
This report presents an assessment study for the use of the code RELAP 5/MOD3/5M5 in the calculation of transient hydrodynamic loads on safety and relief discharge pipes. Its predecessor, RELAP 5/MOD1, was found adequate for this kind of calculations by EPRI. The hydrodynamic loads are very important for the discharge piping design because of the fast opening of the valves and the presence of liquid in the upstream loop seals. The code results are compared to experimental load measurements performed at the Combustion Engineering Laboratory in Windsor (US). Those measurements were part of the PWR Valve Test Program undertaken by EPRI after the TMI-2 accident. This particular kind of transients challenges the applicability of the following code models: two-phase choked discharge; interphase drag in conditions with large density gradients; heat transfer to metallic structures in fast changing conditions; two-phase flow at abrupt expansions. The code applicability to this kind of transients is investigated. Some sensitivity analyses to different code and model options are performed. Finally, the suitability of the code and some modeling guidelines are discussed
Numerical study on heat transfer characteristics of thermosyphon heat pipes using nanofluids
International Nuclear Information System (INIS)
Huminic, Gabriela; Huminic, Angel
2013-01-01
Highlights: • Numerical study of nanofluid heat transfer in thermosyphon heat pipes is performed. • Effect of nanoparticle concentration and operating temperature are studied. • Fe 2 O 3 –water nanofluid with 5.3% volume concentration shows the best performance. • Results show the improvement the thermal performances of thermosyphon heat pipe with nanofluids. - Abstract: In this work, a three-dimensional analysis is used to investigate the heat transfer of thermosyphon heat pipe using water and nanofluids as the working fluid. The study focused mainly on the effects of volume concentrations of nanoparticles and the operating temperature on the heat transfer performance of the thermosyphon heat pipe using the nanofluids. The analysis was performed for water and γ-Fe 2 O 3 nanoparticles, three volume concentrations of nanoparticles (0 vol.%, 2 vol.% and 5.3 vol.%) and four operating temperatures (60, 70, 80 and 90 °C). The numerical results show that the volume concentration of nanoparticles had a significant effect in reducing the temperature difference between the evaporator and condenser. Experimental and numerical results show qualitatively that the thermosyphon heat pipe using the nanofluid has better heat transfer characteristics than the thermosyphon heat pipe using water
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.)
International Nuclear Information System (INIS)
Zheng Bin; Lu Yuechuan; Zang Fenggang; Sun Yingxue
2009-01-01
In order to widen the application of the engineering method of EPRI, with a series of analysis on the 3D elastic and elastic-plastic fracture mechanics finite element, the crack open displacements (COD) of cracked pipe were calculated and a key influence function h 2 in EPRI engineering method was studied against the COD results of FEM. A calculation method of h2 under the condition of tension and bending combined load was introduced in detail. In order to validate this method, the calculated h 2 results were compared with that of EPRI, and the calculated COD results based on the h 2 results were compared with that of PICEP. The compared results indicated that the calculated h 2 results as well as the COD results and the corresponding reference values were respectively accordant, and the calculation method in this paper was validated accordingly. (authors)
Superconducting pipes and levitating magnets.
Levin, Yan; Rizzato, Felipe B
2006-12-01
Motivated by a beautiful demonstration of the Faraday and the Lenz laws in which a small neodymium magnet falls slowly through a conducting nonferromagnetic tube, we consider the dynamics of a magnet falling coaxially through a superconducting pipe. Unlike the case of normal conducting pipes, in which the magnet quickly reaches the terminal velocity, inside a superconducting tube the magnet falls freely. On the other hand, to enter the pipe the magnet must overcome a large electromagnetic energy barrier. For sufficiently strong magnets, the barrier is so large that the magnet will not be able to penetrate it and will be levitated over the mouth of the pipe. We calculate the work that must done to force the magnet to enter a superconducting tube. The calculations show that superconducting pipes are very efficient at screening magnetic fields. For example, the magnetic field of a dipole at the center of a short pipe of radius a and length L approximately > a decays, in the axial direction, with a characteristic length xi approximately 0.26a. The efficient screening of the magnetic field might be useful for shielding highly sensitive superconducting quantum interference devices. Finally, the motion of the magnet through a superconducting pipe is compared and contrasted to the flow of ions through a trans-membrane channel.
Volume calculation from limited number of MR imaging sections
International Nuclear Information System (INIS)
Wang, J.; Mezrich, R.; Sebok, D.
1988-01-01
Magnetic resonance imaging is an accurate and noninvasive way to obtain cardiac geometrical information. For the quantification of left ventricular dynamic parameters, sections are taken along the long axis of the ventricle. Due to the limited number of sections that can be obtained in a reasonable amount of scanning time, the estimation of longitudinal dimension is usually the cause of error in volume calculation. The starting and ending sections are best estimated by guidance of the short axis cuts. This can only guarantee first-order accuracy. Simpson's rule for summation of areas to calculate volume, which is the commonly used method, assumes an accurate knowledge of the starting and ending points of integration. When this assumption is not perfectly met, Simpson's rule tends to unsystemically over- or underestimate the true volume. Due to this concern, some researchers adopt the images from the short axis cut to aid the volume calculation. This can improve the accuracy, but makes the already long scanning time longer. The authors have derived a method of extrapolation and intrapolation based on no more information than usually available to correct the volume over- or underestimated by the Simpson's rule
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.
International Nuclear Information System (INIS)
Shin, Y.W.; Wiedermann, A.H.
1984-02-01
A method was published, based on the integral method of characteristics, by which the junction and boundary conditions needed in computation of a flow in a piping network can be accurately formulated. The method for the junction and boundary conditions formulation together with the two-step Lax-Wendroff scheme are used in a computer program; the program in turn, is used here in calculating sample problems related to the blowdown transient of a two-phase flow in the piping network downstream of a PWR pressurizer. Independent, nearly exact analytical solutions also are obtained for the sample problems. Comparison of the results obtained by the hybrid numerical technique with the analytical solutions showed generally good agreement. The good numerical accuracy shown by the results of our scheme suggest that the hybrid numerical technique is suitable for both benchmark and design calculations of PWR pressurizer blowdown transients
Radiation transmission pipe thickness measurement system
International Nuclear Information System (INIS)
Higashi, Yasuhiko
2010-01-01
Fuji Electric Systems can be measured from the outer insulation of the transmission Characteristics and radiation detection equipment had been developed that can measure pipe wall thinning in plant and running, the recruitment of another three-beam calculation method by pipe thickness measurement system was developed to measure the thickness of the pipe side. This equipment has been possible to measure the thickness of the circumferential profile of the pipe attachment by adopting automatic rotation. (author)
International Nuclear Information System (INIS)
Dupas, P.; Le Delliou, P.
1997-01-01
We calculate with a finite element program the residual stresses generated by quenching and grinding a cast duplex stainless steel pipe. These calculations are performed with Code Aster (developed by EDF/R and D D). They are preliminary to a 3D study concerning an elbow made of the same material. Quenching is simulated by an axisymmetric thermomechanical calculation. Grinding are simulated either by lowering mechanical properties in ground parts of the pipe, either by the releasing the nodes. Stresses due to quenching are in high compression in the skin and tensile in the middle. After grinding (the first concerning both internal and external skins, the second concerning only the internal skin), stresses become tensile on the skin. These results are compared to those obtained in a similar study by CEA and also to the measurement. Some important differences appear in the thermal results between the two FE programs, due to a too coarse time step in the CASTEM 2000 calculation. However, the effect on the residual stress field is not very important. Two complementary studies have shown a negligible influence of mesh size, as well as an equivalence of the two numerical methods used for simulating grinding (lowering the Young modulus and releasing the nodes), according the values given at the notes of the skin by the first method are corrected. (authors)
[Evaluation of tidal volume delivered by ventilators during volume-controlled ventilation].
Zhou, Juan; Yan, Yong; Cao, Desen
2014-12-01
To study the ways which ensure the delivery of enough tidal volume to patients under various conditions close to the demand of the physician. The volume control ventilation model was chosen, and the simulation lung type was active servo lung ASL 5000 or Michigan lung 1601. The air resistance, air compliance and lung type in simulation lungs were set. The tidal volume was obtained from flow analyzer PF 300. At the same tidal volume, the displaying values of tidal volume of E5, Servo i, Evital 4, and Evital XL ventilators with different lung types of patient, compliance of gas piping, leakage, gas types, etc. were evaluated. With the same setting tidal volume of a same ventilator, the tidal volume delivered to patients was different with different lung types of patient, compliance of gas piping, leakage, gas types, etc. Reducing compliance and increasing resistance of the patient lungs caused high peak airway pressure, the tidal volume was lost in gas piping, and the tidal volume be delivered to the patient lungs was decreased. If the ventilator did not compensate to leakage, the tidal volume delivered to the patient lungs was decreased. When the setting gas type of ventilator did not coincide with that applying to the patient, the tidal volume be delivered to the patient lungs might be different with the setting tidal volume of ventilator. To ensure the delivery of enough tidal volume to patients close to the demand of the physician, containable factors such as the compliance of gas piping, leakage, and gas types should be controlled.
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
Mathematical model of polyethylene pipe bending stress state
Serebrennikov, Anatoly; Serebrennikov, Daniil
2018-03-01
Introduction of new machines and new technologies of polyethylene pipeline installation is usually based on the polyethylene pipe flexibility. It is necessary that existing bending stresses do not lead to an irreversible polyethylene pipe deformation and to violation of its strength characteristics. Derivation of the mathematical model which allows calculating analytically the bending stress level of polyethylene pipes with consideration of nonlinear characteristics is presented below. All analytical calculations made with the mathematical model are experimentally proved and confirmed.
Alpha contamination assessment for D ampersand D activities: Monitoring pipe interiors
International Nuclear Information System (INIS)
Rawool-Sullivan, M.W.; Conaway, J.G.; MacArthur, D.W.; Vaccarella, J.
1996-02-01
We have developed a prototype instrument capable of assessing alpha-emitting contamination on interior surfaces of ducts, pipes, tanks, and other enclosed volumes without inserting a probe. Air is drawn through the potentially contaminated volume and then through a detection grid, where ions created in the air by alpha particles are collected and the resulting charge measured with a sensitive electrometer. A filter at the intake end of the contaminated volume excludes externally created ions, so only ions generated inside the volume are detected. We have studied the response of this prototype in initial experiments using calibrated alpha sources with various pipe diameters and configurations, air flows, and source locations in the pipes. The results of these experiments indicate that this method can be an effective approach to assessing internal contamination
Design of a micro-irrigation system based on the control volume method
Directory of Open Access Journals (Sweden)
Chasseriaux G.
2006-01-01
Full Text Available A micro-irrigation system design based on control volume method using the back step procedure is presented in this study. The proposed numerical method is simple and consists of delimiting an elementary volume of the lateral equipped with an emitter, called « control volume » on which the conservation equations of the fl uid hydrodynamicʼs are applied. Control volume method is an iterative method to calculate velocity and pressure step by step throughout the micro-irrigation network based on an assumed pressure at the end of the line. A simple microcomputer program was used for the calculation and the convergence was very fast. When the average water requirement of plants was estimated, it is easy to choose the sum of the average emitter discharge as the total average fl ow rate of the network. The design consists of exploring an economical and effi cient network to deliver uniformly the input fl ow rate for all emitters. This program permitted the design of a large complex network of thousands of emitters very quickly. Three subroutine programs calculate velocity and pressure at a lateral pipe and submain pipe. The control volume method has already been tested for lateral design, the results from which were validated by other methods as fi nite element method, so it permits to determine the optimal design for such micro-irrigation network
Thermal-hydraulic analysis of the improved TOPAZ-II power system using a heat pipe radiator
Energy Technology Data Exchange (ETDEWEB)
Zhang, Wenwen; Zhang, Dalin, E-mail: dlzhang@mail.xjtu.edu.cn; Tian, Wenxi; Qiu, Suizheng; Su, G.H.
2016-10-15
Highlights: • The system thermal-hydraulic model of the improved space thermionic reactor is developed. • The temperature reactivity feedback effects of the moderator, UO2 fuel, electrodes and reflector are considered. • The alkali metal heat pipe radiator is modeled with the two dimensional heat pipe model. • The steady state and the start-up procedure of the system are analyzed. - Abstract: A system analysis code coupled with the heat pipe model is developed to analyze the thermal-hydraulic characteristics of the improved TOPAZ-II reactor power system with a heat pipe radiator. The core thermal-hydraulic model, neutron physics model, and the coolant loop component models (including pump, volume accumulator, pipes and plenums) are established. The designed heat pipe radiator, which replaces the original pumped loop radiator, is also modeled, including two-dimensional heat pipe analysis model, fin model and coolant transport duct model. The system analysis code and the heat pipe model is coupled in the transport duct model. Steady state condition and start-up procedure of the improved TOPAZ-II system are calculated. The results show that the designed radiator can satisfy the waste heat rejection requirement of the improved power system. Meanwhile, the code can be used to obtained the thermal characteristics of the system transients such as the start-up process.
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.))
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
BED-Volume histograms calculation for routine clinical dosimetry in brachytherapy
International Nuclear Information System (INIS)
Galelli, M.; Feroldi, P.
1995-01-01
The consideration of volumes is essential in Brachytherapy clinical dosimetry (I.C.R.U). Indeed, several indices, all based on dose-volume histograms (DVHs), have been designed in order to evaluate: before the therapy the volumetric quality of different possible implant geometries; during the therapy the consistency of the real and the previsional implants. Radiobiological evaluations, considering the dose deposition temporal pattern of treatment, can be usefully added to dosimetric calculations, to compare different treatment schedules. The Linear-Quadratic model is the most used: radiobiological modelisation and Biologically Effective Dose (BED) is principal related dosimetric quantity. Therefore, the consideration of BED-volume histogram (BED-VHs) is a straightforward extension of DVHs. In practice, BED-VHs can help relative comparisons and optimisations in treatment planning when combined to dose-volume histograms. Since 1994 the dosimetric calculations for all the gynecological brachytherapy treatments are performed considering also DVHs and BED-VHs. In this presentation we show the methods of BEDVHs calculation, together with some typical results
Neutron Streaming in D{sub 2}O Pipes
Energy Technology Data Exchange (ETDEWEB)
Braun, J; Randen, K
1962-07-01
An investigation has been carried out concerning the attenuation of neutrons inside D{sub 2}O-filled pipes penetrating a concrete shield. As the purpose has been to simulate the conditions around a heavy water power reactor, pipes surrounded by an annular air gap have also been considered. Thermal, epithermal and fast neutron fluxes have been measured in three separate pipes (15, 22 and 28 cm in diameter and 100 cm long) with annulii ranging from 0 to 9.7 cm in width. The thermal flux distribution has been predicted theoretically by assuming it to be composed of three components originating from a fast exponential volume source and two surface sources at the origin; a 1/E-distributed source and a thermal source, of which the latter proved to be negligible. The fast flux distribution has been approximated by a single exponential expression for the configuration with no annulus and with the sum of two exponentials when an annulus is present. The agreement between measured and calculated values for the thermal flux is better than a factor 1.5 after about 10 cm. For deep penetration (>40 cm) the agreement is within 20 % and only the fast volume source contributes appreciably to the thermal flux. This holds for all cases both with and without annulus. The agreement in the case of the fast flux is within a factor 1.3 (>40 cm) for no annulus geometry and about 2-3 with annulus. Curves are presented for obtaining necessary parameters ('removal' cross section and extrapolated radius) for other geometries than those covered in this report.
Computer simulations of magnetic fluids in laminar pipe flows
International Nuclear Information System (INIS)
Ramos, D.M.; Cunha, F.R.; Sobral, Y.D.; Fontoura Rodrigues, J.L.A.
2005-01-01
Finite volume method is adapted to simulate momentum and magnetic coupled equations of a laminar magnetic fluid flow. An evolution equation is used to calculate the fluid magnetization. Pressure-driven flow under steady and oscillatory magnetic field is investigated. The magnetostatic limit of the Maxwell's equations is treated in terms of a Poisson equation numerically integrated. The SIMPLE algorithm is used to calculate the pressure-velocity coupling when the pressure field is not prescribed. Suitable boundary conditions for velocity, magnetization and field intensity on the pipe wall are described. Results are obtained for velocity and pressure response under several conditions of the identified physical parameters of the flow. The simulations are verified by comparing numerical results and asymptotic theory, and they show a very good agreement
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.
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.
Heat transfer capability analysis of heat pipe for space reactor
International Nuclear Information System (INIS)
Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang
2015-01-01
To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)
Modelling of fiberglass pipe destruction process
Directory of Open Access Journals (Sweden)
А. К. Николаев
2017-03-01
Full Text Available The article deals with important current issue of oil and gas industry of using tubes made of high-strength composite corrosion resistant materials. In order to improve operational safety of industrial pipes it is feasible to use composite fiberglass tubes. More than half of the accidents at oil and gas sites happen at oil gathering systems due to high corrosiveness of pumped fluid. To reduce number of accidents and improve environmental protection we need to solve the issue of industrial pipes durability. This problem could be solved by using composite materials from fiberglass, which have required physical and mechanical properties for oil pipes. The durability and strength can be monitored by a fiberglass winding method, number of layers in composite material and high corrosion-resistance properties of fiberglass. Usage of high-strength composite materials in oil production is economically feasible; fiberglass pipes production is cheaper than steel pipes. Fiberglass has small volume weight, which simplifies pipe transportation and installation. In order to identify the efficiency of using high-strength composite materials at oil production sites we conducted a research of their physical-mechanical properties and modelled fiber pipe destruction process.
Directory of Open Access Journals (Sweden)
Adil Abbas AL-Moosawy
2016-09-01
Full Text Available Experimental study of γ /Al2O3 with mean diameter of less than 50 nm was dispersed in the distilled water that flows through a pipe consist of five sections as work station ,four sections made of carbon steel metal and one sections made of Pyrex glass pipe, with five nanoparticles volume concentrations of 0%,0.1%,0.2%,0.3%,and 0.4% with seven different volume flow rates 100, 200 , 300, 400, 500, 600 ,and 700ℓ/min were investigated to calculated pressure distribution for the cases without rubber ,with 3mm rubber and with 6mm rubber used to support the pipe. Reynolds number was between 20000 and 130000. Frequency value through pipe was measured for all stations of pipe for all cases. The results show that the pressure drop and wall shear stress of the nanofluid increase by increasing the nanoparticles volume concentrations or Reynolds number, the values of frequency through the pipe increase continuously when wall shear stress increases and the ratio of increment increases as nanofluid concentrations increase. Increasing of vibration frequency lead to increasing the friction factor between the pipe and the wall and thus increasing in pressure drop. Several equations between the wall shear stress and frequency for all volume concentration and for three cases without rubber, with rubber has 3mm thickness ,and with rubber has 6mm thickness. Finally, the results led to that γ /Al2O3 could function as a good and alternative conventional working fluid in heat transfer applications. A good agreement is seen between the experimental and those available in the literature
Seismic design evaluation guidelines for buried piping for the DOE HLW Facilities
International Nuclear Information System (INIS)
Lin, Chi-Wen; Antaki, G.; Bandyopadhyay, K.; Bush, S.H.; Costantino, C.; Kennedy, R.
1995-01-01
This paper presents the seismic design and evaluation guidelines for underground piping for the Department of Energy (DOE) High-Level-Waste (HLW) Facilities. The underground piping includes both single and double containment steel pipes and concrete pipes with steel lining, with particular emphasis on the double containment piping. The design and evaluation guidelines presented in this paper follow the generally accepted beam-on-elastic-foundation analysis principle and the inertial response calculation method, respectively, for piping directly in contact with the soil or contained in a jacket. A standard analysis procedure is described along with the discussion of factors deemed to be significant for the design of the underground piping. The following key considerations are addressed: the design feature and safety requirements for the inner (core) pipe and the outer pipe; the effect of soil strain and wave passage; assimilation of the necessary seismic and soil data; inertial response calculation for the inner pipe; determination of support anchor movement loads; combination of design loads; and code comparison. Specifications and justifications of the key parameters used, stress components to be calculated and the allowable stress and strain limits for code evaluation are presented
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.)
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)
Fatigue crack growth in austenitic stainless steel piping
International Nuclear Information System (INIS)
Bethmont, M.; Cheissoux, J.L.; Lebey, J.
1981-04-01
The study presented in this paper is being carried out with a view to substantiating the calculations of the fatigue crack growth in pipes made of 316 L stainless steel. The results obtained may be applied to P.W.R. primary piping. It is divided into two parts. First, fatigue tests (cyclic pressure) are carried out under hot and cold conditions with straight pipes machined with notches of various dimensions. The crack propagation and the fatigue crack growth rate are measured here. Second, calculations are made in order to interpret experimental results. From elastic calculations the stress intensity factor is assessed to predict the crack growth rate. The results obtained until now and presented in this paper relate to longitudinal notches
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)
Internal ultrasonic inspection of flexible pipe
Energy Technology Data Exchange (ETDEWEB)
Baltzersen, O. (IKU Petroleumsforskning A/S, Trondheim (Norway) Norwegian Inst. of Tech., Trondheim (Norway). Div. of Petroleum Engineering and Applied Geophysics); Waag, T.I. (IKU Petroleumsforskning A/S, Trondheim (Norway))
1993-10-01
Methods for internal ultrasonic inspection of flexible pipe have been investigated through experiments with a short sample of Coflexip pipe. Ultrasonic backscatter methods using normal and non-normal incidence have been used for qualitative high contrast ultrasonic imaging of the inner surface of the pipe. Analysis of the internal cross-section has been performed based on the use of a non-contact ultrasonic caliper, and processing procedures which enable calculation of, and compensation for, eccentricity of the tool in the pipe. The methods developed can be used to quantitatively estimate the thickness of the internal carcass, and perform high resolution topographic mapping of the inner surface. (Author)
Mahdavi, Mahboobe; Tiari, Saeed; Qiu, Songgang
2016-11-01
Heat pipes are two-phase heat transfer devices, which operate based on evaporation and condensation of a working fluid inside a sealed container. In the current work, an experimental study was conducted to investigate the performance of a copper-water heat pipe. The performance was evaluated by calculating the corresponding thermal resistance as the ratio of temperature difference between evaporator and condenser to heat input. The effects of inclination angle and the amount of working fluid were studied on the equivalent thermal resistance. The results showed that if the heat pipe is under-filled with the working fluid, energy transferring capacity of the heat pipe decreases dramatically. However, overfilling heat pipe causes over flood and degrades heat pipe performance. The minimum thermal resistances were obtained for the case that 30% of the heat pipe volume was filled with working fluid. It was also found that in gravity-assisted orientations, the inclination angle does not have significant effect on the performance of the heat pipe. However, for gravity-opposed orientations, as the inclination angle increases, the temperature difference between the evaporator and condensation increases and higher thermal resistances are obtained. Authors appreciate the financial support by a research Grant from Temple University.
Efficient methods of piping cleaning
Directory of Open Access Journals (Sweden)
Orlov Vladimir Aleksandrovich
2014-01-01
Full Text Available The article contains the analysis of the efficient methods of piping cleaning of water supply and sanitation systems. Special attention is paid to the ice cleaning method, in course of which biological foil and various mineral and organic deposits are removed due to the ice crust buildup on the inner surface of water supply and drainage pipes. These impurities are responsible for the deterioration of the organoleptic properties of the transported drinking water or narrowing cross-section of drainage pipes. The co-authors emphasize that the use of ice compared to other methods of pipe cleaning has a number of advantages due to the relative simplicity and cheapness of the process, economical efficiency and lack of environmental risk. The equipment for performing ice cleaning is presented, its technological options, terms of cleansing operations, as well as the volumes of disposed pollution per unit length of the water supply and drainage pipelines. It is noted that ice cleaning requires careful planning in the process of cooking ice and in the process of its supply in the pipe. There are specific requirements to its quality. In particular, when you clean drinking water system the ice applied should be hygienically clean and meet sanitary requirements.In pilot projects, in particular, quantitative and qualitative analysis of sediments adsorbed by ice is conducted, as well as temperature and the duration of the process. The degree of pollution of the pipeline was estimated by the volume of the remote sediment on 1 km of pipeline. Cleaning pipelines using ice can be considered one of the methods of trenchless technologies, being a significant alternative to traditional methods of cleaning the pipes. The method can be applied in urban pipeline systems of drinking water supply for the diameters of 100—600 mm, and also to diversion collectors. In the world today 450 km of pipelines are subject to ice cleaning method.Ice cleaning method is simple
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
International Nuclear Information System (INIS)
Papes, Iva; Degroote, Joris; Vierendeels, Jan
2016-01-01
Highlights: • A multi-chamber model is developed from the mass and energy conservation laws. • To better predict inlet pipe pulsations a 3D inlet pipe model is coupled to it. • Flow coefficients are derived from 3D CFD calculations. • Maximal deviation between the full CFD and the presented model is around 5%. • This model is a good compromise between accuracy and computational resources. - Abstract: A twin screw expander is a positive displacement machine used in various applications of waste heat recovery. The performance of this machine is influenced by internal leakages, gas pulsations formed in the inlet pipe and the properties of the refrigerant. In this paper a multi-chamber mathematical model of a twin screw expander is presented to predict its performance. From the mass and energy conservation laws, differential equations are derived which are then solved together with the appropriate Equation of State (EoS) in the instantaneous control volumes. In order to calculate the mass flow rates through leakage paths more accurately, flow coefficients used in the converging nozzle model were derived from 3D Computational Fluid Dynamic (CFD) calculation. Due to high gas pulsation levels at the inlet port, a coupling with a 3D CFD inlet pipe model is introduced in order to better predict throttling losses. The maximal deviation between predictions by the developed model and 3D CFD calculations of the complete machine is around 5% for the mass flow rate and the power output.
Diameter structure modeling and the calculation of plantation volume of black poplar clones
Directory of Open Access Journals (Sweden)
Andrašev Siniša
2004-01-01
Full Text Available A method of diameter structure modeling was applied in the calculation of plantation (stand volume of two black poplar clones in the section Aigeiros (Duby: 618 (Lux and S1-8. Diameter structure modeling by Weibull function makes it possible to calculate the plantation volume by volume line. Based on the comparison of the proposed method with the existing methods, the obtained error of plantation volume was less than 2%. Diameter structure modeling and the calculation of plantation volume by diameter structure model, by the regularity of diameter distribution, enables a better analysis of the production level and assortment structure and it can be used in the construction of yield and increment tables.
Calculating regional tissue volume for hyperthermic isolated limb perfusion: Four methods compared.
Cecchin, D; Negri, A; Frigo, A C; Bui, F; Zucchetta, P; Bodanza, V; Gregianin, M; Campana, L G; Rossi, C R; Rastrelli, M
2016-12-01
Hyperthermic isolated limb perfusion (HILP) can be performed as an alternative to amputation for soft tissue sarcomas and melanomas of the extremities. Melphalan and tumor necrosis factor-alpha are used at a dosage that depends on the volume of the limb. Regional tissue volume is traditionally measured for the purposes of HILP using water displacement volumetry (WDV). Although this technique is considered the gold standard, it is time-consuming and complicated to implement, especially in obese and elderly patients. The aim of the present study was to compare the different methods described in the literature for calculating regional tissue volume in the HILP setting, and to validate an open source software. We reviewed the charts of 22 patients (11 males and 11 females) who had non-disseminated melanoma with in-transit metastases or sarcoma of the lower limb. We calculated the volume of the limb using four different methods: WDV, tape measurements and segmentation of computed tomography images using Osirix and Oncentra Masterplan softwares. The overall comparison provided a concordance correlation coefficient (CCC) of 0.92 for the calculations of whole limb volume. In particular, when Osirix was compared with Oncentra (validated for volume measures and used in radiotherapy), the concordance was near-perfect for the calculation of the whole limb volume (CCC = 0.99). With methods based on CT the user can choose a reliable plane for segmentation purposes. CT-based methods also provides the opportunity to separate the whole limb volume into defined tissue volumes (cortical bone, fat and water). Copyright © 2016 Elsevier Ltd. All rights reserved.
Reliability analysis of stiff versus flexible piping
International Nuclear Information System (INIS)
Lu, S.C.
1985-01-01
The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. The authors then investigated a couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design. They concluded that these changes substantially reduce calculated piping responses and allow piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements. Furthermore, the more flexible piping redesigns are capable of exhibiting reliability levels equal to or higher than the original stiffer design. An investigation of the malfunction of pipe whip restraints confirmed that the malfunction introduced higher thermal stresses and tended to reduce the overall piping reliability. Finally, support and component reliabilities were evaluated based on available fragility data. Results indicated that the support reliability usually exhibits a moderate decrease as the piping flexibility increases. Most on-line pumps and valves showed an insignificant reduction in reliability for a more flexible piping design
ICPP calcined solids storage facility closure study. Volume III: Engineering design files
International Nuclear Information System (INIS)
1998-02-01
The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project's scope of work. Should more accurate numbers be required, a new analysis would be necessary
ICPP calcined solids storage facility closure study. Volume III: Engineering design files
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-02-01
The following information was calculated to support cost estimates and radiation exposure calculations for closure activities at the Calcined Solids Storage Facility (CSSF). Within the estimate, volumes were calculated to determine the required amount of grout to be used during closure activities. The remaining calcine on the bin walls, supports, piping, and floor was also calculated to approximate the remaining residual calcine volumes at different stages of the removal process. The estimates for remaining calcine and vault void volume are higher than what would actually be experienced in the field, but are necessary for bounding purposes. The residual calcine in the bins may be higher than was is experienced in the field as it was assumed that the entire bin volume is full of calcine before removal activities commence. The vault void volumes are higher as the vault roof beam volumes were neglected. The estimations that follow should be considered rough order of magnitude, due to the time constraints as dictated by the project`s scope of work. Should more accurate numbers be required, a new analysis would be necessary.
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.
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.
Pipe Decontamination Involving String-Foam Circulation
International Nuclear Information System (INIS)
Turchet, J.P.; Estienne, G.; Fournel, B.
2002-01-01
Foam applications number for nuclear decontamination purposes has recently increased. The major advantage of foam decontamination is the reduction of secondary liquid wastes volumes. Among foam applications, we focus on foam circulation in contaminated equipment. Dynamic properties of the system ensures an homogeneous and rapid effect of the foam bed-drifted chemical reagents present in the liquid phase. This paper describes a new approach of foam decontamination for pipes. It is based on an alternated air and foam injections. We called it 'string-foam circulation'. A further reduction of liquid wastes is achieved compared to continuous foam. Secondly, total pressure loss along the pipe is controlled by the total foam length in the pipe. It is thus possible to clean longer pipes keeping the pressure under atmospheric pressure value. This ensures the non dispersion of contamination. This study describes experimental results obtained with a neutral foam as well with an acid foam on a 130 m long loop. Finally, the decontamination of a 44 meters pipe is presented. (authors)
Fatigue check of nuclear safety class 1 reactor coolant pipe
International Nuclear Information System (INIS)
Wang Qing; Fang Yonggang; Chu Qibao; Xu Yu; Li Hailong
2015-01-01
Fatigue and thermal ratcheting analyses of nuclear safety Class 1 reactor coolant pipe in a nuclear power plant were independently carried out in this paper. The software used for calculation is ROCOCO, which is based on RCC-M code. The difference of nuclear safety Class 1 pipe fatigue evaluation between RCC-M code and ASME code was compared. The main aspects of comparison include the calculation scoping of fatigue design, the calculation method of primary plus secondary stress intensity, the elastic-plastic correction coefficient calculation, and the dynamic load combination method etc. By correcting inconsistent algorithm of ASME code within ROCOCO, the fatigue usage factor and thermal ratcheting design margin of 65 mm and 55 mm wall thickness of the pipe were obtained. The results show that the minimum wall thickness of the pipe must exceed 55 mm and the design value of the thermal ratcheting of 55 mm wall thickness reaches 95% of the allowable value. (authors)
Study on Monitoring Rock Burst through Drill Pipe Torque
Zhonghua Li; Liyuan Zhu; Wanlei Yin; Yanfang Song
2015-01-01
This paper presents a new method to identify the danger of rock burst from the response of drill pipe torque during drilling process to overcome many defects of the conventional volume of drilled coal rubble method. It is based on the relationship of rock burst with coal stress and coal strength. Through theoretic analysis, the change mechanism of drill pipe torque and the relationship of drill pipe torque with coal stress, coal strength, and drilling speed are investigated. In light of the a...
Technical considerations for flexible piping design in nuclear power plants
International Nuclear Information System (INIS)
Lu, S.C.; Chou, C.K.
1985-01-01
The overall objective of this research project is to develop a technical basis for flexible piping designs which will improve piping reliability and minimize the use of pipe supports, snubbers, and pipe whip restraints. The current study was conducted to establish the necessary groundwork based on the piping reliability analysis. A confirmatory piping reliability assessment indicated that removing rigid supports and snubbers tends to either improve or affect very little the piping reliability. A couple of changes to be implemented in Regulatory Guide (RG) 1.61 and RG 1.122 aimed at more flexible piping design were investigated. It was concluded that these changes substantially reduce calculated piping responses and allows piping redesigns with significant reduction in number of supports and snubbers without violating ASME code requirements
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)
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
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
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
Bandriyana, B.; Utaja
2010-06-01
Thermal stratification introduces thermal shock effect which results in local stress and fatique problems that must be considered in the design of nuclear power plant components. Local stress and fatique calculation were performed on the Pressurize Surge Line piping system of the Pressurize Water Reactor of the Nuclear Power Plant. Analysis was done on the operating temperature between 177 to 343° C and the operating pressure of 16 MPa (160 Bar). The stagnant and transient condition with two kinds of stratification model has been evaluated by the two dimensional finite elements method using the ANSYS program. Evaluation of fatigue resistance is developed based on the maximum local stress using the ASME standard Code formula. Maximum stress of 427 MPa occurred at the upper side of the top half of hot fluid pipe stratification model in the transient case condition. The evaluation of the fatigue resistance is performed on 500 operating cycles in the life time of 40 years and giving the usage value of 0,64 which met to the design requirement for class 1 of nuclear component. The out surge transient were the most significant case in the localized effects due to thermal stratification.
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)
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.
International Nuclear Information System (INIS)
Kaufman, A.M.
1978-01-01
A rapid method is presented for calculating transport in a network of one-dimensional flow paths or ''pipes''. The method defines a Green's function for each flow path and prescribes a method of combining these Green's functions to produce an overall Green's function for the flow path network. A unique feature of the method is the use of the Laplace transform of these Green's functions to carry out most of the calculations
Heat pipe with PCM for electronic cooling
International Nuclear Information System (INIS)
Weng, Ying-Che; Cho, Hung-Pin; Chang, Chih-Chung; Chen, Sih-Li
2011-01-01
This article experimentally investigates the thermal performances of a heat pipe with phase change material for electronic cooling. The adiabatic section of heat pipe is covered by a storage container with phase change material (PCM), which can store and release thermal energy depending upon the heating powers of evaporator and fan speeds of condenser. Experimental investigations are conducted to obtain the system temperature distributions from the charge, discharge and simultaneous charge/discharge performance tests. The parameters in this study include three kinds of PCMs, different filling PCM volumes, fan speeds, and heating powers in the PCM cooling module. The cooling module with tricosane as PCM can save 46% of the fan power consumption compared with the traditional heat pipe.
International Nuclear Information System (INIS)
Hu Lin; Cui Wei; Shi Hanwen; Tian Yingping; Wang Weigang; Feng Yanguang; Huang Xueyan; Liu Zhisheng
2003-01-01
Objective: To compare the relative accuracy of three methods measuring left ventricular volume by X-ray ventriculography: single plane area-length method, biplane area-length method, and single-plane Simpson's method. Methods: Left ventricular casts were obtained within 24 hours after death from 12 persons who died from non-cardiac causes. The true left ventricular cast volume was measured by water displacement. The calculated volume of the casts was obtained with 3 angiographic methods, i.e., single-plane area-length method, biplane area-length method, and single-plane Simpson's method. Results: The actual average volume of left ventricular casts was (61.17±26.49) ml. The left ventricular volume was averagely (97.50±35.56) ml with single plane area-length method, (90.51±36.33) ml with biplane area-length method, and (65.00± 23.63) ml with single-plane Simpson's method. The left ventricular volumes calculated with single-plane and biplane area-length method were significantly larger than that the actual volumes (P 0.05). The left ventricular volumes calculated with single-plane and biplane area-length method were significantly larger than those calculated with single-plane Simpson's method (P 0.05). The over-estimation of left ventricular volume by single plane area-length method (36.34±17.98) ml and biplane area-length method (29.34±15.59) ml was more obvious than that calculated by single-plane Simpson's method (3.83±8.48) ml. Linear regression analysis showed that there was close correlations between left ventricular volumes calculated with single plane area-length method, biplane area-length method, Simpson's method and the true volume (all r>0.98). Conclusion: Single-plane Simpson's method is more accurate than single plane area-length method and biplane area-length method for left ventricular volume measurement; however, both the single-plane and biplane area-length methods could be used in clinical practice, especially in those imaging modality
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
A thermal study of pipes with outer transverse fins
Directory of Open Access Journals (Sweden)
S. Gil
2016-10-01
Full Text Available This paper provides results of thermal investigations on pipes with outer transverse fins produced by placing a strip, being a form of helical spring which functions as a radiator, on the basis pipe. The investigations were carried out at the facility that enables measurements with respect to both natural and forced convection. Performance of the investigated pipes was assessed in relation to a non-finned pipe and a pipe welded with the use of Metal Active Gas (MAG technology. The experiments have shown that the finned pipe welding technology does not markedly affect their thermal efficiency, which has been confirmed by performed model calculations, while the welding technology has a crucial impact on their operating performance.
Study on pipe deflection by using numerical method
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.
Crack-opening area calculations for circumferential through-wall pipe cracks
Energy Technology Data Exchange (ETDEWEB)
Kishida, K.; Zahoor, A.
1988-08-01
This report describes the estimation schemes for crack opening displacement (COD) of a circumferential through-wall crack, then compares the COD predictions with pipe experimental data. Accurate predictions for COD are required to reliably predict the leak rate through a crack in leak-before-break applications.
Crack-opening area calculations for circumferential through-wall pipe cracks
International Nuclear Information System (INIS)
Kishida, K.; Zahoor, A.
1988-08-01
This report describes the estimation schemes for crack opening displacement (COD) of a circumferential through-wall crack, then compares the COD predictions with pipe experimental data. Accurate predictions for COD are required to reliably predict the leak rate through a crack in leak-before-break applications
Study on Monitoring Rock Burst through Drill Pipe Torque
Directory of Open Access Journals (Sweden)
Zhonghua Li
2015-01-01
Full Text Available This paper presents a new method to identify the danger of rock burst from the response of drill pipe torque during drilling process to overcome many defects of the conventional volume of drilled coal rubble method. It is based on the relationship of rock burst with coal stress and coal strength. Through theoretic analysis, the change mechanism of drill pipe torque and the relationship of drill pipe torque with coal stress, coal strength, and drilling speed are investigated. In light of the analysis, a new device for testing drill pipe torque is developed and a series of experiments is performed under different conditions; the results show that drill pipe torque linearly increases with the increase of coal stress and coal strength; the faster the drilling speed, the larger the drill pipe torque, and vice versa. When monitoring rock burst by drill pipe torque method, the index of rock burst is regarded as a function in which coal stress index and coal strength index are principal variables. The results are important for the forecast of rock burst in coal mine.
Heat pipe technology. a bibliography with abstracts. Quarterly update, 31 March 1975
International Nuclear Information System (INIS)
1975-01-01
Heat Pipe Technology is a continuing bibliographic summary of research on the subject of the heat pipe. The first volume was published in 1971. The 1972, 1973, and 1974 Annual Supplements have been published and distributed. This update cites additional references for 1975
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
factors to this may be the smaller dry well volume per blowdown pipe ratio and the lack of dry well internal structures in the PPOOLEX facility. Furthermore, the pipe material seemed to have an effect on the condensation process inside the pipe. Polycarbonate has two orders of magnitude smaller thermal conductivity than steel. (Author)
Resolution of concerns in auxiliary feedwater piping
International Nuclear Information System (INIS)
Bain, R.A.; Testa, M.F.
1994-01-01
Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures
International Nuclear Information System (INIS)
Jo, Jong Chull; Kim, Wee Kyung; Kim, Yun Il; Cho, Sang Jin; Choi, Seok Ki
2000-01-01
A detailed numerical analysis of initial evolution of thermal stratification in a curved pipe with a finite wall thickness is performed. A primary emphasis of the present study is placed on the investigation of the effect of existence of pipe wall thickness on the evolution of thermal stratification. A simple and convenient numerical method of treating the unsteady conjugate heat transfer in Cartesian as well as non-orthogonal coordinate systems is presented. The proposed unsteady conjugate heat transfer analysis method is implemented in a finite volume thermal-hydraulic computer code based on a cell-centered, non-staggered grid arrangement, the SIMPLEC algorithm and a higher-order bounded convection scheme. Calculations are performed for initial evolution of thermal stratification with high Richardson number in a curved pipe. The predicted results show that the thermally stratified flow and transient conjugate heat transfer in a curved pipe with a specified wall thickness can be satisfactorily analyzed by using the numerical method presented in this paper. As the result, the present analysis method is considered to be effective for the determination of transient temperature distributions in the wall of curved piping system subjected to internally thermal stratification. In addition, the method can be extended to be applicable for the simulation of turbulent flow of thermally stratified fluid
International Nuclear Information System (INIS)
Poirier, M; Fernando Fondeur, F; Samuel Fink, S
2006-01-01
The Department of Energy (DOE) identified the caustic side solvent extraction (CSSX) process as the preferred technology to remove cesium from radioactive waste solutions at the Savannah River Site (SRS). As a result, Washington Savannah River Company (WSRC) began designing and building a Modular CSSX Unit (MCU) in the SRS tank farm to process liquid waste for an interim period until the Salt Waste Processing Facility (SWPF) begins operations. Both the solvent and the strip effluent streams could contain high concentrations of cesium which must be removed from the contactors, process tanks, and piping prior to performing contactor maintenance. When these vessels are drained, thin films or drops will remain on the equipment walls. Following draining, the vessels will be flushed with water and drained to remove the flush water. The draining reduces the cesium concentration in the vessels by reducing the volume of cesium-containing material. The flushing, and subsequent draining, reduces the cesium in the vessels by diluting the cesium that remains in the film or drops on the vessel walls. MCU personnel requested that Savannah River National Laboratory (SRNL) researchers conduct a literature search to identify models to calculate the thickness of the liquid films remaining in the contactors, process tanks, and piping following draining of salt solution, solvent, and strip solution. The conclusions from this work are: (1) The predicted film thickness of the strip effluent is 0.010 mm on vertical walls, 0.57 mm on horizontal walls and 0.081 mm in horizontal pipes. (2) The predicted film thickness of the salt solution is 0.015 mm on vertical walls, 0.74 mm on horizontal walls, and 0.106 mm in horizontal pipes. (3) The predicted film thickness of the solvent is 0.022 mm on vertical walls, 0.91 mm on horizontal walls, and 0.13 mm in horizontal pipes. (4) The calculated film volume following draining is: (a) Salt solution receipt tank--1.6 gallons; (b) Salt solution feed
International Nuclear Information System (INIS)
Trevin, Stephane; Moutrille, Marie-Pierre; Qiu, Gonghao; Miller, Cecile; Mellin, Nicolas
2012-09-01
EDF has developed during these 15 last years a software called BRT-CICERO TM for the surveillance of the secondary piping system of its Pressurized Water Reactors (PWRs). This software enables the operator to calculate the FAC wear rates taking into account all the influencing parameters such as: pipe isometrics, chromium content of the steel, chemical conditioning and operating parameters of the secondary circuit (temperature, pressure, etc.). This is a major tool for the operators to organize the maintenance and to plan the inspections. In the framework of the French pressure vessel law issued on March 15, 2000, the software BRT-CICERO TM has been recognized by the French authority for the FAC surveillance on the secondary pressure piping lines of the EDF 58 NPPs. It takes advantage of the experience feedback of EDF's fleet, of the R and D improvements (especially from the laboratory tests conducted on EDF's CIROCO loop) and is frequently updated. Kinetics calculations made with BRT-CICERO TM are highly dependent of chromium, copper and molybdenum contents of steel. These values are measured on site by X-ray portable fluorescence. EDF elaborated a measurement procedure with a validation process and verification of the measurement devices using certified blocks standard. This procedure enables EDF and service provider companies to measure more than 6 thousand components per year. These values are input in BRT-CICERO TM and the flow accelerated corrosion kinetic is calculated with a higher accuracy than before alloy contents measurement. The next version of BRT-CICERO will take into account chromium, copper and molybdenum contents. The actual version is using only chromium contents. This paper describes the X-Ray fluorescence and the procedure used at EDF. The advantage and drawbacks of this technique are discussed. According to research and development studies, the future algorithm for FAC calculation with these 3 alloys contents is described. Because of
Development of Calculation Algorithm for ECCS Kinematic Shock
Energy Technology Data Exchange (ETDEWEB)
Lee, Seung-Chan; Yoon, Duk-Joo; Ha, Sang-Jun [KHNP-CRI, Daejeon (Korea, Republic of)
2014-10-15
The void fraction of inverted U-pipes in front of SI(Safety Injection) pumps impact on the pipe system of ECCS(Emergency Core Cooling Systems). This phenomena is called as 'Kinematic Shock'. The purpose of this paper is to achieve the more exactly calculation when the kinematic shock is calculated by simplified equation. The behavior of the void packet of the ECCS pipes is illustrated by the simplified (other name is kinematic shock equation).. The kinematic shock is defined as the depth of total length of void clusters in the pipes of ECCS when the void cluster is continually reached along the part of pipes in vertical direction. In this paper, the simplified equation is evaluated by comparing calculation error each other.]. The more exact methods of calculating the depth of the kinematic shock in ECCS is achieved. The error of kinematic shock calculation is strongly depended on the calculation search gap and the order of Taylor's expansion. From this study, to select the suitable search gap and the suitable calculation order, differential root method, secant method, and Taylor's expansion form are compared one another.
Rupture hardware minimization in pressurized water reactor piping
International Nuclear Information System (INIS)
Mukherjee, S.K.; Ski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.F.; Quinones, D.F.; Server, W.L.
1989-01-01
For much of the high-energy piping in light reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but also improves the overall safety and integrity of the plant since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied a Beaver Valley Power Station- Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferrutic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in. (152-mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel line as small as 3-in. (76-mm) diameter (outside containment) can qualify for pipe rupture hardware elemination
A Corrosion Risk Assessment Model for Underground Piping
Datta, Koushik; Fraser, Douglas R.
2009-01-01
The Pressure Systems Manager at NASA Ames Research Center (ARC) has embarked on a project to collect data and develop risk assessment models to support risk-informed decision making regarding future inspections of underground pipes at ARC. This paper shows progress in one area of this project - a corrosion risk assessment model for the underground high-pressure air distribution piping system at ARC. It consists of a Corrosion Model of pipe-segments, a Pipe Wrap Protection Model; and a Pipe Stress Model for a pipe segment. A Monte Carlo simulation of the combined models provides a distribution of the failure probabilities. Sensitivity study results show that the model uncertainty, or lack of knowledge, is the dominant contributor to the calculated unreliability of the underground piping system. As a result, the Pressure Systems Manager may consider investing resources specifically focused on reducing these uncertainties. Future work includes completing the data collection effort for the existing ground based pressure systems and applying the risk models to risk-based inspection strategies of the underground pipes at ARC.
Fast Near-Field Calculation for Volume Integral Equations for Layered Media
DEFF Research Database (Denmark)
Kim, Oleksiy S.; Meincke, Peter; Breinbjerg, Olav
2005-01-01
. Afterwards, the scattered electric field can be easily computed at a regular rectangular grid on any horizontal plane us-ing a 2-dimensional FFT. This approach provides significant speedup in the near-field calculation in comparison to a straightforward numerical evaluation of the ra-diation integral since......An efficient technique based on the Fast Fourier Transform (FFT) for calculating near-field scattering by dielectric objects in layered media is presented. A higher or-der method of moments technique is employed to solve the volume integral equation for the unknown induced volume current density...
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
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.
Stresses in a curved pipe subject to an in-plane bending moment
International Nuclear Information System (INIS)
Hofmann, E.; Heeschen, U.
1979-01-01
The design of the KWU-primary component supports is mainly defined by the loads of the postulated pipe breaks. To estimate the maximum loading of a component support it is necessary to know the maximum in-plane bending moment (opening and closing) that can be transmitted by a pipe bend. Another reason for such information is that the displacements and distortions of the components cause higher stresses in elbows than in straight pipes. With a detailed knowledge of the deformation characteristic of a pipe bend an integrity analysis could be done without an expensive plastic system analysis. With this purpose in mind experiments were performed with straight pipes and pipe bends of different dimensions subject to in-plane bending moments. The experimental results give the ratio between the maximum transmittable moment of a pipe bend to that of a straight pipe or, the distortion of the end cross-sections and the flattening of the elbow cross-section. An attempt is made to derive simple expressions for estimating the behaviour at pipe elbows. Parallel to the experiments calculations were done for the straight pipe and elbow with a finite difference code with plastic capabilities. The results of the experiment and calculation are compared with the formulas of the ASME-Code section III subjection NB. (orig.)
A method for bubble volume calculating in vertical two-phase flow
International Nuclear Information System (INIS)
Wang, H Y; Dong, F
2009-01-01
The movement of bubble is a basic subject in gas-liquid two-phase flow research. A method for calculating bubble volume which is one of the most important characters in bubble motion research was proposed. A suit of visualized experimental device was designed and set up. Single bubble rising in stagnant liquid in a rectangular tank was studied using the high-speed video system. Bubbles generated by four orifice with different diameter (1mm, 2mm, 3mm, 4mm) were recorded respectively. Sequences of recorded high-speed images were processed by digital image processing method, such as image noise remove, binary image transform, bubble filling, and so on. then, Several parameters could be obtained from the processed image. Bubble area, equivalent diameter, bubble velocity, bubble acceleration are all indispensable in bubble volume calculating. In order to get the force balance equation, forces that work on bubble along vertical direction, including drag force, virtual mass force, buoyancy, gravity and liquid thrust, were analyzed. Finally, the bubble volume formula could be derived from the force balance equation and bubble parameters. Examples were given to shown the computing process and results. Comparison of the bubble volume calculated by geomettic method and the present method have shown the superiority of the proposed method in this paper.
Mechanical assessment of local thinned pipings
International Nuclear Information System (INIS)
Meister, E.
2007-01-01
Local wall thinning is likely to be found in some piping systems of nuclear power plant under, for example, Flow Accelerated Corrosion in raw water systems or by loss of metal during the grinding of the weld seam. To assess the mechanical integrity in such situations, EDF/SEPTEN has developed calculation methods for the RSE-M (In Service Inspection Rules for the Mechanical components of PWR nuclear power islands) code. This paper focuses on the methodology used for internal pressure resistance evaluation based on limit load calculations. Beyond the Nuclear Safety classification and requirements given by the RSE-M code, this problem is general for Power Piping and the associated in service rules. (author) [fr
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
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
Investigations on penetration control for automated pipe welding system
International Nuclear Information System (INIS)
Fujiki, Daisuke; Sato, Akihiro; Funamoto, Takao; Matsumoto, Toshimi; Kobayashi, Masahiro
1995-01-01
We have been investigating process conditions forming sound root bead by orbital welding technique for nuclear power stations. Specimens used were stainless steel (SUS304) pipes (318.5 mm outside diameter and 15.4 mm thickness), and pulsed gas tungsten-arc (GTA) welder was adopted. We have found process conditions to form sound root bead by changing both heat input conditions and joint designs. It is found that reducing volume of molten metal is necessary to form sound root bead. And it is also found that changing joint designs is effective to reduce volume of molten metal. By selecting proper joint designs, we could form sound root bead in constant heat input conditions in every position of pipe. (author)
Study of a two-pipe chilled beam system for both cooling and heating of office buildings
Energy Technology Data Exchange (ETDEWEB)
Norouzi, R. [Univ. of Boraes, Boraes (Sweden); Hultmark, G. [Lindab Comfort A/S, Farum (Denmark); Afshari, A. (ed.); Bergsoee, N.C. [Aalborg Univ.. Statens Byggeforskningsinstitut (SBi), Copenhagen (Denmark)
2013-05-15
The main aim of this master thesis was to investigate possibilities and limitations of a new system in active chilled beam application for office buildings. Lindab Comfort A/S pioneered the presented system. The new system use two-pipe system, instead of the conventional active chilled beam four-pipe system for heating and cooling purposes. The Two-Pipe System which is studied in this project use high temperature cooling and low temperature heating with water temperatures of 20 deg. C to 23 deg. C, available for free most of the year. The system can thus take advantage of renewable energy. It was anticipated that a Two-Pipe System application enables transfer of energy from warm spaces to cold spaces while return flows, from cooling and heating beams, are mixed. BSim software was chosen as a simulation tool to model a fictional office building and calculate heating and cooling loads of the building. Moreover, the effect of using outdoor air as a cooling energy source (free cooling) is investigated through five possible scenarios in both the four pipe system and the Two-Pipe System. The calculations served two purposes. Firstly, the effect of energy transfer in the Two-Pipe System were calculated and compared with the four pipe system. Secondly, free cooling effect was calculated in the Two-Pipe System and compared with the four pipe system. The simulation results showed that the energy transfer, as an inherent characteristic in the Two-Pipe System, is able to reduce up to 3 % of annual energy use compared to the four pipe system. Furthermore, different free cooling applications in the Two-Pipe System and the four pipe system respectively showed that the Two-Pipe System requires 7-15 % less total energy than the four pipe system in one year. In addition, the Two-Pipe System can save 18-57 % of annual cooling energy when compared to the four pipe system. (Author)
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
Assessment of water pipes durability under pressure surge
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.
International Nuclear Information System (INIS)
Minato, Akihiko; Nagoyoshi, Takuji; Nakamura, Akira; Fujii, Yuzo; Aya, Izuo; Yamane, Kenji
2004-01-01
Subcooled water injection into steam flow in piping systems may generate a water column containing a large steam slug. The steam slug collapses due to rapid condensation and interfaces on both sides collides with each other. Water hammer takes place and sharp pressure pulse propagates through the pipe. The purpose of this study is to show capability of the present numerical simulation method for predictions of pressure transient and loads on a piping system following steam slug collapse. A three-dimensional computer code for transient gas-liquid two-phase flow was applied to simulate an experiment of steam-condensation-induced water hammer with a horizontal polycarbonate pipe. The code was based on the extended two-fluid model, which treated interface motion using the VOF (Volume of Fluid) technique. The Godunov scheme of highly compressible single-phase flow was modified for application to the Riemann problem solution of gas-liquid mixture. Analysis of local steam slug collapse resulted in comparable peak pressure and pulse width of pressure transients with the observation. The calculation of pressure pulse propagation and impact load on piping system showed the quasi-steady pressure load was imposed especially on elbow at 1/10 of water hammer peak pressure. (author)
International Nuclear Information System (INIS)
Nogueira, A.C.R.
1981-10-01
The numerical calculation of the main variables of the laminar, incompressible, axissimmetric, steady flow around a circunferential square obstacle placed at the wall of a circular pipe, is done. The velocity profiles, the separating length and the shape of the separating streamline are compared with experimental available data and a good agreement is achieved. (E.G.) [pt
Directory of Open Access Journals (Sweden)
Ney Procópio Lopes
2011-12-01
Full Text Available A presente pesquisa em escala piloto, simulando trecho de uma rede interligada a um ramal predial, visou quantificar o volume de ar aferido pelos hidrômetros residenciais. Para tal fim, testaram-se ventosas, bloqueadores de ar e válvulas eliminadoras de ar. Sob condições normais de operação da rede, o volume de ar medido pelos hidrômetros é comparável ao encontrado na água natural, não justificando a instalação de equipamento de eliminação de ar de qualquer natureza. Todavia, logo após esvaziamento da rede interligada ao ramal, a sobremedição pode atingir até 21% em condições de pressão máxima na rede de distribuição (500 kPa. Por fim, verificou-se que a menor vazão afluente associa-se ao maior volume de ar aferido pelo hidrômetro. Dessa forma, é possível supor que os consumidores situados na menor faixa de consumo sejam os mais prejudicados pela situação de desabastecimento.The purpose of the present work is to evaluate the volume of air measured in domestic water supply pipe connections. Tests were performed to evaluate the efficiency and applicability of air reducing valves in domestic water supply connections. The results obtained under regular water supply conditions showed that the volume of air in the water measured by the hydrometers is comparable to the one found in natural waters. On the other hand, other tests, right after emptying the network connected to the domestic water supply pipe, revealed that the volume of water which gets to the gauged reservoir comprehends up to 21% of the total air-water volume recorded by the hydrometer for the experiments performed under pressure of 500 kPa.
Reza Khoshravan Azar, Mohammad; Emami Satellou, Ali Akbar; Shishesaz, Mohammad; Salavati, Bahram
2013-04-01
Given the increasing use of composite materials in various industries, oil and gas industry also requires that more attention should be paid to these materials. Furthermore, due to variation in choice of materials, the materials needed for the mechanical strength, resistance in critical situations such as fire, costs and other priorities of the analysis carried out on them and the most optimal for achieving certain goals, are introduced. In this study, we will try to introduce appropriate choice for use in the natural gas transmission composite pipelines. Following a 4-layered filament-wound (FW) composite pipe will consider an offer our analyses under internal pressure. The analyses' results will be calculated for different combinations of angles 15 deg, 30 deg, 45 deg, 55 deg, 60 deg, 75 deg, and 80 deg. Finally, we will compare the calculated values and the optimal angle will be gained by using the Approximation methods. It is explained that this layering is as the symmetrical.
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)
Increasing the volumetric efficiency of Diesel engines by intake pipes
List, Hans
1933-01-01
Development of a method for calculating the volumetric efficiency of piston engines with intake pipes. Application of this method to the scavenging pumps of two-stroke-cycle engines with crankcase scavenging and to four-stroke-cycle engines. The utility of the method is demonstrated by volumetric-efficiency tests of the two-stroke-cycle engines with crankcase scavenging. Its practical application to the calculation of intake pipes is illustrated by example.
MAXIMUM AIR SUCTION INTO HORIZONTAL OPEN ENDED CYLINDRICAL LOUVERED PIPE
Directory of Open Access Journals (Sweden)
SAMEER RANJAN SAHU
2017-02-01
Full Text Available The main approach behind the present numerical investigation is to estimate the mass flow rate of air sucked into a horizontal open-ended louvered pipe from the surrounding atmosphere. The present numerical investigation has been performed by solving the conservation equations for mass, momentum and energy along with two equation based k-ɛ model for a louvered horizontal cylindrical pipe by finite volume method. It has been found from the numerical investigation that mass suction rate of air into the pipe increases with increase in louvered opening area and the number of nozzles used. Keeping other parameters fixed, for a given mass flow rate there exists an optimum protrusion of nozzle for highest mass suction into the pipe. It was also found from the numerical investigation that increasing the pipe diameter the suction mass flow rate of air was increased.
105-KW Sandfilter Backwash Pit sludge volume calculation
International Nuclear Information System (INIS)
Dodd, E.N. Jr.
1995-01-01
The volume of sludge contained in the 100-KW Sandfilter Backwash Pit (SFBWP) was calculated from depth measurements of the sludge, pit dimension measurements and analysis of video tape recordings taken by an underwater camera. The term sludge as used in this report is any combination of sand, sediment, or corrosion products visible in the SFBWP area. This work was performed to determine baseline volume for use in determination of quantities of uranium and plutonium deposited in the pit from sandfilter backwashes. The SFBWP has three areas where sludge is deposited: (1) the main pit floor, (2) the transfer channel floor, and (3) the surfaces and structures in the SFBWP. The depths of sludge and the uniformity of deposition varies significantly between these three areas. As a result, each of the areas was evaluated separately. The total volume of sludge determined was 3.75 M 3 (132.2 ft 3 )
Influence of plastic deformation on seismic response of piping
International Nuclear Information System (INIS)
Yao Yanping; Chen Yong; Lu Mingwan
2000-01-01
On the basis of a brief summary of linear elastic seismic analysis methods, the importance for consideration of plastic deformation during the dynamic response analysis of piping system is indicated. The present methods of considering plasticity and the disadvantages of these methods are discussed. And the authors point out that in order to reduce the conservatism of present codes and to put forward appropriate and realistic piping seismic design methods, the key is to understand the plastic dynamic failure mode for piping under seismic excitation and to calculate the inelastic energy dissipation. The analysis and evaluation are applicable to nuclear piping systems
Transient Simulation of Accumulating Particle Deposition in Pipe Flow
Hewett, James; Sellier, Mathieu
2015-11-01
Colloidal particles that deposit in pipe systems can lead to fouling which is an expensive problem in both the geothermal and oil & gas industries. We investigate the gradual accumulation of deposited colloids in pipe flow using numerical simulations. An Euler-Lagrangian approach is employed for modelling the fluid and particle phases. Particle transport to the pipe wall is modelled with Brownian motion and turbulent diffusion. A two-way coupling exists between the fouled material and the pipe flow; the local mass flux of depositing particles is affected by the surrounding fluid in the near-wall region. This coupling is modelled by changing the cells from fluid to solid as the deposited particles exceed each local cell volume. A similar method has been used to model fouling in engine exhaust systems (Paz et al., Heat Transfer Eng., 34(8-9):674-682, 2013). We compare our deposition velocities and deposition profiles with an experiment on silica scaling in turbulent pipe flow (Kokhanenko et al., 19th AFMC, 2014).
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)
Monitoring of pipe displacements in French LMFBR SUPERPHENIX
International Nuclear Information System (INIS)
Foucher, N.; Debaene, J.P.; Renault, Y.; Blin, B.
1993-01-01
In order to check that pipe supports work properly and that the locking of snubbers or the loss of supports do not put a pipe in unacceptable loading conditions, a monitoring of the behaviour of the main pipes of SUPERPHENIX is planned. This monitoring system consists in measuring the displacements at selected points of the pipe by means of measuring rods and checking that these displacements remain inside allowable domains. These allowable domains are defined so that, if the displacements of the pipe are inside all these domains, the plant operator is sure that the stresses verify the allowable limits and then no additional inspection is carried out. In the opposite case, the operator will inspect the pipe in detail in order to determine the consequences and repair if necessary before restarting. Selection of points for monitoring was done with the to minimize the number of measures to be carried out and to use as far as possible the measuring rods that were installed to check that pipe displacements were consistent with what has been obtained in design calculations. However, it appears necessary to ensure that any incident occurring at any point of the pipe can be detected and, if necessary, additional measuring rods may be installed. An incident is said detectable if it induces on at least one measuring rod a deviation with respect to expected displacement not lower than 5 mm. It has been chosen so that small normal changes in measured displacements are not mistaken as incidents. The incidents that are supposed likely to occur are: 1) loss of a support which induces mainly primary stresses, 2) locking of a snubber which induces mainly secondary stresses. Monitoring of pipe displacements is a simple and effective way of checking that no damaging perturbation has occurred on the pipe. Calculations carried out on the DHR loops of SUPERPHENIX show that allowable domains of acceptable size may be obtained using a relatively small number of measuring rods. The method
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.
International Nuclear Information System (INIS)
Fukue, Hisayoshi; Mochizuki, Yoji; Nakamura, Harushige; Kobo, Hiroshi; Nitta, Tetsuo; Kawakami, Kiyoshi
1986-01-01
A pipe bending apparatus has recently been developed by applying high frequency induction heating. However, the smaller the radius of pipe bending, the greater becomes the reduction in wall thickness and the ovality of the pipe form. This makes it impossible to manufacture pipe bending which will meet the nuclear pipe design code. In order to solve this problem it is crucial to obtain a temperature distributions in a pipe which is moving. It is calculated by giving the following boundary conditions : distribution of the heat generation rate, and that of heat transfer of cooling water. In the process of analyzing these distributions, the following results were obtained. (1) The distribution of the heat generation rate is determined by the sink of energy flux of Poynting vectors. The coil efficiency thus calculated was sixty percent. This figure accords with the test data. (2) The distribution of heat transfer coefficient of cooling water is mainly determined by the rate of liquid film heat transfer, but departure from nucleate boiling and dryout has to be taken into consideration. (3) TRUMP CODE is modified so that the temperature distribution in moving pipes can be calculated by taking the boundary conditions into account. The calculated results were in accordance with the test data. (author)
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.)
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
Pipe rupture hardware minimization in pressurized water reactor system
International Nuclear Information System (INIS)
Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.
1987-01-01
For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination
Kimberlite Wall Rock Fragmentation: Venetia K08 Pipe Development
Barnett, W.; Kurszlaukis, S.; Tait, M.; Dirks, P.
2009-05-01
Volcanic systems impose powerful disrupting forces on the country rock into which they intrude. The nature of the induced brittle deformation or fragmentation can be characteristic of the volcanic processes ongoing within the volcanic system, but are most typically partially removed or obscured by repeated, overprinting volcanic activity in mature pipes. Incompletely evolved pipes may therefore provide important evidence for the types and stages of wall rock fragmentation, and mechanical processes responsible for the fragmentation. Evidence for preserved stages of fragmentation is presented from a detailed study of the K08 pipe within the Cambrian Venetia kimberlite cluster, South Africa. This paper investigates the growth history of the K08 pipe and the mechanics of pipe development based on observations in the pit, drill core and thin sections, from geochemical analyses, particle size distribution analyses, and 3D modeling. Present open pit exposures of the K08 pipe comprise greater than 90% mega-breccia of country rock clasts (gneiss and schist) with Drill core shows that below about 225 m the CRB includes increasing quantities of kimberlite. The breccia clasts are angular, clast-supported with void or carbonate cement between the clasts. Average clast sizes define sub-horizontal layers tens of metres thick across the pipe. Structural and textural observations indicate the presence of zones of re-fragmentation or zones of brittle shearing. Breccia textural studies and fractal statistics on particle size distributions (PSD) is used to quantify sheared and non- sheared breccia zones. The calculated energy required to form the non-sheared breccia PSD implies an explosive early stage of fragmentation that pre-conditions the rock mass. The pre-conditioning would have been caused by explosions that are either phreatic or phreatomagmatic in nature. The explosions are likely to have been centered on a dyke, or pulses of preceding volatile-fluid phases, which have
Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation
International Nuclear Information System (INIS)
CAROTHERS, K.G.
1999-01-01
Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18, 1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30, 1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''
Waste Retrieval Sluicing System Campaign Number 3 Solids Volume Transferred Calculation
International Nuclear Information System (INIS)
CAROTHERS, K.G.
1999-01-01
Waste Retrieval Sluicing System (WRSS) operations at tank 241-C-106 began on Wednesday, November 18,1998. The purpose of this system is to retrieve and transfer the high-heat sludge from the tank for storage in double-shell tank 241-AY-102, thereby resolving the high-heat safety issue for the tank, and to demonstrate modernized past-practice retrieval technology for single-shell tank waste. Performance Agreement (PA) TWR 1.2.2, C-106 Sluicing, was established by the Department of Energy, Office of River Protection (ORP) for achieving completion of sluicing retrieval of waste from tank 241-C-106 by September 30,1999. This level of sludge removal is defined in the PA as either removal of approximately 72 inches of sludge or removal of 172,000 gallons of sludge (approximately 62 inches) and less than 6,000 gallons (approximately 2 inches) of sludge removal per 12 hour sluice batch for three consecutive batches. Preliminary calculations of the volume of tank 241-C-106 sludge removed as of September 29, 1999 were provided to ORP documenting completion of PA TWR 1.2.2 (Allen 1999a). The purpose of this calculation is to document the final sludge volume removed from tank 241-C-106 up through September 30, 1999. Additionally, the results of an extra batch completed October 6, 1999 is included to show the total volume of sludge removed through the end of WRSS operations. The calculation of the sludge volume transferred from the tank is guided by engineering procedure HNF-SD-WM-PROC-021, Section 15.0,Rev. 3, sub-section 4.4, ''Calculation of Sludge Transferred.''
EARTHWORK VOLUME CALCULATION FROM DIGITAL TERRAIN MODELS
Directory of Open Access Journals (Sweden)
JANIĆ Milorad
2015-06-01
Full Text Available Accurate calculation of cut and fill volume has an essential importance in many fields. This article shows a new method, which has no approximation, based on Digital Terrain Models. A relatively new mathematical model is developed for that purpose, which is implemented in the software solution. Both of them has been tested and verified in the praxis on several large opencast mines. This application is developed in AutoLISP programming language and works in AutoCAD environment.
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.)
Seminar on countermeasures for pipe cracking in BWRs. Volume 4 of 4
Energy Technology Data Exchange (ETDEWEB)
None
1980-05-01
Intergranular stress corrosion cracking of welded type 304 stainless steel in the recirculation piping of boiling water reactors has had an impact on plant availability and reliability since the fall of 1974. Investigations of this problem have resulted in significant progress in understanding the phenomenon and providing an engineering resolution by developing and qualifying countermeasures. A number of these countermeasures including solution heat treatment, corrosion resistant clad, alternate pipe materials, induction heating stress improvement and heat sink welding have been implemented. Separate abstracts are included for each of the papers presented.
Seminar on countermeasures for pipe cracking in BWRs. Volume 2 of 4
Energy Technology Data Exchange (ETDEWEB)
None
1980-05-01
Intergranular stress corrosion cracking of welded type 304 stainless steel in the recirculation piping of boiling water reactors has had an impact on plant availability and reliability since the fall of 1974. Investigtions of this problem have resulted in significant progress in understanding the phenomenon and providing an engineering resolution by developing and qualifying countermeasures. A number of these countermeasures including solution heat treatment, corrosion resistant clad, alternate pipe materials, induction heating stress improvement and heat sink welding have been implemented. Separate abstracts are included for each of the papers presented.
Haas, Matthias; Günzel, Karsten; Miller, Kurt; Hamm, Bernd; Cash, Hannes; Asbach, Patrick
2017-01-01
Prostate volume in multiparametric MRI (mpMRI) is of clinical importance. For 3-Tesla mpMRI without endorectal coil, there is no distinctive standard for volume calculation. We tested the accuracy of the ellipsoid formula with planimetric volume measurements as reference and investigated the correlation of gland volume and cancer detection rate on MRI/ultrasound (MRI/US) fusion-guided biopsy. One hundred forty-three patients with findings on 3-Tesla mpMRI suspicious of cancer and subsequent MRI/US fusion-guided targeted biopsy and additional systematic biopsy were analyzed. T2-weighted images were used for measuring the prostate diameters and for planimetric volume measurement by a segmentation software. Planimetric and calculated prostate volumes were compared with clinical data. The median prostate volume was 48.1 ml (interquartile range (IQR) 36.9-62.1 ml). Volume calculated by the ellipsoid formula showed a strong concordance with planimetric volume, with a tendency to underestimate prostate volume (median volume 43.1 ml (IQR 31.2-58.8 ml); r = 0.903, p Tesla mpMRI without endorectal coil. It allows a fast, valid volume calculation in prostate MRI datasets. © 2016 S. Karger AG, Basel.
International Nuclear Information System (INIS)
Jasso G, J.; Diaz S, A.; Mendoza G, G.; Sainz M, E.; Garcia de la C, F. M.
2014-10-01
The growth and the cracks propagation by fatigue are a typical degradation mechanism that is presented in the nuclear industry as in the conventional industry; the unstable propagation of a crack can cause the catastrophic failure of a metallic component even with high ductility; for this reason, activities of programmed maintenance have been established in the industry using inspection and visual techniques and/or ultrasound with an established periodicity allowing to follow up to these growths, controlling the undesirable effects; however, these activities increase the operation costs; and in the peculiar case of the nuclear industry, they increase the radiation exposure to the participant personnel. The use of mathematical processes that integrate concepts of uncertainty, material properties and the probability associated to the inspection results, has been constituted as a powerful tool of evaluation of the component reliability, reducing costs and exposure levels. In this work the evaluation of the failure probability by cracks growth preexisting by fatigue is presented, in pipes of a Reactor Core Isolation Cooling system (Rcic) in a nuclear power station. The software WinPRAISE 07 (Piping Reliability Analysis Including Seismic Events) was used supported in the probabilistic fracture mechanics principles. The obtained values of failure probability evidenced a good behavior of the analyzed pipes with a maximum order of 1.0 E-6, therefore is concluded that the performance of the lines of these pipes is reliable even extrapolating the calculations at 10, 20, 30 and 40 years of service. (Author)
Refined pipe theory for mechanistic modeling of wood development.
Deckmyn, Gaby; Evans, Sam P; Randle, Tim J
2006-06-01
We present a mechanistic model of wood tissue development in response to changes in competition, management and climate. The model is based on a refinement of the pipe theory, where the constant ratio between sapwood and leaf area (pipe theory) is replaced by a ratio between pipe conductivity and leaf area. Simulated pipe conductivity changes with age, stand density and climate in response to changes in allocation or pipe radius, or both. The central equation of the model, which calculates the ratio of carbon (C) allocated to leaves and pipes, can be parameterized to describe the contrasting stem conductivity behavior of different tree species: from constant stem conductivity (functional homeostasis hypothesis) to height-related reduction in stem conductivity with age (hydraulic limitation hypothesis). The model simulates the daily growth of pipes (vessels or tracheids), fibers and parenchyma as well as vessel size and simulates the wood density profile and the earlywood to latewood ratio from these data. Initial runs indicate the model yields realistic seasonal changes in pipe radius (decreasing pipe radius from spring to autumn) and wood density, as well as realistic differences associated with the competitive status of trees (denser wood in suppressed trees).
Influence of transient flow in the formation of condensate and in the calculation of steam line
International Nuclear Information System (INIS)
Bazzo, E.
1989-01-01
The piping design is analyzed in unsteady-state conditions, with the main goal of minimizing operational costs and initial investments of a plant. All heat losses are calculated by applying the control volume method. The results confirm the applicability of the method and show that the influence of the transient regime on the condensation rate and economical insulation thickness must be considered. (author)
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.)
Design and analysis for piping systems
International Nuclear Information System (INIS)
Sterkel, H.-P.; Cutrim, J.H.C.
1981-01-01
The procedure and the typical techniques that are used in NUCLEN for the design and the calculation of the piping of Nuclear Plants. The classification system are generically described and the analysis techniques which are used for the design and verification of the piping systems, i.e. pressure design for the dimensioning of the wallthicknesses, temperature and dead weight analysis together with determination of support points, are shown. The techniques of dynamic design and analyses are described for earthquake and pressure impulse loadings. (Author) [pt
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
Modeling of Pressure Drop During Refrigerant Condensation in Pipe Minichannels
Sikora, Małgorzata; Bohdal, Tadeusz
2017-12-01
Investigations of refrigerant condensation in pipe minichannels are very challenging and complicated issue. Due to the multitude of influences very important is mathematical and computer modeling. Its allows for performing calculations for many different refrigerants under different flow conditions. A large number of experimental results published in the literature allows for experimental verification of correctness of the models. In this work is presented a mathematical model for calculation of flow resistance during condensation of refrigerants in the pipe minichannel. The model was developed in environment based on conservation equations. The results of calculations were verified by authors own experimental investigations results.
A simplified dynamic analysis for reactor piping systems under blowdown conditions
International Nuclear Information System (INIS)
Chen, M.M.
1975-01-01
In the design of pipelines in a nuclear power plant for blowdown conditions, is it customary to conduct dynamic analysis of the piping system to obtain the responses and the resulting stresses. Calculations are repeated for each design modification in piping geometry or supporting system until the design codes are met. The numerical calculations are, in general, very costly and time consuming. Until now, there have been no simple means for calculating the dynamic responses for the design. The proposed method reduces the dynamic calculation to a quasi-static one, and can be beneficially used for the preliminary design. The method is followed by a complete dynamical analysis to improve the final results. The new formulations greatly simplify the numerical computation and provide design guides. When used to design a given piping system, the method saved approximately one order of magnitude of computer time. The approach can also be used for other types of structures
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
An investigation for design and operational procedures to avoid water hammer in NPP piping systems
International Nuclear Information System (INIS)
Kim, Jin Weon
1993-02-01
To predict waterhammer initiation due to water slug formation in the horizontal section of piping system and to calculate its impact pressure by using the results of waterhammer initiation model, waterhammer initiation model and impact pressure calculation model have been developed. In the impact pressure calculation model, the effects of water layer depth at which water slug formation and water temperature variation with time and space have been included to calculate a more realistic impact pressure. Prediction of waterhammer initiation is compared with experimental data for the various 'L' shaped pipes. The results show that the present waterhammer initiation model well predicts the waterhammer initiation boundary for inverted vertical 'L' shaped pipe filled through the horizontal pipe. Impact pressure calculated by present model also gives good agreement with the range of impact pressure of steam bubble collapse experiment. Impact pressure is calculated at the waterhammer initiation boundary by using the conditions obtained from waterhammer initiation model. From this result, it is seen that low inlet subcooling results in not only low frequency of waterhammer but also minor impact pressure if it does occur
ICPP tank farm closure study. Volume 2: Engineering design files
International Nuclear Information System (INIS)
1998-02-01
Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks
ICPP tank farm closure study. Volume 2: Engineering design files
Energy Technology Data Exchange (ETDEWEB)
NONE
1998-02-01
Volume 2 contains the following topical sections: Tank farm heel flushing/pH adjustment; Grouting experiments for immobilization of tank farm heel; Savannah River high level waste tank 20 closure; Tank farm closure information; Clean closure of tank farm; Remediation issues; Remote demolition techniques; Decision concerning EIS for debris treatment facility; CERCLA/RCRA issues; Area of contamination determination; Containment building of debris treatment facility; Double containment issues; Characterization costs; Packaging and disposal options for the waste resulting from the total removal of the tank farm; Take-off calculations for the total removal of soils and structures at the tank farm; Vessel off-gas systems; Jet-grouted polymer and subsurface walls; Exposure calculations for total removal of tank farm; Recommended instrumentation during retrieval operations; High level waste tank concrete encasement evaluation; Recommended heavy equipment and sizing equipment for total removal activities; Tank buoyancy constraints; Grout and concrete formulas for tank heel solidification; Tank heel pH requirements; Tank cooling water; Evaluation of conservatism of vehicle loading on vaults; Typical vault dimensions and approximately tank and vault void volumes; Radiological concerns for temporary vessel off-gas system; Flushing calculations for tank heels; Grout lift depth analysis; Decontamination solution for waste transfer piping; Grout lift determination for filling tank and vault voids; sprung structure vendor data; Grout flow properties through a 2--4 inch pipe; Tank farm load limitations; NRC low level waste grout; Project data sheet calculations; Dose rates for tank farm closure tasks; Exposure and shielding calculations for grout lines; TFF radionuclide release rates; Documentation of the clean closure of a system with listed waste discharge; and Documentation of the ORNL method of radionuclide concentrations in tanks.
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)
Generic safety evaluation report regarding integrity of BWR scram system piping
International Nuclear Information System (INIS)
1981-08-01
Safety concerns associated with postulated pipe breaks in the boiling water reactor (BWR) scram system were identified during the staff's continuing investigation of the Browns Ferry Unit 3 control rod partial insertion failure on June 28, 1980. This report includes an evaluation of the licensing basis for the BWR scram discharge volume (SDV) piping and an assessment of the potential for the SDV piping to fail while in service. A discussion of the means available for mitigation an unlikely SDV system failure is provided. Generic recommendations are made to improve mitigation capability and ensure that system integrity is maintained in service
Programmable calculator programs to solve softwood volume and value equations.
Janet K. Ayer. Sachet
1982-01-01
This paper presents product value and product volume equations as programs for handheld calculators. These tree equations are for inland Douglas-fir, young-growth Douglas-fir, western white pine, ponderosa pine, and western larch. Operating instructions and an example are included.
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
International Nuclear Information System (INIS)
Uehara, Yasushi; Uchida, Shunsuke; Naitoh, Masanori; Okada, Hidetoshi; Koshizuka, Seiichi
2009-01-01
In order to predict and mitigate flow accelerated corrosion (FAC) of carbon steel piping in PWR and BWR secondary systems, computer program packages for evaluating FAC have been developed by coupling one through three dimensional (1-3D) computational flow dynamics (CFD) models and corrosion models. To evaluate corrosive conditions, e.g., oxygen concentration and electrochemical corrosion potential (ECP) along the flow path, flow pattern and temperature in each elemental volume were obtained with 1D computational flow dynamics (CFD) codes. Precise flow turbulence and mass transfer coefficients at the structure surface were calculated with 3D CFD codes to determine wall thinning rates. One of the engineering options is application of k-ε calculation as a 3D CFD code, which has limitation of detail evaluation of flow distribution at very surface of large scale piping. A combination of k-ε calculation and wall function was proposed to evaluate precise distribution of mass transfer coefficients with reasonable CPU volume and computing time and, at the same time, reasonable accuracy. (author)
International Nuclear Information System (INIS)
Wang, Y.K.; Subudhi, M.; Bezler, P.
1983-01-01
This report includes the findings of an investigation of the conservatism associated with different combinations between the primary and secondary stress components for piping systems under dynamic loading, such as in an earthquake event. The primary stresses are induced by piping response to its mass inertia effects. The secondary stresses are induced by relative displacements of piping supports. The study involves an independnent time history analysis of several typical piping models to predict a best estimate of the actual dynamic and pseudo-static pipe responses to an earthquake. These piping systems are also analyzed using the response spectrum method to obtain the maximum primary stress components. Secondary stresses are next calculated by performing a set of static analyses which provide the worst stress condition. The two components are then combined by both SRSS and absolute sum methods as the results are compared with time history solutions. It is found that the SRSS combination of the primary and secondary stress components yield acceptable results provided the secondary stress component is calculated in the most unfavorable phasing relationship among displacements of piping supports
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
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
International Nuclear Information System (INIS)
Sanchez Mazon, J.; Raba Diez, J. L.; Vazquez Rodriguez, J. A.; Pacheco Baldor, M. T.; Mendiguren Santiago, M. A.
2011-01-01
In the Protocol for the control treatment planning systems with ionizing radiation of the proposed SEFM tests to verify proper operation of the calculation in the evaluation of DVH (Dose Volume Histogram). The calculation of the volume that makes a planner may have important implications because it can trigger an overestimation of the dose or otherwise. We present a comparison of the calculation of volumes estimated with 4 different planners.
Computer aided design of piping for a radiochemical plant
Energy Technology Data Exchange (ETDEWEB)
Selvaraj, P G; Chandrasekhar, A; Chandrasekar, A V [Reprocessing Group, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Raju, R P; Mahudeeswaran, K V; Kumar, S V [Reprocessing Group, Bhabha Atomic Research Centre, Mumbai (India)
1994-06-01
In a radiochemical plant such as reprocessing plants, process equipment, storage tanks, liquid transfer systems and the associated pipe lines etc. are housed in series of concrete cells. Availability of limited cell space/volume, provision of various modes of liquid transfers with associated redundancies and instrumentation lines with standby alternatives increase the overall piping density. Designing such high density piping layout without interference is quite complex and needs lot of human efforts. This paper briefly describes development of computer codes for the entire scheme of design, drafting and fabrication of piping for nuclear fuel reprocessing plant. The general organisation of various programs, their functions, the complete sequence of the scheme and the flow of data are presented. High degree of reliability of each routine, considerable error checking facilities, marking legends on the drawings, provision for scaling in drafting and accuracy to the extent of one mm in layout design are some of the important features of this scheme. (author). 1 fig.
Study on the stress and strain during welding of plate-to-pipe joint
Energy Technology Data Exchange (ETDEWEB)
Na, S.J.; Kim, H.W.
1986-09-01
In manufacturing of pipe walls for boiler units, distortion can result in pipe-to-pipe joints from the nonuniform expansion and contraction of the weld metal and the adjacent base metal during heating and cooling cycle of the welding process. In this study, the stresses and strains during longitudinal welding of the plate-to-pipe joint were investigated. Using the method of sucessive elastic solution, longitudinal stresses and strains during and after welding were calculated from the information of temperature distributions obtained by Rosenthal's equations. In order to confirm the validity of the numerical results, the temperature and residual stress distributions were measured and compared with the calculated results. In spite of some assumptions, the one-dimensional analytical results of residual stresses were in fairly good agreement with the experimental ones. The residual stresses due to welding of plate-to-pipe joints are tensile near the weld line and compressive in the base metal as in the welding of plates. The amount and distribution of residual stresses were deeply dependent on the heat input ratio of the plate and pipe.
Analysis of NPP pipes and equipment damage in life time prolongation
International Nuclear Information System (INIS)
Tkachev, V.V.; Zheltukhin, K.K.
2008-01-01
Paper describes a procedure to calculate the probability of pipes and equipment failure taking account of both the service records of the structures under various conditions and their aging. The parameters characterizing applied loads, failures, as well as metal strength, mechanical and thermal properties serve as the arbitrary values used in the described procedure. Paper presents an example of the probability calculation of failure of the RBMK emergency feed pump recirculation pipes when their service life is prolonged [ru
Fatigue analysis of HANARO primary cooling system piping
International Nuclear Information System (INIS)
Ryu, Jeong Soo
1998-05-01
A main form of piping failure which occurring leak before break (LBB) is fatigue failure. The fatigue analysis of HANARO primary cooling system (PCS) piping was performed. The PCS piping had been designed in accordance with ASME Class 3 for service conditions. However fatigue analysis is not required in Class 3. In this study the quantitative fatigue analysis was carried out according to ASME Class 1. The highest stress points which have the largest possibility of ASME class 1. The highest stress points which have the largest possibility of the fatigue were determined from the piping stress analysis for each subsection piping. The fatigue analysis was performed for 3 highest stress points, i.e., branch connection, anchor point and butt welding joint. After calculating the peak stress intensity range the fatigue usage factors were evaluated considering operating cycles and S-N curve. The cumulative usage factors for 3 highest stress points were much less than 1. The results show that the possibility of fatigue failure for PCS piping subjected to thermal expansion and seismic loads is very small. The structural integrity of the HANARO PCS piping for fatigue failure was proved to apply the LBB. (author). 11 tabs., 6 figs
40 CFR 80.596 - How is a refinery motor vehicle diesel fuel volume baseline calculated?
2010-07-01
... 40 Protection of Environment 16 2010-07-01 2010-07-01 false How is a refinery motor vehicle diesel... Requirements § 80.596 How is a refinery motor vehicle diesel fuel volume baseline calculated? (a) For purposes of this subpart, a refinery's motor vehicle diesel fuel volume baseline is calculated using the...
Fracture toughness evaluations of TP304 stainless steel pipes
International Nuclear Information System (INIS)
Rudland, D.L.; Brust, F.W.; Wilkowski, G.M.
1997-02-01
In the IPIRG-1 program, the J-R curve calculated for a 16-inch nominal diameter, Schedule 100 TP304 stainless steel (DP2-A8) surface-cracked pipe experiment (Experiment 1.3-3) was considerably lower than the quasi-static, monotonic J-R curve calculated from a C(T) specimen (A8-12a). The results from several related investigations conducted to determine the cause of the observed toughness difference are: (1) chemical analyses on sections of Pipe DP2-A8 from several surface-cracked pipe and material property specimen fracture surfaces indicate that there are two distinct heats of material within Pipe DP2-A8 that differ in chemical composition; (2) SEN(T) specimen experimental results indicate that the toughness of a surface-cracked specimen is highly dependent on the depth of the initial crack, in addition, the J-R curves from the SEN(T) specimens closely match the J-R curve from the surface-cracked pipe experiment; (3) C(T) experimental results suggest that there is a large difference in the quasi-static, monotonic toughness between the two heats of DP2-A8, as well as a toughness degradation in the lower toughness heat of material (DP2-A8II) when loaded with a dynamic, cyclic (R = -0.3) loading history
Development of solutions to benchmark piping problems
Energy Technology Data Exchange (ETDEWEB)
Reich, M; Chang, T Y; Prachuktam, S; Hartzman, M
1977-12-01
Benchmark problems and their solutions are presented. The problems consist in calculating the static and dynamic response of selected piping structures subjected to a variety of loading conditions. The structures range from simple pipe geometries to a representative full scale primary nuclear piping system, which includes the various components and their supports. These structures are assumed to behave in a linear elastic fashion only, i.e., they experience small deformations and small displacements with no existing gaps, and remain elastic through their entire response. The solutions were obtained by using the program EPIPE, which is a modification of the widely available program SAP IV. A brief outline of the theoretical background of this program and its verification is also included.
Design for whipping pipe impact on reinforced concrete panels
International Nuclear Information System (INIS)
Chen, C.C.; Gurbuz, O.
1984-01-01
This paper describes determination of local and overall effects on reinforced concrete panels due to whipping pipe impact in postulated pipe break events. Local damage includes the prediction of minimum concrete panel thickness required to prevent spalling from the back face of the target reinforced concrete panels. Evaluation of overall effect deals with the ductility ratio calculation for the target reinforced concrete panels. Design curves for determining the minimum panel thickness and the minimum reinforcement of reinforced concrete panels are presented in this paper for some cases commonly encountered in nuclear applications. The methodology and the results provided can be used to determine if an existing reinforced concrete wall is capable of resisting the whipping pipe impact, and consequently, if pipe whip restraints can be eliminated
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)
Numerical study of heat and mass transfer in inertial suspensions in pipes.
Niazi Ardekani, Mehdi; Brandt, Luca
2017-11-01
Controlling heat and mass transfer in particulate suspensions has many important applications such as packed and fluidized bed reactors and industrial dryers. In this work, we study the heat and mass transfer within a suspension of spherical particles in a laminar pipe flow, using the immersed boundary method (IBM) to account for the solid fluid interactions and a volume of fluid (VoF) method to resolve temperature equation both inside and outside of the particles. Tracers that follow the fluid streamlines are considered to investigate mass transfer within the suspension. Different particle volume fractions 5, 15, 30 and 40% are simulated for different pipe to particle diameter ratios: 5, 10 and 15. The preliminary results quantify the heat and mass transfer enhancement with respect to a single-phase laminar pipe flow. We show in particular that the heat transfer from the wall saturates for volume fractions more than 30%, however at high particle Reynolds numbers (small diameter ratios) the heat transfer continues to increase. Regarding the dispersion of tracer particles we show that the diffusivity of tracers increases with volume fraction in radial and stream-wise directions however it goes through a peak at 15% in the azimuthal direction. European Research Council, Grant No. ERC-2013-CoG- 616186, TRITOS; SNIC (the Swedish National Infrastructure for Computing).
Wall Thickness Measurement Of Insulated Pipe By Tangential Radiography Technique Using Ir 192
International Nuclear Information System (INIS)
Soedarjo
2000-01-01
Insulation pipe wall thickness by tangential radiography technique has been carried out using 41 Curie Iridium 192 source has activity for two carbon steel pipes. The outer diameter of the first pipe is 90 mm, wall thickness is 75.0 mm, source film film distance is 609.5 mm, source tangential point of insulation is 489.5 mm and exposure time 3 minute and 25 second. From the calculation, the first pipe thickness is found to be 12.54 mm and for the second pipe is 8.42 mm. The thickness is due to inaccuracy in reading the pipe thickness on radiography film and the geometry distortion radiation path
International Nuclear Information System (INIS)
Ohta, Takahiro; Terasaki, Toshio
2009-01-01
The new process called L-SIP (outer surface irradiated Laser Stress Improvement Process) is developed to improve the tensile residual stress of the inner surface near the butt welded joints of pipes in the compression stress. The temperature gradient occurs in the thickness of pipes in heating the outer surface rapidly by laser beam. By the thermal expansion difference between the inner surface and the outer surface, the compression plastic strain generates near the outer surface and the tensile plastic strain generates near the inner surface of pipes. The compression stress occurs near the inner surface of pipes by the plastic deformation. In this paper, the theoretical equation which calculates residual stress distribution from the inherent strain distribution in the thickness of pipes is derived. And, the relation between the distribution of temperature and the residual stress in the thickness is examined for various pipes size. (1) By rapidly heating from the outer surface, the residual stress near the inner surface of the pipe is improved to the compression stress. (2) Pipes size hardly affects the distribution of the residual stress in the stainless steel pipes for piping (JISG3459). (3) The temperature rising area from the outside is smaller, the area of the compression residual stress near the inner surface becomes wider. (author)
Turbulent Heat Transfer in Curved Pipe Flow
Kang, Changwoo; Yang, Kyung-Soo
2013-11-01
In the present investigation, turbulent heat transfer in fully-developed curved pipe flow with axially uniform wall heat flux has been numerically studied. The Reynolds numbers under consideration are Reτ = 210 (DNS) and 1,000 (LES) based on the mean friction velocity and the pipe radius, and the Prandtl number (Pr) is 0.71. For Reτ = 210 , the pipe curvature (κ) was fixed as 1/18.2, whereas three cases of κ (0.01, 0.05, 0.1) were computed in the case of Reτ = 1,000. The mean velocity, turbulent intensities and heat transfer rates obtained from the present calculations are in good agreement with the previous numerical and experimental results. To elucidate the secondary flow structures due to the pipe curvature, the mean quantities and rms fluctuations of the flow and temperature fields are presented on the pipe cross-sections, and compared with those of the straight pipe flow. To study turbulence structures and their influence on turbulent heat transfer, turbulence statistics including but not limited to skewness and flatness of velocity fluctuations, cross-correlation coefficients, an Octant analysis, and turbulence budgets are presented and discussed. Based on our results, we attempt to clarify the effects of Reynolds number and the pipe curvature on turbulent heat transfer. This research was supported by Basic Science Research Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Education, Science and Technology (2010-0008457).
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
Comparative study on heat pipe performance using aqueous solutions of alcohols
Energy Technology Data Exchange (ETDEWEB)
Senthilkumar, R.; Vaidyanathan, S.; Sivaraman, B. [Annamalai University, Department of Mechanical Engineering, Annamalai Nagar, Tamil Nadu (India)
2012-12-15
This paper deals with the performance characterization of heat pipes using an aqueous solution of long chain alcohols like n-Butanol, n-Pentanol, n-Hexanol and n-Heptanol as working mediums. These solutions are called as self-rewetting fluids, since these fluid mixtures possess a non-linear dependence of the surface tension with temperature. A cylindrical heat pipe made up of copper with two layers of wrapped screen is used as a wick material and partially filled with the self-rewetting fluid water mixture and tested for its heat transport capability like thermal efficiency and thermal resistance at different inclinations and input power levels. A number of tests have been performed with heat pipes, filled with various aqueous solutions of alcohols with a concentration of 2 ml/l in de-ionized water (DI water) on volume basis. The results obtained for heat pipes using self rewetting fluids show improved performances, when compared to DI water heat pipes. (orig.)
Rectangular-section mirror light pipes
Energy Technology Data Exchange (ETDEWEB)
Swift, P.D.; Lawlor, R. [School of Physical Sciences, Dublin City University, Dublin 9 (Ireland); Smith, G.B.; Gentle, A. [Department of Applied Physics, University of Technology, Sydney, Broadway, NSW 2007 (Australia)
2008-08-15
Using an integrated-ray approach an expression for the transmission of rectangular section mirror light pipe (MLP) has been derived for the case of collimated light input. The transmittance and the irradiance distribution at the exit aperture of rectangular-section MLPs have been measured experimentally and calculated theoretically for the case of collimated light input. The results presented extend the description of MLPs from the cylindrical case. Measured and calculated transmittances and irradiance distributions are in good agreement. (author)
Nuclear power plant piping damping parametric effects
International Nuclear Information System (INIS)
Ware, A.G.
1983-01-01
The present NRC guidelines for structural damping to be used in the dynamic stress analyses of nuclear power plant piping systems are generally considered to be overly conservative. As a result, plant designers have in many instances used a considerable number of seismic supports to keep stresses calculated by large scale piping computer codes below the allowable limits. In response to this problem, the NRC and EG and G Idaho are engaged in programs to evaluate piping system damping, in order to provide more realistic and less conservative values to be used in seismic analyses. To generate revised guidelines, solidly based on technical data, new experimental data need to be generated and assessed, and the parameters which influence piping system damping need to be quantitatively identified. This paper presents the current state-of-the-art knowledge in the United States on parameters which influence piping system damping. Examples of inconsistencies in the data and areas of uncertainty are explained. A discussion of programs by EG and G Idaho and other organizations to evaluate various effects is included, and both short and long range goals of the program are outlined
Energy Technology Data Exchange (ETDEWEB)
Ohk, Seung Min; Chung, Bum Jin [Kyunghee University, Yongin (Korea, Republic of)
2016-05-15
The Passive Cooling System (PCS) driven by natural forces drew research attention since Fukushima nuclear power plant accident. This study investigated the natural convection heat transfer inside of vertical pipe with emphasis on the phenomena regarding the boundary layer interaction. Numerical calculations were carried out using FLUENT 6.3. Experiments were performed for the parts of the cases to explore the accuracy of calculation. Based on the analogy, heat transfer experiment is replaced by mass transfer experiment using sulfuric acid copper sulfate (CuSO{sub 4}. H{sub 2}SO{sub 4}) electroplating system. The natural convection heat transfer inside a vertical pipe is studied experimentally and numerically. Experiments were carried out using sulfuric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) based on the analogy concept between heat and mass transfer system. Numerical analysis was carried out using FLUENT 6.3. It is concluded that the boundary layer interaction along the flow passage influences the heat transfer, which is affected by the length, diameter, and Prandtl number. For the large diameter and high Prandtl number cases, where the thermal boundary layers do not interfered along the pipe, the heat transfer agreed with vertical flat plate for laminar and turbulent natural convection correlation within 8%. When the flow becomes steady state, the forced convective flow appears in the bottom of the vertical pipe and natural convection flow appears near the exit. It is different behavior from the flow on the parallel vertical flat plates. Nevertheless, the heat transfer was not different greatly compared with those of vertical plate.
Design analysis of liquid metal pipe supports
International Nuclear Information System (INIS)
Margolin, L.L.; LaSalle, F.R.
1979-02-01
Design guidelines pertinent to liquid metal pipe supports are presented. The numerous complex conditions affecting the support stiffness and strength are addressed in detail. Topics covered include modeling of supports for natural frequency and stiffness calculations, support hardware components, formulas for deflection due to torsion, plate bending, and out-of-plane flexibility. A sample analysis and a discussion on stress analysis of supports are included. Also presented are recommendations for design improvements for increasing the stiffness of pipe supports and which were utilized in the FFTF system
A coupled approach for the three-dimensional simulation of pipe leakage in variably saturated soil
Peche, Aaron; Graf, Thomas; Fuchs, Lothar; Neuweiler, Insa
2017-12-01
In urban water pipe networks, pipe leakage may lead to subsurface contamination or to reduced waste water treatment efficiency. The quantification of pipe leakage is challenging due to inaccessibility and unknown hydraulic properties of the soil. A novel physically-based model for three-dimensional numerical simulation of pipe leakage in variably saturated soil is presented. We describe the newly implemented coupling between the pipe flow simulator HYSTEM-EXTRAN and the groundwater flow simulator OpenGeoSys and its validation. We further describe a novel upscaling of leakage using transfer functions derived from numerical simulations. This upscaling enables the simulation of numerous pipe defects with the benefit of reduced computation times. Finally, we investigate the response of leakage to different time-dependent pipe flow events and conclude that larger pipe flow volume and duration lead to larger leakage while the peak position in time has a small effect on leakage.
Strain measurements at the HDR-pipe-system under LOCA-load: Effects on elbows and displaced weldings
International Nuclear Information System (INIS)
Hunger, H.
1985-01-01
This paper characterizes some effects which have been detected during strain gauge measurements on a test piping with feed water check valve oscillating under blowdown-load. The ovalization of a pipe elbow subjected to in-plane-bending affects the connected straight pipe; this is shown by means of circumferential stresses. Very high LOCA-load produces plastic strain and changes the pipe dynamics. Artificial displaced welds increase the local strain but no defects have occurred. One example compares stresses from measurement and post-calculation. Moreover there are given some remarks on the optimization of the comparison of measurement and calculation. (orig.)
Investigations on the necessity of dose calculations for several planes of the target volume
International Nuclear Information System (INIS)
Richter, E.
1987-01-01
In radiotherapy planning, the shape of a target volume can at present be exactly delimited by means of computed tomography. A method often applied is to project the largest target volume scan on the plane of the central ray and to calculate the dose in this plane. This method does not allow to take into account any change of the target volume scan which will be mainly due to the body contours of the patient. The results of dose calculations made in several planes for pharyngeal and laryngeal tumors are presented. With this procedure, 33 out of 60 irradiation techniques for nine tumor sites meet the requirements with regard to the central ray plane. If several planes are regarded, this is only true for ten irradiation plans. If is therefore absolutely necessary to calculate the doses of several planes if the target volume has an irregular shape or if the body contours vary considerably. This is the only way to prevent a false treatment caused by possibly severe dose excesses or dose insufficiencies in radiotherapy. (orig.) [de
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.
Piping hydrodynamic loads for a PWR power up-rate with steam generator replacement
International Nuclear Information System (INIS)
Julie M Jarvis; Allen T Vieira; James M Gilmer
2005-01-01
Full text of publication follows: Pipe break hydrodynamic loads are calculated for various systems in a PWR for a Power Up-rate (PUR) with a Steam Generator Replacement (SGR). PUR with SGR can change the system pressures, mass flowrates and pipe routing/configuration. These changes can alter the steam generator piping steam/water hammer loads. This paper discusses the need to benchmark against the original design basis, the use of different modeling techniques, and lessons learned. Benchmarking for licensing in the United States is vital in consideration of 10CFR50.59 and other licensing and safety issues. RELAP5 and its force post-processor R5FORCE are used to model the transient loads for various piping systems such as main feedwater and blowdown systems. Other modeling applications, including the Bechtel GAFT program, are used to evaluate loadings in the main steam piping. Forces are calculated for main steam turbine stop valve closure, feedwater pipe breaks and subsequent check valve slam, and blowdown isolation valve closure. These PUR/SGR forces are compared with the original design basis forces. Modeling techniques discussed include proper valve closure modeling, sonic velocity changes due to pipe material changes, and two phase flow effects. Lessons learned based on analyses done for several PWR PUR with SGR are presented. Lessons learned from these analyses include choosing the optimal replacement piping size and routing to improve system performance without resulting in excessive piping loads. (authors)
Yu, Yi-Lin; Lee, Meei-Shyuan; Juan, Chun-Jung; Hueng, Dueng-Yuan
2013-08-01
The ABC/2 equation is commonly applied to measure the volume of intracranial hematoma. However, the precision of ABC/2 equation in estimating the tumor volume of acoustic neuromas is less addressed. The study is to evaluate the accuracy of the ABC/2 formula by comparing with planimetry method for estimating the tumor volumes. Thirty-two patients diagnosed with acoustic neuroma received contrast-enhanced magnetic resonance imaging of brain were recruited. The volume was calculated by the ABC/2 equation and planimetry method (defined as exact volume) at the same time. The 32 patients were divided into three groups by tumor volume to avoid volume-dependent overestimation (6 ml). The tumor volume by ABC/2 method was highly correlated to that calculated by planimetry method using linear regression analysis (R2=0.985). Pearson correlation coefficient (r=0.993, pABC/2 formula is an easy method in estimating the tumor volume of acoustic neuromas that is not inferior to planimetry method. Copyright © 2013 Elsevier B.V. All rights reserved.
Factors affecting volume calculation with single photon emission tomography (SPECT) method
International Nuclear Information System (INIS)
Liu, T.H.; Lee, K.H.; Chen, D.C.P.; Ballard, S.; Siegel, M.E.
1985-01-01
Several factors may influence the calculation of absolute volumes (VL) from SPECT images. The effect of these factors must be established to optimize the technique. The authors investigated the following on the VL calculations: % of background (BG) subtraction, reconstruction filters, sample activity, angular sampling and edge detection methods. Transaxial images of a liver-trunk phantom filled with Tc-99m from 1 to 3 μCi/cc were obtained in 64x64 matrix with a Siemens Rota Camera and MDS computer. Different reconstruction filters including Hanning 20,32, 64 and Butterworth 20, 32 were used. Angular samplings were performed in 3 and 6 degree increments. ROI's were drawn manually and with an automatic edge detection program around the image after BG subtraction. VL's were calculated by multiplying the number of pixels within the ROI by the slice thickness and the x- and y- calibrations of each pixel. One or 2 pixel per slice thickness was applied in the calculation. An inverse correlation was found between the calculated VL and the % of BG subtraction (r=0.99 for 1,2,3 μCi/cc activity). Based on the authors' linear regression analysis, the correct liver VL was measured with about 53% BG subtraction. The reconstruction filters, slice thickness and angular sampling had only minor effects on the calculated phantom volumes. Detection of the ROI automatically by the computer was not as accurate as the manual method. The authors conclude that the % of BG subtraction appears to be the most important factor affecting the VL calculation. With good quality control and appropriate reconstruction factors, correct VL calculations can be achieved with SPECT
ANSPipe: An IBM-PC interactive code for pipe-break assessment
International Nuclear Information System (INIS)
Fullwood, R.R.; Harrington, M.
1988-01-01
The advanced neutron source (ANS) being designed at Oak Ridge National Laboratory will be the world's highest flux neutron source and best facility for associated basic and applied research. The ANSPipe code was written as an aid for the piping configuration and material selection to enhance safety and availability. The primary calculation is based on the Thomas mode. which models pipe leak or break probabilities as proportional to the length of the segment and diameter and the inverse square of the wall thickness. This scaling, based on experience, is adjusted for radiation effects, using the Regulatory Guide 1.99 model, and for cyclic fatigue, stress corrosion, and inspection, using adaptations form the PRAISE-B code. The key to an ANSPipe analysis is the definition of the pipe segments. A pipe segment is defined as a length of pipe in which all the parameters affecting the pipe are constant or reasonably so. Thus, a segment would be a length of pipe of constant diameter, thickness, material type, internal pressure, flux distribution, stress, and submergence or nonsubmergence
Balance-of-plant options for the Heat-Pipe Power System
International Nuclear Information System (INIS)
Berte, M.; Capell, B.
1997-09-01
The Heat-Pipe Power System (HPS) is a near-term, low-cost space fission power system with the potential for utilizing various option for balance-of-plant options. The following options have been studied: a low-power thermoelectric design (14-kWe output), a small Brayton cycle system (60--75 kWe), and a large Brayton cycle system (250 kWe). These systems were analyzed on a preliminary basis, including mass, volume, and structure calculations. These analyses have shown that the HPS system can provide power outputs from 10--250 kWe with specific powers of ∼ 14 W/kg for a 14-kWe model to ∼ 100 W/kg for a 250-kWe model. The system designs considered in this study utilize a common component base to permit easy expansion and development
Waste retrieval sluicing system campaign number 1 solids volume transferred calculation
International Nuclear Information System (INIS)
BAILEY, J.W.
1999-01-01
This calculation has been prepared to document the volume of sludge removed from tank 241-C-106 during Waste Retrieval Sluicing System (WRSS) Sluicing Campaign No.1. This calculation will be updated, if necessary, to incorporate new data. This calculation supports the declaration of completion of WRSS Campaign No.1 and, as such, is also the documentation for completion of Performance Agreement TWR 1.2.1 , C-106 Sluicing Performance Expectations. It documents the performance of all the appropriate tank 241-C-106 mass transfer verifications, evaluations, and appropriate adjustments discussed in HNF-SD-WM-PROC-021, Chapter 23, ''Process Engineering Calculations for Tank 241-C-106 Sluicing and Retrieval''
Waste retrieval sluicing system campaign number 1 solids volume transferred calculation
International Nuclear Information System (INIS)
BAILEY, J.W.
1999-01-01
This calculation has been prepared to document the volume of sludge removed from tank 241-C-106 during Waste Retrieval Sluicing System (WRSS) Sluicing Campaign No.1. This calculation will be updated, if necessary, to incorporate new data. This calculation supports the declaration of completion of WRSS Campaign No.1 and, as such, is also the documentation for completion of Performance Agreement TWR 1.2.1 C-106 Sluicing Performance Expectations. It documents the performance of all the appropriate tank 241-C-106 mass transfer verifications, evaluations, and appropriate adjustments discussed in HNF-SD-WM-PROC-021, Chapter 23, ''Process Engineering Calculations for Tank 241-C-106 Sluicing and Retrieval''
Joustra, S.D.; Plas, E.M. van der; Goede, J.; Oostdijk, W.; Delemarre-van de Waal, H.A.; Hack, W.W.M.; Buuren, S. van; Wit, J.M.
2015-01-01
Aim Accurate calculations of testicular volume standard deviation (SD) scores are not currently available. We constructed LMS-smoothed age-reference charts for testicular volume in healthy boys. Methods The LMS method was used to calculate reference data, based on testicular volumes from
Dynamic response of piping system subject to flow acoustic excitation
International Nuclear Information System (INIS)
Wang, T.; Sun, Y.S.
1988-01-01
Through the use of a theoretically derived and test data-calibrated forcing function, the dynamic response of a piping system subject to flow-acoustic induced vibration is analyzed. It is shown that the piping behavior can be predicted when consideration is given to both the wall flexural vibration and the piping system vibration. Piping responded as a system to the transversal excitation due to the swirling motion of the fluid flow, as well as flexurally to the high-frequency acoustic excitations. The transverse piping system response was calculated using a lumped mass piping model. The piping model has more stringent requirements than its counterpart for waterhammer and seismic modeling due to the shorter spiral wavelength and higher frequency of the forcing function. Proper modeling ensured that both the moment stress caused by system excitation and the local stress induced by the support reaction load were properly accounted for. Flexural vibration not only poses a threat to nipples and branch connections, but also contributes substantially to the resultant total stress experienced by the pipe. The forcing function approach has the advantage that the critical locations on the piping system can be identified by means of analysis, facilitating surveillance and inspection, as well as fatigue evaluation
TSTA piping and flame arrestor operating experience data
Energy Technology Data Exchange (ETDEWEB)
Cadwallader, Lee C., E-mail: Lee.Cadwallader@inl.gov [Idaho National Laboratory, Idaho Falls, ID (United States); Willms, R. Scott [ITER International Organization, Cadarache (France)
2015-10-15
Highlights: • Experiences from the Tritium Systems Test Assembly were examined. • Failure rates of copper piping and a flame arrestor were calculated. • The calculated failure rates compared well to similar data from the literature. • Tritium component failure rate data support fusion safety assessment. - Abstract: The Tritium Systems Test Assembly (TSTA) was a facility dedicated to tritium handling technology and experiment research at the Los Alamos National Laboratory. The facility was operated with tritium for its research and development program from 1984 to 2001, running a prototype fusion fuel processing loop with ∼100 g of tritium as well as small experiments. There have been several operating experience reports written on this facility's operation and maintenance experience. This paper describes reliability analysis of two additional components from TSTA, small diameter copper gas piping that handled tritium in a nitrogen carrier gas, and the flame arrestor used in this piping system. The component failure rates for these components are discussed in this paper. Comparison data from other applications are also presented.
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
40 CFR 80.599 - How do I calculate volume balances for designation purposes?
2010-07-01
... (CONTINUED) AIR PROGRAMS (CONTINUED) REGULATION OF FUELS AND FUEL ADDITIVES Motor Vehicle Diesel Fuel... June 30, 2013. July 1, 2013 May 31, 2014. (2) [Reserved] (b) Volume balance for motor vehicle diesel fuel. (1) A facility's motor vehicle diesel fuel volume balance is calculated as follows: MVB = MVI−MVO...
Heat losses through pipe connections in hot water stores
DEFF Research Database (Denmark)
Andersen, Elsa; Fan, Jianhua; Furbo, Simon
2007-01-01
The heat loss from pipe connections at the top of hot water storage tanks with and without a heat trap is investigated theoretically and compared to similar experimental investigations. Computational Fluid Dynamics (CFD) is used for the theoretical analysis. The investigations show that the heat...... loss from an ideally insulated pipe connected to the top of a hot water tank is mainly due to a natural convection flow in the pipe, that the heat loss coefficient of pipes connected to the top of a hot water tank is high, and that a heat trap can reduce the heat loss coefficient significantly. Further......, calculations show that the yearly thermal performance of solar domestic hot water systems is strongly reduced if the hot water tank has a thermal bridge located at the top of the tank....
Risk analysis of in-service pressure piping containing defects
International Nuclear Information System (INIS)
Lin, Y.C.; Xie, Y.J.; Wang, X.H.; Luo, H.
2004-01-01
The reliability of pressure piping containing defects is important in engineering. The failure probability of pressure piping containing defects may be used as a guide to the most economic deployment of resources on maintenance, inspection and repair. This paper presents a probabilistic assessment methodology for in-service pressure piping containing defects, which is especially designed for programming. It is based on three assessment codes, BS 7910, R6 and SAPV-99, considering uncertainties in operating loadings, flaw sizes, material fracture toughness and flow stress. A general sampling computation method of stress intensity factor (SIF), in the form of the relationship between SIF and axial force and bending moment and torsion, is adopted. This relationship has been successfully used in developing software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), to assess planar and non-planar flaws. A numerical example is presented to illustrate the application of SAPP-2003 for calculating the failure probabilities of separate defects and for the assessed pressure piping
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)
Comparison of leak opening and leak rate calculations to HDR experimental results
International Nuclear Information System (INIS)
Grebner, H.; Hoefler, A.; Hunger, H.
1993-01-01
During the last years a number of calculations of leak opening and leak rate for through cracks in piping components have been performed. Analyses are pre- or mostly post-calculations to experiments performed at the HDR facility under PWR operating conditions. Piping components under consideration were small diameter straight pipes with circumferential cracks, pipe bends with longitudinal or circumferential cracks and pipe branches with weldment cracks. The components were loaded by internal pressure and opening as well as closing bending moment. The finite element method and two-phase flow leak rate programs were 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
International Nuclear Information System (INIS)
Fondeur, F; Michael Poirier, M; Samuel Fink, S
2007-01-01
Effluent streams from the Modular Caustic-Side Solvent Extraction Unit (MCU) will transfer to the tank farms and to the Defense Waste Processing Facility (DWPF). These streams will contain entrained solvent. A significant portion of the Strip Effluent (SE) pipeline (i.e., acid stream containing Isopar(reg s ign) L residues) length is within one inch of a sludge stream. Personnel envisioned the sludge stream temperature may reach 100 C during operation. The nearby SE stream may receive heat from the sludge stream and reach temperatures that may lead to flammability issues once the contents of the SE stream discharge into a larger reservoir. To this end, personnel used correlations from the literature to estimate the maximum temperature rise the SE stream may experience if the nearby sludge stream reaches boiling temperature. Several calculation methods were used to determine the temperature rise of the SE stream. One method considered a heat balance equation under steady state that employed correlation functions to estimate heat transfer rate. This method showed the maximum temperature of the acid stream (SE) may exceed 45 C when the nearby sludge stream is 80 C or higher. A second method used an effectiveness calculation used to predict the heat transfer rate in single pass heat exchanger. By envisioning the acid and sludge pipes as a parallel flow pipe-to-pipe heat exchanger, this method provides a conservative estimation of the maximum temperature rise. Assuming the contact area (i.e., the area over which the heat transfer occurs) is the whole pipe area, the results found by this method nearly matched the results found with the previous calculation method. It is recommended that the sludge stream be maintained below 80 C to minimize a flammable vapor hazard from occurring
The assessment of water loss from a damaged distribution pipe using the FEFLOW software
Directory of Open Access Journals (Sweden)
Iwanek Małgorzata
2017-01-01
Full Text Available Common reasons of real water loss in distribution systems are leakages caused by the failures or pipe breakages. Depending on the intensity of leakage from a damaged buried pipe, water can flow to the soil surface just after the failure occurs, much later or never at all. The localization of the place where the pipe breakage occurs is relatively easy when water outflow occurs on the soil surface. The volume of lost water strongly depends on the time it takes to localize the place of a pipe breakage. The aim of this paper was to predict the volume of water lost between the moment of a failure occurring and the moment of water outflow on the soil surface, during a prospective failure in a distribution system. The basis of the analysis was a numerical simulation of a water pipe failure using the FEFLOW v. 5.3 software (Finite Element subsurface FLOW systems for a real middle-sized distribution system. Simulations were conducted for variants depending on pipes’ diameter (80÷200 mm for minimal and maximal hydraulic pressure head in the system (20.14 and 60.41 m H2O, respectively. FEFLOW software application enabled to select places in the water system where possible failures would be difficult to detect.
Enhanced seismic criteria for piping
International Nuclear Information System (INIS)
Touboul, F. . E-mail francoise.touboul@cea.fr; Blay, N.; Sollogoub, P.; Chapuliot, S.
2006-01-01
In situ or laboratory experiments have shown that piping systems exhibit satisfactory seismic behavior. Seismic motion is not severe enough to significantly damage piping systems unless large differential motions of anchorage are imposed. Nevertheless, present design criteria for piping are very severe and require a large number of supports, which creates overly rigid piping systems. CEA, in collaboration with EDF, FRAMATOME and IRSN, has launched a large R and D program on enhanced design methods which will be less severe, but still conservative, and compatible with defect justification during operation. This paper presents the background of the R and D work on this matter, and CEA proposed equations. Our approach is based on the difference between the real behavior (or the best estimated computed one) with the one supposed by codified methods. Codified criteria are applied on an elastically calculated behavior that can be significantly different from the real one: the effect of plasticity may be very meaningful, even with low incursion in the plastic domain. Moreover, and particularly in piping systems, the elastic follow-up effect affects stress distribution for both seismic and thermal loads. For seismic load, we have proposed to modify the elastic moment limitation, based on the interpretation of experimental results on piping systems. The methods have been validated on more industrial cases, and some of the consequences of the changes have been studied: modification of the drawings and of the number of supports, global displacements, forces in the supports, stability of potential defects, etc. The basic aim of the studies undertaken is to make a decision on the stress classification problem, one that is not limited to seismic induced stresses, and to propose simplified methods for its solution
Piriev, S. A.
2018-01-01
This paper describes the study of scattered fracture of a thick-walled pipe filled with an aggressive medium, which creates uniform pressure on the inner surface of the pipe. It is assumed that the aggressive medium affects only the value of instantaneous strength. Damageability is described by an integral operator of the hereditary type. The problem is solved with allowance for residual strength of the pipe material behind the fracture front. Numerical calculation is carried out, and relationships between the fracture front coordinate and time for various concentrations of the aggressive medium and residual strength behind the fracture front are constructed.
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
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)
Commercial high efficiency dehumidification systems using heat pipes
Energy Technology Data Exchange (ETDEWEB)
1993-09-01
An improved heat pipe design using separately connected two-section one-way flow heat pipes with internal microgrooves instead of wicks is described. This design is now commercially available for use to increase the dehumidification capacity of air conditioning systems. The design also includes a method of introducing fresh air into buildings while recovering heat and controlling the humidity of the incoming air. Included are applications and case studies, load calculations and technical data, and installation, operation, and maintenance information.
Piping structural design for the ITER thermal shield manifold
Energy Technology Data Exchange (ETDEWEB)
Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)
2015-10-15
Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.
Effect of inlet cone pipe angle in catalytic converter
Amira Zainal, Nurul; Farhain Azmi, Ezzatul; Arifin Samad, Mohd
2018-03-01
The catalytic converter shows significant consequence to improve the performance of the vehicle start from it launched into production. Nowadays, the geometric design of the catalytic converter has become critical to avoid the behavior of backpressure in the exhaust system. The backpressure essentially reduced the performance of vehicles and increased the fuel consumption gradually. Consequently, this study aims to design various models of catalytic converter and optimize the volume of fluid flow inside the catalytic converter by changing the inlet cone pipe angles. Three different geometry angles of the inlet cone pipe of the catalytic converter were assessed. The model is simulated in Solidworks software to determine the optimum geometric design of the catalytic converter. The result showed that by decreasing the divergence angle of inlet cone pipe will upsurge the performance of the catalytic converter.
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
Friction factors referring to laminar flow through pipe bundles with longitudinal webs
Energy Technology Data Exchange (ETDEWEB)
Schenkel, G
1983-09-01
Pipe bundles with continuous webs or ribs between adjacent pipes, as well as between outer pipes and channel walls, are much more vibrational proof than web-free systems. In addition, the change-over from a multiple-connected web-free cross-section to a set of singly-connected cross-sections facilitates the calculation of friction factors. The investigation is concerned with isothermal steady fully-developed laminar flow of Newtonian fluids. In particularly, pipe bundles with squares and hexagonal arrays in respective channels are treated. Friction factors for the subchannels are taken from a former paper of the author.
Failure frequencies and probabilities applicable to BWR and PWR piping
International Nuclear Information System (INIS)
Bush, S.H.; Chockie, A.D.
1996-03-01
This report deals with failure probabilities and failure frequencies of nuclear plant piping and the failure frequencies of flanges and bellows. Piping failure probabilities are derived from Piping Reliability Analysis Including Seismic Events (PRAISE) computer code calculations based on fatigue and intergranular stress corrosion as failure mechanisms. Values for both failure probabilities and failure frequencies are cited from several sources to yield a better evaluation of the spread in mean and median values as well as the widths of the uncertainty bands. A general conclusion is that the numbers from WASH-1400 often used in PRAs are unduly conservative. Failure frequencies for both leaks and large breaks tend to be higher than would be calculated using the failure probabilities, primarily because the frequencies are based on a relatively small number of operating years. Also, failure probabilities are substantially lower because of the probability distributions used in PRAISE calculations. A general conclusion is that large LOCA probability values calculated using PRAISE will be quite small, on the order of less than 1E-8 per year (<1E-8/year). The values in this report should be recognized as having inherent limitations and should be considered as estimates and not absolute values. 24 refs 24 refs
Method and device for characterization of two-phase flow in pipes
International Nuclear Information System (INIS)
Skarsvaag, K.; Sunde, A.J.
1993-01-01
Gamma radiation transmission measurements are made with one-shot-collimation to determine the distribution of voids within a gas-liquid mixture flowing in a pipe. The distribution of voids in selected portions of the pipe, taken together with statistical and logical tests applied thereto, provides information from which are determined: type of flow pattern or flow regime, the profile of a large gas bubble in slug flow, and the gas and the liquid volume flow rates in slug flow. 4 refs
Strain Limits within the Scope of the Integrity Assessment of Piping Systems
International Nuclear Information System (INIS)
Mutz, Alexander
2008-01-01
Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)
Strain Limits within the Scope of the Integrity Assessment of Piping Systems
Energy Technology Data Exchange (ETDEWEB)
Mutz, Alexander [EnBW, Durlacher Allee 93, Karlsruhe 76131 (Germany)
2008-07-01
Allowable stresses in nuclear power plant piping resulting from loading conditions to be considered in Germany are determined on the basis of the German Safety Standards of the Nuclear Safety Standards Commission, KTA. The limitation of the different stress categories within the analysis of the mechanical behaviour is based on a linear elastic material behaviour. Because of the ductile material used in high energy nuclear piping, a more realistic assessment can be performed on the basis of allowable strains using elastic plastic material behaviour. In the present work comparison between the analysis of piping systems considering the elastic material model and the actual elastic plastic material behaviour is performed. The possibilities of allocating plastic strains to calculated elastic stresses is discussed. A parametric study on straight pipes with the actual elastic plastic material model under pure bending is the basis of deriving the elastic plastic strains for the calculated elastic stresses. Strain limits are suggested which correspond to the different stress categories. The aim is to utilize the deformation possibilities of ductile materials used in German nuclear piping and the allocation of maximum strains to different load categories. Keywords: strain limit, ductile material, stress category. (author)
Russian regulatory approaches to seismic design and seismic analysis of NPP piping
International Nuclear Information System (INIS)
Kaliberda, Y.V.
2003-01-01
The paper presents an overview of Russian regulatory approaches to seismic design and seismic analysis of NPP piping. The paper is focused on categorization and seismic analysis of nuclear power plant items (piping, equipment, supports, valves, but not building structures). The paper outlines the current seismic recommendations, corresponding methods with the examples of calculation models. The paper considers calculation results of the mechanisms of dynamic behavior and the problems of developing a rational and economical approaches to seismic design and seismic protection. (author)
Heat pipes. Design and industrial applications
International Nuclear Information System (INIS)
Semeria, R.
1974-01-01
Heat pipes are thermosiphons with vaporization where we can distinguish a boiler, a condenser, and eventually an adiabatic zone. To insure the returning liquid flow from the condenser to the boiler, surface tension forces, associated with the gravity forces, if need be, are used. For this, the condensing liquid is sucked by a capillary structure, generally situated against the inner wall. The review of the design methods, and particularly the prediction of the maximal performances shows the advantages and limitations of such devices. The main difficulties are technological for the heat pipes with high temperature liquid metals. The thermohydrodynamical limitations are: the maximum power which can be calculated by a balance between the friction forces and the active ones, the maximum heat flux leading to the dry-out of the evaporator, the critical conditions for the start up associated with the sonic conditions in the vapour phase. The description of heat pipes designed for some industrial applications (mainly for space) is given [fr
Numerical studies of temperature effect on the extrusion fracture and swell of plastic micro-pipe
Ren, Zhong; Huang, Xingyuan; Xiong, Zhihua
2018-03-01
Temperature is a key factor that impacts extrusion forming quality of plastic micro-pipe. In this study, the effect of temperature on extrusion fracture and swell of plastic micro-pipe was investigated by numerical method. Under a certain of the melt’s flow volume, the extrusion pattern, extrusion swelling ratio of melt are obtained under different temperatures. Results show that the extrusion swelling ratio of plastic micro-pipe decreases with increasing of temperature. In order to study the reason of temperature effect, the physical distributions of plastic micro-pipe are gotten. Numerical results show that the viscosity, pressure, stress value of melt are all decreased with the increasing of temperature, which leads to decrease the extrusion swell and fracture phenomenon for the plastic micro-pipe.
Steel Fibers Reinforced Concrete Pipes - Experimental Tests and Numerical Simulation
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
Directory of Open Access Journals (Sweden)
Chao Hu
2015-04-01
Full Text Available Slope excavation is one of the most crucial steps in the construction of a hydraulic project. Excavation project quality assessment and excavated volume calculation are critical in construction management. The positioning of excavation projects using traditional instruments is inefficient and may cause error. To improve the efficiency and precision of calculation and assessment, three-dimensional laser scanning technology was used for slope excavation quality assessment. An efficient data acquisition, processing, and management workflow was presented in this study. Based on the quality control indices, including the average gradient, slope toe elevation, and overbreak and underbreak, cross-sectional quality assessment and holistic quality assessment methods were proposed to assess the slope excavation quality with laser-scanned data. An algorithm was also presented to calculate the excavated volume with laser-scanned data. A field application and a laboratory experiment were carried out to verify the feasibility of these methods for excavation quality assessment and excavated volume calculation. The results show that the quality assessment indices can be obtained rapidly and accurately with design parameters and scanned data, and the results of holistic quality assessment are consistent with those of cross-sectional quality assessment. In addition, the time consumption in excavation quality assessment with the laser scanning technology can be reduced by 70%–90%, as compared with the traditional method. The excavated volume calculated with the scanned data only slightly differs from measured data, demonstrating the applicability of the excavated volume calculation method presented in this study.
The measurement of the dielectric constant of concrete pipes and clay pipes
McGraw, David
To optimize the effectiveness of the rehabilitation of underground utilities, taking in consideration limitation of available resources, there is a need for a cost effective and efficient sensing systems capable of providing effective, in real time and in situ, measurement of infrastructural characteristics. To carry out accurate non-destructive condition assessment of buried and above ground infrastructure such as sewers, bridges, pavements and dams, an advanced ultra-wideband (UWB) based radar was developed at Trenchless Technology Centre (TTC) and Centre for Applied Physics Studies (CAPS) at Louisiana Tech University (LTU). One of the major issues in designing the FCC compliant UWB radar was the contribution of the pipe wall, presence of complex soil types and moderate-to-high moisture levels on penetration depth of the electromagnetic (EM) energy. The electrical properties of the materials involved in designing the UWB radar exhibit a significant variation as a result of the moisture content, mineral content, bulk density, temperature and frequency of the electromagnetic signal propagating through it. Since no measurements of frequency dependence of the dielectric permittivity and conductivities of the pipe wall material in the FCC approved frequency range exist, in this thesis, the dielectric constant of concrete and clay pipes are measured over a microwave frequency range from 1 Ghz to 10 Ghz including the effects of moisture and chloride content. A high performance software package called MU-EPSLN(TM) was used for the calculations. Data reduction routines to calculate the complex permeability and permittivity of materials as well as other parameters are also provided. The results obtained in this work will be used to improve the accuracy of the numerical simulations and the performances of the UWB radar system.
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...
Soil map, area and volume calculations in Orrmyrberget catchment basin at Gideaa, Northern Sweden
International Nuclear Information System (INIS)
Ittner, T.; Tammela, P.T.; Gustafsson, E.
1991-06-01
Fallout studies in the Gideaa study site after the Chernobyl fallout in 1986, has come to the point that a more exact surface mapping of the studied catchment basin is needed. This surface mapping is mainly made for area calculations of different soil types within the study site. The mapping focus on the surface, as the study concerns fallout redistribution and it is extended to also include materials down to a depth of 0.5 meter. Volume calculations are made for the various soil materials within the top 0.5 m. These volume and area calculations will then be used in the modelling of the migration and redistribution of the fallout radionuclides within the studied catchment basin. (au)
46 CFR 197.462 - Pressure vessels and pressure piping.
2010-10-01
... that each pressure vessel, including each volume tank, cylinder and PVHO, and each pressure piping... tests conducted in accordance with this section shall be either hydrostatic tests or pneumatic tests. (1... times the maximum allowable working pressure. (2) When a pneumatic test is conducted on a pressure...
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
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
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.
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
Energy Technology Data Exchange (ETDEWEB)
None
1979-09-01
A design study for a central receiver/fossil fuel hybrid power system using molten salts for heat transfer and heat storage is presented. This volume contains the appendices: (A) parametric salt piping data; (B) sample heat exchanger calculations; (C) salt chemistry and salt/materials compatibility evaluation; (D) heliostat field coordinates; (E) data lists; (F) STEAEC program input data; (G) hybrid receiver design drawings; (H) hybrid receiver absorber tube thermal math model; (I) piping stress analysis; (J) 100-MWe 18-hour storage solar central receiver hybrid power system capital cost worksheets; and (K) 500-MWe 18-hour solar central receiver hybrid power system cost breakdown. (WHK)
Probabilistic assessment of critically flawed LMFBR PHTS piping elbows
International Nuclear Information System (INIS)
Balkey, K.R.; Wallace, I.T.; Vaurio, J.K.
1982-01-01
One of the important functions of the Primary Heat Transport System (PHTS) of a large Liquid Metal Fast Breeder Reactor (LMFBR) plant is to contain the circulating radioactive sodium in components and piping routed through inerted areas within the containment building. A significant possible failure mode of this vital system is the development of cracks in the piping components. This paper presents results from the probabilistic assessment of postulated flaws in the most-critical piping elbow of each piping leg. The criticality of calculated maximum sized flaws is assessed against an estimated material fracture toughness to determine safety factors and failure probability estimates using stress-strength interference theory. Subsequently, a different approach is also employed in which the randomness of the initial flaw size and loading are more-rigorously taken into account. This latter approach yields much smaller probability of failure values when compared to the stress-strength interference analysis results
Analysis of the thermal performance of heat pipe radiators
Boo, J. H.; Hartley, J. G.
1990-01-01
A comprehensive mathematical model and computational methodology are presented to obtain numerical solutions for the transient behavior of a heat pipe radiator in a space environment. The modeling is focused on a typical radiator panel having a long heat pipe at the center and two extended surfaces attached to opposing sides of the heat pipe shell in the condenser section. In the set of governing equations developed for the model, each region of the heat pipe - shell, liquid, and vapor - is thermally lumped to the extent possible, while the fin is lumped only in the direction normal to its surface. Convection is considered to be the only significant heat transfer mode in the vapor, and the evaporation and condensation velocity at the liquid-vapor interface is calculated from kinetic theory. A finite-difference numerical technique is used to predict the transient behavior of the entire radiator in response to changing loads.
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
International Nuclear Information System (INIS)
Thompson, W.L.; Deutsch, O.L.; Booth, T.E.
1980-04-01
Several Monte Carlo techniques are compared in the transport of neutrons of different source energies through two different deep-penetration problems each with two parts. The first problem involves transmission through a 200-cm concrete slab. The second problem is a 90 0 bent pipe jacketed by concrete. In one case the pipe is void, and in the other it is filled with liquid sodium. Calculations are made with two different Los Alamos Monte Carlo codes: the continuous-energy code MCNP and the multigroup code MCMG
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.)
Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe
International Nuclear Information System (INIS)
Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol
2015-01-01
Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of
Nonlinear optimal perturbations in a curved pipe
Rinaldi, Enrico; Canton, Jacopo; Marin, Oana; Schanen, Michel; Schlatter, Philipp
2017-11-01
We investigate the effect of curvature on transition to turbulence in pipes by comparing optimal perturbations of finite amplitude that maximise their energy growth in a toroidal geometry to the ones calculated in the absence of curvature. Our interest is motivated by the fact that even small curvatures, of the order of d =Rpipe /Rtorus art numerical algorithms, capable of tackling the optimisation problem on large computational domains, coupled to a high-order spectral-element code, which is used to perform direct numerical simulations (DNS) of the full Navier-Stokes and their adjoint equations. Results are compared to the corresponding states in straight pipes and differences in their structure and evolution are discussed. Furthermore, the newly calculated initial conditions are used to identify coherent flow structures that are compared to the ones observed in recent DNS of weakly turbulent and relaminarising flows in the same toroidal geometry.
Experiments on vertical gas-liquid pipe flows using ultrafast X-ray tomography
Energy Technology Data Exchange (ETDEWEB)
Banowski, M.; Beyer, M.; Lucas, D.; Hoppe, D.; Barthel, F. [Helmholtz-Zentrum Dresden-Rossendorf (Germany). Inst. fuer Sicherheitsforschung
2016-12-15
For the qualification and validation of two-phase CFD-models for medium and large-scale industrial applications dedicated experiments providing data with high temporal and spatial resolution are required. Fluid dynamic parameter like gas volume fraction, bubble size distribution, velocity or turbulent kinetic energy should be measured locally. Considering the fact, that the used measurement techniques should not affect the flow characteristics, radiation based tomographic methods are the favourite candidate for such measurements. Here the recently developed ultrafast X-ray tomography, is applied to measure the local and temporal gas volume fraction distribution in a vertical pipe. To obtain the required frame rate a rotating X-ray source by a massless electron beam and a static detector ring are used. Experiments on a vertical pipe are well suited for development and validation of closure models for two-phase flows. While vertical pipe flows are axially symmetrically, the boundary conditions are well defined. The evolution of the flow along the pipe can be investigated as well. This report documents the experiments done for co-current upwards and downwards air-water and steam-water flows as well as for counter-current air-water flows. The details of the setup, measuring technique and data evaluation are given. The report also includes a discussion on selected results obtained and on uncertainties.
International Nuclear Information System (INIS)
Meier, S.; Mildenberger, P.; Pitton, M.; Thelen, M.; Schenk, A.; Bourquain, H.
2004-01-01
Purpose: computed tomography has become the preferred method in detecting liver carcinomas. The introduction of spiral CT added volumetric assessment of intrahepatic tumors, which was unattainable in the clinical routine with incremental CT due to complex planimetric revisions and excessive computing time. In an ongoing clinical study, a new software tool was tested for the automatic detection of tumor volume and the time needed for this procedure. Materials and methods: we analyzed patients suffering from hepatocellular carcinoma (HCC). All patients underwent treatment with repeated transcatheter chemoembolization of the hepatic arteria. The volumes of the HCC lesions detected in CT were measured with the new software tool in HepaVison (MeVis, Germany). The results were compared with manual planimetric calculation of the volume performed by three independent radiologists. Results: our first results in 16 patients show a correlation between the automatically and the manually calculated volumes (up to a difference of 2 ml) of 96.8%. While the manual method of analyzing the volume of a lesion requires 2.5 minutes on average, the automatic method merely requires about 30 seconds of user interaction time. Conclusion: These preliminary results show a good correlation between automatic and manual calculations of the tumor volume. The new software tool requires less time for accurate determination of the tumor volume and can be applied in the daily clinical routine. (orig.) [de
Residual stress measurement in 304 stainless steel weld overlay pipes
International Nuclear Information System (INIS)
Yen, H.J.; Lin, M.C.C.; Chen, L.J.
1996-01-01
Welding overlay repair (WOR) is commonly employed to rebuild piping systems suffering from intergranular stress corrosion cracking (IGSCC). To understand the effects of this repair, it is necessary to investigate the distribution of residual stresses in the welding pipe. The overlay welding technique must induce compressive residual stress at the inner surface of the welded pipe to prevent IGSCC. To understand the bulk residual stress distribution, the stress profile as a function of location within wall is examined. In this study the full destructive residual stress measurement technique -- a cutting and sectioning method -- is used to determine the residual stress distribution. The sample is type 304 stainless steel weld overlay pipe with an outside diameter of 267 mm. A pipe segment is cut from the circular pipe; then a thin layer is removed axially from the inner to the outer surfaces until further sectioning is impractical. The total residual stress is calculated by adding the stress relieved by cutting the section away to the stress relieved by axially sectioning. The axial and hoop residual stresses are compressive at the inner surface of the weld overlay pipe. Compressive stress exists not only at the surface but is also distributed over most of the pipe's cross section. On the one hand, the maximum compressive hoop residual stress appears at the pipe's inner surface. The thermal-mechanical induced crack closure from significant compressive residual stress is discussed. This crack closure can thus prevent IGSCC very effectively
Calculation of Steam Volume Fraction in Subcooled Boiling
Energy Technology Data Exchange (ETDEWEB)
Rouhani, S Z
1967-06-15
An analysis of subcooled boiling is presented. It is assumed that heat is removed by vapor generation, heating of the liquid that replaces the detached bubbles, and to some extent by single phase heat transfer. Two regions of subcooled boiling are considered and a criterion is provided for obtaining the limiting value of subcooling between the two regions. Condensation of vapor in the subcooled liquid is analysed and the relative velocity of vapor with respect to the liquid is neglected in these regions. The theoretical arguments result in some equations for the calculation of steam volume fraction and true liquid subcooling.
The method of calculation of pipelines laid on supports
Directory of Open Access Journals (Sweden)
Benin D.M.
2017-08-01
Full Text Available this article focuses on the issue of laying pipelines on supports and the method of calculation of vertical and horizontal loads acting on the support. As pipelines can be water piping systems, heat networks, oil and mazout lines, condensate lines, steam lines, etc. this article describes the calculations of supports for pipelines laid above ground, in crowded channels, premises, on racks, in impassable channels, hanging supports, etc. The paper explores recommendations for placement of the supports on the route of the pipelines, calculation of loads on rotating and stationary supports of pipelines; inspection of stresses in the metal pipe, resulting from elongation of the piping from the temperature from the thermal expansion of the metal during operation.
Energy Technology Data Exchange (ETDEWEB)
1985-04-01
This document reports the position and recommendations of the NRC Piping Review Committee, Task Group on Seismic Design. The Task Group considered overlapping conservation in the various steps of seismic design, the effects of using two levels of earthquake as a design criterion, and current industry practices. Issues such as damping values, spectra modification, multiple response spectra methods, nozzle and support design, design margins, inelastic piping response, and the use of snubbers are addressed. Effects of current regulatory requirements for piping design are evaluated, and recommendations for immediate licensing action, changes in existing requirements, and research programs are presented. Additional background information and suggestions given by consultants are also presented.
Development on methods for evaluating structure reliability of piping components
International Nuclear Information System (INIS)
Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.
2003-01-01
In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)
Seismic behaviour of un-cracked and cracked thin pipes
International Nuclear Information System (INIS)
Blay, N.; Brunet, G.; Gantenbein, F.; Aguilar, J.
1995-01-01
In order to evaluate the seismic behaviour of un-cracked and cracked thin pipes, subjected to high acceleration levels, seismic tests and calculations have been performed on straight thin pipes made of 316L stainless steel, loaded in pure bending by a permanent static and dynamic loading. The seismic tests were carried out on the AZALEE shaking table of the CEA laboratory TAMARIS. The influence of the elasto-plastic model with isotropic or kinematic hardening are studied. 5 refs., 7 figs., 2 tabs
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
International Nuclear Information System (INIS)
Clement, Jason; Wang Xia
2013-01-01
A pulsating heat pipe (PHP) is a closed loop, passive heat transfer device. Its operation depends on the phase change of a working fluid within the loop. Design and performance testing of a pulsating heat pipe was conducted under conditions to simulate heat dissipation requirements of a proton exchange membrane (PEM) fuel cell stack. Integration of pulsating heat pipes within bipolar plates of the stack would eliminate the need for ancillary cooling equipment, thus also reducing parasitic losses and increasing energy output. The PHP under investigation, having dimensions of 46.80 cm long and 14.70 cm wide, was constructed from 0.3175 cm copper tube. Heat pipes effectiveness was found to be dependent upon several factors such as energy input, types of working fluid and its filling ratio. Power inputs to the evaporator side of the pulsating heat pipe varied from 80 to 180 W. Working fluids tested included acetone, methanol, and deionized water. Filling ratios between 30 and 70 percent of the total working volume were also examined. Methanol outperformed other fluids tested; with a 45 percent fluid fill ratio and a 120 W power input, the apparatus took the shortest time to reach steady state and had one of the smallest steady state temperature differences. The various conditions studied were chosen to assess the heat pipe's potential as cooling media for PEM fuel cells. - Highlights: ► Methanol as a working fluid outperformed both acetone and water in a pulsating heat pipe. ► Performance for the PHP peaked with methanol and a fill ratio of 45 percent fluid to total volume. ► A smaller resistance was associated with a higher power input to the system.
Directory of Open Access Journals (Sweden)
Ramachandran Raghavan Nair
2016-01-01
Full Text Available Experiments were conducted to study the thermal performance of meshed wick heat pipe by varying the working fluid and heat input. In this work four screen mesh wicked heat pipes were fabricated and tested. All the heat pipes were tested for heat input from 50W to 250W each with an increment of 50W in each step. The heat input range selected in this study is commonly encountered in most of the electronic application devices. The thermal resistance of all the heat pipes charged with different working fluids such as DI water, Al2O3/DI water nanofluid of volume concentration 0.1 % and hybrid nanofluid volume concentration 0.1%( with two different combinations of (Al2O3 50%- CuO 50%/DI water and (Al2O3 25%- CuO 75%/DI waterwas determined. The maximum percentage reduction was found to be 58.87% for the hybrid nanofluid of (Al2O3 25%- CuO 75%/DI water compared to base fluid. An important observation from the study is that, use of hybrid nanofluid can raise the operating range of the heat pipe beyond 250W which makes hybrid nanofluid as a potential substitute for the conventional working fluid.
The effect of cyclic and dynamic loads on carbon steel pipe
International Nuclear Information System (INIS)
Rudland, D.L.; Scott, P.M.; Wilkowski, G.M.
1996-02-01
This report presents the results of four 152-mm (6-inch) diameter, unpressurized, circumferential through-wall-cracked, dynamic pipe experiments fabricated from STS410 carbon steel pipe manufactured in Japan. For three of these experiments, the through-wall crack was in the base metal. The displacement histories applied to these experiments were a quasi-static monotonic, dynamic monotonic, and dynamic, cyclic (R = -1) history. The through-wall crack for the third experiment was in a tungsten-inert-gas weld, fabricated in Japan, joining two lengths of STS410 pipe. The displacement history for this experiment was the same history applied to the dynamic, cyclic base metal experiment. The test temperature for each experiment was 300 C (572 F). The objective of these experiments was to compare a Japanese carbon steel pipe material with US pipe material, to ascertain whether this Japanese steel was as sensitive to dynamic and cyclic effects as US carbon steel pipe. In support of these pipe experiments, quasi-static and dynamic, tensile and fracture toughness tests were conducted. An analysis effort was performed that involved comparing experimental crack initiation and maximum moments with predictions based on available fracture prediction models, and calculating J-R curves for the pipe experiments using the η-factor method
Updated pipe break analysis for Advanced Neutron Source Reactor conceptual design
International Nuclear Information System (INIS)
Wendel, M.W.; Chen, N.C.J.; Yoder, G.L.
1994-01-01
The Advanced Neutron Source Reactor (ANSR) is a research reactor to be built at the Oak Ridge National Laboratory that will supply the highest continuous neutron flux levels of any reactor in the world. It uses plate-type fuel with high-mass-flux and highly subcooled heavy water as the primary coolant. The Conceptual Safety Analysis for the ANSR was completed in June 1992. The thermal-hydraulic pipe-break safety analysis (performed with a specialized version of RELAP5/MOD3) focused primarily on double-ended guillotine breaks of the primary piping and some core-damage mitigation options for such an event. Smaller, instantaneous pipe breaks in the cold- and hot-leg piping were also analyzed to a limited extent. Since the initial analysis for the conceptual design was completed, several important changes to the RELAP5 input model have been made reflecting improvements in the fuel grading and changes in the elevation of the primary coolant pumps. Also, a new philosophy for pipe-break safety analysis (similar to that adopted for the New Production Reactor) accentuates instantaneous, limited flow area pipe-break accidents in addition to finite-opening-time, double-ended guillotine breaks of the major coolant piping. This paper discloses the results of the most recent instantaneous pipe-break calculations
Pressure wave propagation in the discharge piping with water pool
International Nuclear Information System (INIS)
Bang, Young S.; Seul, Kwang W.; Kim, In Goo
2004-01-01
Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined
241-U-701 new compressor building and instrument air piping analyses
International Nuclear Information System (INIS)
Huang, F.H.
1994-01-01
Building anchorage analysis is performed to qualify the design of the new compressor building foundation given in the ECN ''241-U-701 New Compressor Building.'' Recommendations for some changes in the ECN are made accordingly. Calculations show that the 6-in.-slab is capable of supporting the pipe supports, and that the building foundation, air compressor and dryer anchorage, and electric rack are adequate structurally. Analysis also shows that the instrument air piping and pipe supports for the compressed air system meet the applicable code requirements and are acceptable. The building is for the U-Farm instrument air systems
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
Impedance of a slotted-pipe kicker
Energy Technology Data Exchange (ETDEWEB)
Zhou, Feng [Academia Sinica, Beijing, BJ (China). Inst. of High Energy Physics
1996-08-01
This paper introduces the principle of a new slotted kicker simply, which is made by using vacuum pipe itself with proper slits as current conductors, and then, presents a rough estimation of its longitudinal and transverse impedance, respectively. Calculation shows that its impedance is reduced significantly compared to our present air-coil kicker. (author)
Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber
Energy Technology Data Exchange (ETDEWEB)
Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)
2015-05-15
Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by
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
Electron cloud density measurements in accelerator beam-pipe using resonant microwave excitation
Energy Technology Data Exchange (ETDEWEB)
Sikora, John P., E-mail: jps13@cornell.edu [CLASSE, Cornell University, Ithaca, NY 14853 (United States); Carlson, Benjamin T. [Carnegie Mellon University, Pittsburgh, PA 15213 (United States); Duggins, Danielle O. [Gordon College, Wenham, MA 01984 (United States); Hammond, Kenneth C. [Columbia University, New York, NY 10027 (United States); De Santis, Stefano [LBNL, Berkeley, CA 94720 (United States); Tencate, Alister J. [Idaho State University, Pocatello, ID 83209 (United States)
2014-08-01
An accelerator beam can generate low energy electrons in the beam-pipe, generally called electron cloud, that can produce instabilities in a positively charged beam. One method of measuring the electron cloud density is by coupling microwaves into and out of the beam-pipe and observing the response of the microwaves to the presence of the electron cloud. In the original technique, microwaves are transmitted through a section of beam-pipe and a change in EC density produces a change in the phase of the transmitted signal. This paper describes a variation on this technique in which the beam-pipe is resonantly excited with microwaves and the electron cloud density calculated from the change that it produces in the resonant frequency of the beam-pipe. The resonant technique has the advantage that measurements can be localized to sections of beam-pipe that are a meter or less in length with a greatly improved signal to noise ratio.
Directory of Open Access Journals (Sweden)
DONG Peng
2017-01-01
Full Text Available When one end of a fluid-filled pipe with an elastic wall is fixed and a harmonic force effect acts on the other end,a steady longitudinal vibration will be produced. Compared to the pipeline resonance mode,the amplitude of the steady longitudinal vibration of an elastic pipe is greater,and the effect on the sound is also greater. The study of the steady longitudinal vibration of pipes can better describe the effects of fluid-filled pipelines on the radiation sound field of the pipe opening. Through the contrast between the analysis calculation of the equivalent beam model and the experimental results,the accuracy of the equivalent beam model for the calculation of the steady longitudinal vibration of pipelines is verified,and a method of isolating the steady longitudinal vibration state is proposed and verified.
Finite volume thermal-hydraulics and neutronics coupled calculations - 15300
International Nuclear Information System (INIS)
Araujo Silva, V.; Campagnole dos Santos, A.A.; Mesquit, A.Z.; Bernal, A.; Miro, R.; Verdu, G.; Pereira, C.
2015-01-01
The computational power available nowadays allows the coupling of neutronics and thermal-hydraulics codes for reactor studies. The present methodology foresees at least one constraint to the separated codes in order to perform coupled calculations: both codes must use the same geometry, however, meshes can be different for each code as long as the internal surfaces stays the same. Using the finite volume technique, a 3D diffusion nodal code was implemented to deal with neutron transport. This code can handle non-structured meshes which allows for complicated geometries calculations and therefore more flexibility. A computational fluid dynamics (CFD) code was used in order to obtain the same level of details for the thermal hydraulics calculations. The chosen code is OpenFOAM, an open-source CFD tool. Changes in OpenFOAM allow simple coupled calculations of a PWR fuel rod with neutron transport code. OpenFOAM sends coolant density information and fuel temperature to the neutron transport code that sends back power information. A mapping function is used to average values when one node in one side corresponds to many nodes in the other side. Data is exchanged between codes by library calls. As the results of a fuel rod calculations progress, more complicated and processing demanding geometries will be simulated, aiming to the simulation of a real scale PWR fuel assembly
Magnetic forces on a ferromagnetic HT-9 first wall/blanket and coolant pipe
International Nuclear Information System (INIS)
Lechtenberg, T.A.; Dahms, C.; Attaya, H.; Univ. of Wisconsin, Madison)
1984-01-01
The GFUN 3D code was used to model the toroidal fields and determine the magnetic body forces on the STARFIRE design for coolant pipes exiting the first wall sector and first wall/blanket modules. The HT-9 coolant pipes were modeled on the basis of a square bar having the same length and material volume as the coolant pipes. The stress analysis was performed using these magnetic forces applied to a pipe of 4 meters length, 8.25 cm O.D., and 0.75 cm thickness by the MODSAP stress analysis code. For the first wall/blanket module, GFUN 3D does not allow full modeling of the complex thin-walled structure or numerous small tubes because of the element aspect ratio limitations. Therefore, to obtain three dimensional loads, a solid homogeneous equivalent structure was used
Treating network junctions in finite volume solution of transient gas flow models
Bermúdez, Alfredo; López, Xián; Vázquez-Cendón, M. Elena
2017-09-01
A finite volume scheme for the numerical solution of a non-isothermal non-adiabatic compressible flow model for gas transportation networks on non-flat topography is introduced. Unlike standard Euler equations, the model takes into account wall friction, variable height and heat transfer between the pipe and the environment which are source terms. The case of one single pipe was considered in a previous reference by the authors, [8], where a finite volume method with upwind discretization of the flux and source terms has been proposed in order to get a well-balanced scheme. The main goal of the present paper is to go a step further by considering a network of pipes. The main issue is the treatment of junctions for which container-like 2D finite volumes are introduced. The couplings between pipes (1D) and containers (2D) are carefully described and the conservation properties are analyzed. Numerical tests including real gas networks are solved showing the performance of the proposed methodology.
International Nuclear Information System (INIS)
Kondo, Koichi; Yoshida, Kenji; Okawa, Tomio; Kataoka, Isao
2004-01-01
Experiment and numerical calculation were carried out for upward, turbulent bubbly two-phase flow in a vertical pipe with an axisymmetric sudden expansion, which is one of the typical multi-dimensional channel geometries. The void fraction, the liquid velocity and turbulent intensity along the flow direction below and the above the sudden expansion point were measured for various turbulent flow conditions by using a point-electrode resistivity probe and a hot-film anemometry probe. They showed quite complicated behaviors depending upon flow rates of gas and liquid phases and bubble size. In particular, the geometry of sudden expansion affected on the bubble behaviors in multi-dimensional two-phase flow, such as the bubble-stagnation, the bubble-deformation, the enhancement and suppression effects due to the two-phase turbulence etc. Through the measurements, fundamental parameters of the two-phase flow were clarified for the sudden expansion channel. Moreover, a three-dimensional one-way bubble tracking simulation of a single bubble behavior in turbulent flow field along the downstream of the sudden expansion was also demonstrated where equation of motion of bubble was solved by assuming appropriate constitutive models and turbulence model. Based on the trajectories of large number of bubbles, the void fraction distribution was predicted in this calculation. It concretely revealed that the lift force and the two-phase turbulence model were the most important parameters in determining the multi-dimensional void fraction distribution and the calculation should be considered by using the measured experimental data. (author)
International Nuclear Information System (INIS)
Frericks, Bernd B.; Caldarone, Franco C.; Savellano, Dagmar Hoegemann; Stamm, Georg; Kirchhoff, Timm D.; Shin, Hoen-Oh; Galanski, Michael; Nashan, Bjoern; Klempnauer, Juergen; Schenk, Andrea; Selle, Dirk; Spindler, Wolf; Peitgen, Heinz-Otto
2004-01-01
The aim of this study was to evaluate a software tool for non-invasive preoperative volumetric assessment of potential donors in living donated liver transplantation (LDLT). Biphasic helical CT was performed in 56 potential donors. Data sets were post-processed using a non-commercial software tool for segmentation, volumetric analysis and visualisation of liver segments. Semi-automatic definition of liver margins allowed the segmentation of parenchyma. Hepatic vessels were delineated using a region-growing algorithm with automatically determined thresholds. Volumes and shapes of liver segments were calculated automatically based on individual portal-venous branches. Results were visualised three-dimensionally and statistically compared with conventional volumetry and the intraoperative findings in 27 transplanted cases. Image processing was easy to perform within 23 min. Of the 56 potential donors, 27 were excluded from LDLT because of inappropriate liver parenchyma or vascular architecture. Two recipients were not transplanted due to poor clinical conditions. In the 27 transplanted cases, preoperatively visualised vessels were confirmed, and only one undetected accessory hepatic vein was revealed. Calculated graft volumes were 1110±180 ml for right lobes, 820 ml for the left lobe and 270±30 ml for segments II+III. The calculated volumes and intraoperatively measured graft volumes correlated significantly. No significant differences between the presented automatic volumetry and the conventional volumetry were observed. A novel image processing technique was evaluated which allows a semi-automatic volume calculation and 3D visualisation of the different liver segments. (orig.)
Operation control of fluids pumping in curved pipes during annular flow: a numerical evaluation
Directory of Open Access Journals (Sweden)
T Andrade
2016-10-01
Full Text Available To generate projects which provide significant volume recovery from heavy oils reservoirs and improve existing projects, is important to develop new production and transport technologies, especially in the scenario of offshore fields. The core-flow technique is one of new technologies used in heavy oil transportation. This core-flow pattern is characterized by a water pellicle that is formed close or adjacent to the inner wall of the pipe, functioning as a lubricant. The oil flows in the center of the pipe causing a reduction in longitudinal pressure drop. In this sense, this work presents a numerical study of heavy oil annular flow (core-flow assisted by computational tool ANSYS CFX® Release 12.0. It was used a three-dimensional, transient and isothermal mathematical model considered by the mixture and turbulence - models to address the water-heavy oil two-phase flow, assuming laminar flow for oil phase and turbulent flow for water phase. Results of the pressure, velocity and volume fraction distributions of the phases and the pressure drop for different operation conditions are presented and evaluated. It was observed that the oil core flowing eccentrically in the pipe and stops of the water flux considerably increases the pressure drop in the pipe after the restart of the pump.
The calculation of coolant leak rate through the cracks using RELAP5 code
International Nuclear Information System (INIS)
Krungeleviciute, V.; Kaliatka, A.
2001-01-01
For reason to choose method of leak detection first of all it is necessary to perform evaluating thermal-hydraulic calculations. These calculations allow to determine flow rate of discharged coolant. For coolant leak rate calculations through possible cracks in Ignalina NPP pipes SQUIRT and RELAP5 thermal-hydraulic codes were used. SQUIRT is well known as computer program that predicts the leakage for cracked pipes in NPP. As this code calculates only water (at subcooled or saturated conditions) leak rate, RELAP5 code model, that calculates water and steam leak rate, was created. For model validation comparison of SQUIRT, RELAP5 and experimental results was performed. Analysis shows RELAP5 code model suitability for calculations of leak through through-wall cracks in pipes. (author)
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
Smith, Peter
2013-01-01
Written for the piping engineer and designer in the field, this two-part series helps to fill a void in piping literature,since the Rip Weaver books of the '90s were taken out of print at the advent of the Computer Aid Design(CAD) era. Technology may have changed, however the fundamentals of piping rules still apply in the digitalrepresentation of process piping systems. The Fundamentals of Piping Design is an introduction to the designof piping systems, various processes and the layout of pipe work connecting the major items of equipment forthe new hire, the engineering student and the vetera
International Nuclear Information System (INIS)
Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.
1986-01-01
A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)
Automatic seismic support design of piping system by an object oriented expert system
International Nuclear Information System (INIS)
Nakatogawa, T.; Takayama, Y.; Hayashi, Y.; Fukuda, T.; Yamamoto, Y.; Haruna, T.
1990-01-01
The seismic support design of piping systems of nuclear power plants requires many experienced engineers and plenty of man-hours, because the seismic design conditions are very severe, the bulk volume of the piping systems is hyge and the design procedures are very complicated. Therefore we have developed a piping seismic design expert system, which utilizes the piping design data base of a 3 dimensional CAD system and automatically determines the piping support locations and support styles. The data base of this system contains the maximum allowable seismic support span lengths for straight piping and the span length reduction factors for bends, branches, concentrated masses in the piping, and so forth. The system automatically produces the support design according to the design knowledge extracted and collected from expert design engineers, and using design information such as piping specifications which give diameters and thickness and piping geometric configurations. The automatic seismic support design provided by this expert system achieves in the reduction of design man-hours, improvement of design quality, verification of design result, optimization of support locations and prevention of input duplication. In the development of this system, we had to derive the design logic from expert design engineers and this could not be simply expressed descriptively. Also we had to make programs for different kinds of design knowledge. For these reasons we adopted the object oriented programming paradigm (Smalltalk-80) which is suitable for combining programs and carrying out the design work
Investigation of the specific mass flow rate distribution in pipes supplied with a pulsating flow
Energy Technology Data Exchange (ETDEWEB)
Olczyk, Aleksander [Institute of Turbomachinery, Technical University of Lodz, Wolczanska 219/223, 90-924 Lodz (Poland)], E-mail: aolczyk@p.lodz.pl
2009-08-15
A pulsating flow is typical of inlet and exhaust pipes of internal combustion engines and piston compressors. Unsteady flow phenomena are especially important in the case of turbocharged engines, because dynamic effects occurring in the exhaust pipe can affect turbine operation conditions and performance. One of the basic parameters describing the unsteady flow is a transient mass flow rate related to the instantaneous flow velocity, which is usually measured by means of hot-wire anemometers. For the flowing gas, it is more appropriate to analyze the specific mass flow rate {phi}{sub m} = {rho}v, which takes into account also variations in the gas density. In order to minimize the volume occupied by measuring devices in the control section, special double-wire sensors for the specific mass flow rate (CTA) and temperature (CCT) measurement were applied. The article describes procedures of their calibration and measurement. Different forms of calibration curves are analyzed as well in order to match the approximation function to calibration points. Special attention is paid to dynamic phenomena related to the resonance occurring in a pipe for characteristic frequencies depending on the pipe length. One of these phenomena is a reverse flow, which makes it difficult to interpret properly the recorded CTA signal. Procedures of signal correction are described in detail. To verify the measurements, a flow field investigation was carried out by displacing probes radially and determining the profiles of the specific mass flow rate under the conditions of a steady and pulsating flow. The presence and general features of a reverse flow, which was identified experimentally, were confirmed by 1-D unsteady flow calculations.
Investigation of the specific mass flow rate distribution in pipes supplied with a pulsating flow
International Nuclear Information System (INIS)
Olczyk, Aleksander
2009-01-01
A pulsating flow is typical of inlet and exhaust pipes of internal combustion engines and piston compressors. Unsteady flow phenomena are especially important in the case of turbocharged engines, because dynamic effects occurring in the exhaust pipe can affect turbine operation conditions and performance. One of the basic parameters describing the unsteady flow is a transient mass flow rate related to the instantaneous flow velocity, which is usually measured by means of hot-wire anemometers. For the flowing gas, it is more appropriate to analyze the specific mass flow rate φ m = ρv, which takes into account also variations in the gas density. In order to minimize the volume occupied by measuring devices in the control section, special double-wire sensors for the specific mass flow rate (CTA) and temperature (CCT) measurement were applied. The article describes procedures of their calibration and measurement. Different forms of calibration curves are analyzed as well in order to match the approximation function to calibration points. Special attention is paid to dynamic phenomena related to the resonance occurring in a pipe for characteristic frequencies depending on the pipe length. One of these phenomena is a reverse flow, which makes it difficult to interpret properly the recorded CTA signal. Procedures of signal correction are described in detail. To verify the measurements, a flow field investigation was carried out by displacing probes radially and determining the profiles of the specific mass flow rate under the conditions of a steady and pulsating flow. The presence and general features of a reverse flow, which was identified experimentally, were confirmed by 1-D unsteady flow calculations.
40 CFR 80.1407 - How are the Renewable Volume Obligations calculated?
2010-07-01
... is calculated as follows: ER26MR10.430 Where: x = Individual batch of gasoline produced or imported in calendar year i. n = Total number of batches of gasoline produced or imported in calendar year i. GX = Volume of batch x of gasoline produced or imported, as defined in paragraph (c) of this section...
40 CFR 80.1107 - How is the Renewable Volume Obligation calculated?
2010-07-01
... this section is calculated as follows: ER01MY07.061 Where: x = Individual batch of gasoline produced or imported in calendar year i. n = Total number of batches of gasoline produced or imported in calendar year i. GX = Volume of batch x of gasoline produced or imported, in gallons. y = Individual batch of...
Energy Technology Data Exchange (ETDEWEB)
Ishida, E.; Isoyama, R. [Japan Engineering Consultants Co., Ltd., Tokyo (Japan). Public Management Research Center; Koganemaru, K.; Shimuzu, Y. [Tokyo Gas Co. Ltd., Tokyo (Japan). Center for Disaster Management and Supply Control; Morimoto, I. [Kiso-Jiban Consultants Co. Ltd., Tokyo (Japan); Yasuda, S. [Tokyo Denki Univ., Tokyo (Japan). Dept. of Civil and Environmental Engineering
2004-07-01
Estimating the degree of damage to city gas pipe networks is difficult because of the lack of damage case data. This paper proposes a method for calculating the amount of earthquake-induced ground displacement at pipe node locations by constructing ground models. Data for the models was obtained from boreholes and by using a simple ground flow formula. The analysis method will make it possible to calculate the allowable limits of damage-causing factors such as ground motion and flow for different pipe network elements. The analysis procedure was conducted using a 2-dimensional liquefaction-induced flow analysis program finite element method. A real time damage estimation system for low pressure gas pipes uses ground motions having a design seismic coefficient of 0.4 in preparing strong earthquake liquefied layer thickness distribution data. Flow calculations were presented as well as a ground revetment database to replace node location data. It was concluded that achieving consistency was desirable. 7 refs., 2 tabs., 5 figs.
Calculation of partial molar volume of components in supercritical ammonia synthesis system
Institute of Scientific and Technical Information of China (English)
Cunwen WANG; Chuanbo YU; Wen CHEN; Weiguo WANG; Yuanxin WU; Junfeng ZHANG
2008-01-01
The partial molar volumes of components in supercritical ammonia synthesis system are calculated in detail by the calculation formula of partial molar volume derived from the R-K equation of state under different conditions. The objectives are to comprehend phase beha-vior of components and to provide the theoretic explana-tion and guidance for probing novel processes of ammonia synthesis under supercritical conditions. The conditions of calculation are H2/N2= 3, at a concentra-tion of NH3 in synthesis gas ranging from 2% to 15%, Concentration of medium in supercritical ammonia syn-thesis system ranging from 20% to 50%, temperature ran-ging from 243 K to 699 K and pressure ranging from 0.1 MPa to 187 MPa. The results show that the ammonia synthesis system can reach supercritical state by adding a suitable supercritical medium and then controlling the reaction conditions. It is helpful for the supercritical ammonia synthesis that medium reaches supercritical state under the conditions of the corresponding total pres-sure and components near the normal temperature or near the critical temperature of medium or in the range of tem-perature of industrialized ammonia synthesis.
Application of new developments in coupled seismic analysis of piping systems
International Nuclear Information System (INIS)
Gupta, A.; Gupta, A.K.
1995-01-01
The current practice of calculating the seismic response is to perform the analysis of the primary structure (buildings) and the secondary systems (piping) separately. Earthquake input to the primary system in terms of a design response spectrum. An acceleration time history compatible with the design response spectrum is developed (a non-unique process) and primary system is analyzed to obtain the acceleration histories at the desired floors. Floor time histories are used for generating the corresponding instructure response spectrum (IRIS). The instructure response spectra are used as input at the supports of secondary systems. Further, in case of multiple supports, an envelope spectrum (introducing conservatism) is obtained from the individual support IRS. The effect of relative support motion is incorporated by a worst-case separate static analysis (adding to the conservatism). In the above method, mass interaction between the secondary and primary system is ignored, which may have significant effect at resonant frequencies (further adding to the conservatism). The calculated response may be an order of magnitude higher than they should be. Two computer programs, CREST and CREST-IRIS, were developed at Center for NUclear Power Plant Structures, Equipment and Piping. Any one of the two computer programs together with a piping analysis program can be used to perform an accurate coupled seismic analysis of piping systems. The two computer programs have been validated against the time history analysis for simple problems. In the present study, we have applied CREST to analyze two real-life piping systems. The piping analysis program used in this research is the commercial software PIPESTRESS, developed by DST Computer Services of Geneva, Switzerland. (author). 4 refs., 3 figs., 2 tabs
Heat pipe and method of production of a heat pipe
International Nuclear Information System (INIS)
Kemp, R.S.
1975-01-01
The heat pipe consists of a copper pipe in which a capillary network or wick of heat-conducting material is arranged in direct contact with the pipe along its whole length. Furthermore, the interior space of the tube contains an evaporable liquid for pipe transfer. If water is used, the capillary network consists of, e.g., a phosphorus band network. To avoid contamination of the interior of the heat pipe during sealing, its ends are closed by mechanical deformation so that an arched or plane surface is obtained which is in direct contact with the network. After evacuation of the interior space, the remaining opening is closed with a tapered pin. The ratio wall thickness/tube diameter is between 0.01 and 0.6. (TK/AK) [de
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.)
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.)
International Nuclear Information System (INIS)
Grigorov, Grigor N.; Chow, James C.L.; Grigorov, Lenko; Jiang, Runqing; Barnett, Rob B.
2006-01-01
The normal tissue complication probability (NTCP) is a predictor of radiobiological effect for organs at risk (OAR). The calculation of the NTCP is based on the dose-volume-histogram (DVH) which is generated by the treatment planning system after calculation of the 3D dose distribution. Including the NTCP in the objective function for intensity modulated radiation therapy (IMRT) plan optimization would make the planning more effective in reducing the postradiation effects. However, doing so would lengthen the total planning time. The purpose of this work is to establish a method for NTCP determination, independent of a DVH calculation, as a quality assurance check and also as a mean of improving the treatment planning efficiency. In the study, the CTs of ten randomly selected prostate patients were used. IMRT optimization was performed with a PINNACLE3 V 6.2b planning system, using planning target volume (PTV) with margins in the range of 2 to 10 mm. The DVH control points of the PTV and OAR were adapted from the prescriptions of Radiation Therapy Oncology Group protocol P-0126 for an escalated prescribed dose of 82 Gy. This paper presents a new model for the determination of the rectal NTCP ( R NTCP). The method uses a special function, named GVN (from Gy, Volume, NTCP), which describes the R NTCP if 1 cm 3 of the volume of intersection of the PTV and rectum (R int ) is irradiated uniformly by a dose of 1 Gy. The function was 'geometrically' normalized using a prostate-prostate ratio (PPR) of the patients' prostates. A correction of the R NTCP for different prescribed doses, ranging from 70 to 82 Gy, was employed in our model. The argument of the normalized function is the R int , and parameters are the prescribed dose, prostate volume, PTV margin, and PPR. The R NTCPs of another group of patients were calculated by the new method and the resulting difference was <±5% in comparison to the NTCP calculated by the PINNACLE3 software where Kutcher's dose
Parisher, Roy A; Parisher
2000-01-01
Pipe designers and drafters provide thousands of piping drawings used in the layout of industrial and other facilities. The layouts must comply with safety codes, government standards, client specifications, budget, and start-up date. Pipe Drafting and Design, Second Edition provides step-by-step instructions to walk pipe designers and drafters and students in Engineering Design Graphics and Engineering Technology through the creation of piping arrangement and isometric drawings using symbols for fittings, flanges, valves, and mechanical equipment. The book is appropriate primarily for pipe
Water Hammer Mitigation on Postulated Pipe Break of Feed Water System
International Nuclear Information System (INIS)
Seong, Ho Je; Woo, Kab Koo; Cho, Keon Taek
2008-01-01
The Feed Water (FW) system supplies feedwater from the deaerator storage tank to the Steam Generators(S/G) at the required pressure, temperature, flow rate, and water chemistry. The part of FW system, from the S/G to Main Steam Valve House just outside the containment building wall, is designed as safety grade because of its safety function. According to design code the safety related system shall be designed to protect against dynamic effects that may results from a pipe break on high energy lines such as FW system. And the FW system should be designed to minimize blowdown volume of S/G secondary side during the postulated pipe break. Also the FW system should be designed to prevent the initiation or to minimize the effects of water hammer transients which may be induced by the pipe break. This paper shows the results of the hydrodynamic loads induced by the pipe break and the optimized design parameters to mitigate water hammer loads of FW system for Shin-Kori Nuclear Power Plant Unit 3 and 4 (SKN 3 and 4)
A generalized relationship for swirl decay in laminar pipe flow
Indian Academy of Sciences (India)
Swirling ﬂow is of great importance in heat and mass transfer enhancements and in ﬂow measurements. In this study, laminar swirling ﬂow in a straight pipe was considered. Steady three-dimensional axisymmetric Navier–Stokes equations were solved numerically using a control volume approach. The swirl number ...
Directory of Open Access Journals (Sweden)
Nawal H. Al – Raheimy
2016-09-01
Full Text Available In this paper the approximate method of Raleigh method can be used to study the effect of additional boundary conditions (clamped – free & clamped – clamped on the free transverse vibrations of uniform pipes which have length, L (1m , inner radius, "Ri" (1cm & thickness, "t" (1mm made from composite materials, where the resin of unsaturated polyester represented the matrix material reinforced by aligned (E-fibers glass in the first case and used aligned fiber (Kevlar-49 in the second case. The length of fibers is in the two types, the first type is long fibers (continuous and the second is short fibers (discontinuous for different length all at volume fraction of fibers, "f" (0.15 & 0.25. At any construction of the pipe in composite material the natural frequency decreased when the velocity of flow increased from zero to critical velocity also can be observed the pipe at clamped – clamped boundary conditions predicts natural frequency & critical velocity greater than that pipe at clamped – free. The natural frequency and critical velocity increase with increasing volume fraction and length of discontinuous fiber. The value of natural frequency for pipes which have continuous fibers is constant at certain velocity of flow while are variable in pipes which have discontinuous fibers according to ratio between length of short fiber to critical length of discontinuous fiber whereas the natural frequency increase with increasing this ratio. Finally the pipes with Kevlar fiber have high critical velocity and natural frequency compare with pipes for fiber glass.
Development of methodologies for coupled water-hammer analysis of piping systems and supports
International Nuclear Information System (INIS)
Kamil, H.; Gantayat, A.; Attia, A.; Goulding, H.
1983-01-01
The paper presents the results of an investigation on the development of methodologies for coupled water-hammer analyses. The study was conducted because the present analytical methods for calculation of loads on piping systems and supports resulting from water-hammer phenomena are overly conservative. This is mainly because the methods do not usually include interaction between the fluid and the piping and thus predict high loads on piping systems and supports. The objective of the investigation presented in this paper was to develop methodologies for coupled water-hammer analyses, including fluid-structure interaction effects, to be able to obtain realistic loads on piping systems and supports, resulting in production of more economical designs. (orig./RW)
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
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
Application of mathematical model for high viscous damper to dynamic analysis of NPP pipings
International Nuclear Information System (INIS)
Kostarev, V.V.; Bercovsky, A.M.; Kireev, O.B.; Vasiliev, P.S.
1993-01-01
The problems of dynamic analysis of Nuclear Power Plants (NPP) piping systems are considered in the paper. The special calculation program for PC has been developed that enables to estimate the seismic margin for any piping system with different antiseismic devices having nonlinear characteristics. The calculated comparison has been done for two antiseismic supports that are widely used now, namely: a High Viscous Damper (HVD) and a Seismic Stop Support (SSS) with the application, as an example, to the well known pipeline BM3 (USNRC). (author)
Application of mathematical model for high viscous damper to dynamic analysis of NPP pipings
Energy Technology Data Exchange (ETDEWEB)
Kostarev, V V; Bercovsky, A M; Kireev, O B; Vasiliev, P S [CKTI VIBROSEISM (CVS), St. Petersburg (Russian Federation)
1993-07-01
The problems of dynamic analysis of Nuclear Power Plants (NPP) piping systems are considered in the paper. The special calculation program for PC has been developed that enables to estimate the seismic margin for any piping system with different antiseismic devices having nonlinear characteristics. The calculated comparison has been done for two antiseismic supports that are widely used now, namely: a High Viscous Damper (HVD) and a Seismic Stop Support (SSS) with the application, as an example, to the well known pipeline BM3 (USNRC). (author)
Directory of Open Access Journals (Sweden)
Jonny Nordström
2017-11-01
Full Text Available Abstract Background Quantitative measurement of myocardial blood flow (MBF is of increasing interest in the clinical assessment of patients with suspected coronary artery disease (CAD. 15O-water positron emission tomography (PET is considered the gold standard for non-invasive MBF measurements. However, calculation of left ventricular (LV volumes and ejection fraction (EF is not possible from standard 15O-water uptake images. The purpose of the present work was to investigate the possibility of calculating LV volumes and LVEF from cardiac-gated parametric blood volume (V B 15O-water images and from first pass (FP images. Sixteen patients with mitral or aortic regurgitation underwent an eight-gate dynamic cardiac-gated 15O-water PET/CT scan and cardiac MRI. V B and FP images were generated for each gate. Calculations of end-systolic volume (ESV, end-diastolic volume (EDV, stroke volume (SV and LVEF were performed with automatic segmentation of V B and FP images, using commercially available software. LV volumes and LVEF were calculated with surface-, count-, and volume-based methods, and the results were compared with gold standard MRI. Results Using V B images, high correlations between PET and MRI ESV (r = 0.89, p 0.86, p < 0.001. Conclusion Calculation of LV volumes and LVEF from dynamic 15O-water PET is feasible and shows good correlation with MRI. However, the analysis method is laborious, and future work is needed for more automation to make the method more easily applicable in a clinical setting.
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)
Heat pipes and use of heat pipes in furnace exhaust
Polcyn, Adam D.
2010-12-28
An array of a plurality of heat pipe are mounted in spaced relationship to one another with the hot end of the heat pipes in a heated environment, e.g. the exhaust flue of a furnace, and the cold end outside the furnace. Heat conversion equipment is connected to the cold end of the heat pipes.
International Nuclear Information System (INIS)
Winslow, D.W.; Brisco, D.P.
1991-01-01
This patent describes a method of stopping flow of fluid up through a pipe bore of a pipe string in a well. It comprises: lowering a bridge plug apparatus on a work string into the pipe string to a position where the pipe bore is to be closed; communicating the pipe bore below a packer of the bridge plug apparatus through the bridge plug apparatus with a low pressure zone above the packer to permit the fluid to flow up through the bridge plug apparatus; engaging the bridge plug apparatus with an internal upset of the pipe string; while the fluid is flowing up through the bridge plug apparatus, pulling upward on the work string and the bridge plug apparatus and thereby sealing the packer against the pipe bore; isolating the pipe bore below the packer from the low pressure zone above the packer and thereby stopping flow of the fluid up through the pipe bore; disconnecting the work string from the bridge plug apparatus; and maintaining the bridge plug apparatus in engagement with the internal upset and sealed against the pipe bore due to an upward pressure differential applied to the bridge plug apparatus by the fluid contained therebelow
The method for measuring residual stress in stainless steel pipes
International Nuclear Information System (INIS)
Shimov, Georgy; Rozenbaum, Mikhail; Serebryakov, Alexandr; Serebryakov, Andrey
2016-01-01
The main reason of appearance and growth of corrosion damages of the nuclear steam generator heat exchanger tubes is the process of stress-corrosion cracking of metal under the influence of residual tensile stress. Methods used in the production for estimating residual stresses (such as a method of ring samples) allow measuring only the average tangential stress of the pipe wall. The method of ring samples does not allow to assess the level of residual stress in the surface layer of the pipe. This paper describes an experimental method for measuring the residual stresses on the pipe surface by etching a thin surface layer of the metal. The construction and working principle of a trial installation are described. The residual stresses in the wall of the tubes 16 × 1.5 mm (steel AISI 321) for nuclear steam generators is calculated. Keywords: heat exchange pipes, stress corrosion cracking, residual stresses, stress distribution, stress measurement.
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.)
Comparison between conventional heat exchanger performance and an heat pipes exchanger
International Nuclear Information System (INIS)
Souza, J.R.G. de; Rocha, N.R.
1989-01-01
The thermal performance of conventional compact heat exchanger and of exchanger with heat pipes are simulated using a digital computer, for equal volumes and the same process conditions. The comparative analysis is depicted in graphs that indicate which of the situations each equipment is more efficient. (author)
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
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
1997-01-01
Small Business Innovation Research contracts from Goddard Space Flight Center to Thermacore Inc. have fostered the company work on devices tagged "heat pipes" for space application. To control the extreme temperature ranges in space, heat pipes are important to spacecraft. The problem was to maintain an 8-watt central processing unit (CPU) at less than 90 C in a notebook computer using no power, with very little space available and without using forced convection. Thermacore's answer was in the design of a powder metal wick that transfers CPU heat from a tightly confined spot to an area near available air flow. The heat pipe technology permits a notebook computer to be operated in any position without loss of performance. Miniature heat pipe technology has successfully been applied, such as in Pentium Processor notebook computers. The company expects its heat pipes to accommodate desktop computers as well. Cellular phones, camcorders, and other hand-held electronics are forsible applications for heat pipes.
Calculated volumes of individual shield volcanoes at the young end of the Hawaiian Ridge
Robinson, Joel E.; Eakins, Barry W.
2006-03-01
High-resolution multibeam bathymetry and a digital elevation model of the Hawaiian Islands are used to calculate the volumes of individual shield volcanoes and island complexes (Niihau, Kauai, Oahu, the Maui Nui complex, and Hawaii), taking into account subsidence of the Pacific plate under the load of the Hawaiian Ridge. Our calculated volume for the Island of Hawaii and its submarine extent (213 × 10 3 km 3) is nearly twice the previous estimate (113 × 10 3 km 3), due primarily to crustal subsidence that had not been accounted for in the earlier work. The volcanoes that make up the Island of Hawaii (Mahukona, Kohala, Mauna Kea, Hualalai, Mauna Loa, Kilauea and Loihi) are generally considered to have been formed within the past million years, and our revised volume for the island indicates that magma supply rates are greater than previously estimated, 0.21 km 3/yr as opposed to ˜ 0.1 km 3/yr. This result also shows that compared with rates calculated for the Hawaiian Islands (0-6 Ma, 0.095 km 3/yr), the Hawaiian Ridge (0-45 Ma, 0.017 km 3/yr), and the Emperor Seamounts (45-80 Ma, 0.010 km 3/yr), magma supply rates have increased dramatically to build the Island of Hawaii.
Examination of the X-ray piping diagnostic system using EGS4 (examination of the film and iron rust)
International Nuclear Information System (INIS)
Kajiwara, G.
2000-01-01
In the X-ray piping diagnosis system, X-ray photograph is taken of the used pipes, and from the density of the image of the pipe on the film, the thickness of the pipe wall is measured using the relationship between the density and the thickness. First, as for the relationship between the absorbed energy and the density on the film, though good agreement was obtained last year, it is improved further by making energy bin smaller in the calculation of EGS4. The reason of the agreement was researched and understood. Next, using EGS4, the calculation of the thickness of the steel was carried out which was covered with the rust, using the element analysis result of the rust sample that was collected in the old pipe. When the thickness changes, the rate of the energy absorption of the steel and the rust layer changes. This relationship between the energy absorption and the thickness of the layers is expressed approximately in a formula. It will be reflected on the diagnosis of the pipes. (author)
Energy Technology Data Exchange (ETDEWEB)
Bennett, W.; Jimenez, A.F.
1987-09-08
This patent describes a method for storing and retrieving a riser pipe, comprising the steps of: providing an upright annular magazine comprised of an inside annular wall and an outside annular wall, the magazine having an open top; storing the riser pipe in a substantially vertically oriented position within the annular magazine; and moving the riser pipe upwardly through the open top of the annular magazine at an angle to the vertical along at least a portion of the length of the riser pipe.
Development of two-phase flow along a large vertical pipe
International Nuclear Information System (INIS)
Dirk Lucas; Prasser, H.M.
2005-01-01
Full text of publication follows: To qualify CFD codes for two-phase flow simulations, closure laws describing the interaction between the phases are needed. Vertical pipe flow is a suitable object for studying the corresponding phenomena in case of dispersed bubbly flow. Here, the bubbles move under clear boundary conditions, resulting in a shear field of nearly constant structure where the bubbles rise for a comparatively long time. This allows to study the lateral motion of the bubbles in a shear flow as well as bubble coalescence and break-up by comparing gas volume fraction distributions and bubble size distributions at different heights. Very detailed data were obtained at the TOPFLOW facility of the Forschungszentrum Rossendorf using an advanced wire-mesh sensor. This sensor measures the instantaneous conductivity distribution over the pipe cross section. The high frequency of the measurement (2500 frames/s) allows the detection of single bubbles by a special evaluation procedure. Bubble size distributions, gas volume fraction distributions and also gas fraction distributions decomposed according to the bubble size are delivered as result of the evaluation procedure. The use of two sensors allows to measure the profile of the gas velocity. In previous works similar data for pipe of 51.2 mm inner diameter were used for the validation of non-drag bubble forces [1] and the evaluation of the influence of radial profiles on the development of the flow pattern [2]. First investigations on scaling effects were done using data obtained at a pipe with an inner diameter of 194 mm [3]. A constant distance between gas injection and measuring plane of L/D ∼ 40 was used. From a new test series now measurements are available for varying distances between the injection device and the wire-mesh sensor. This allows the evaluation of the development of the flow along the pipe. The data are used for the development and validation of mesoscale models for the forces acting on
A study on the temperature distribution in the hot leg pipe
International Nuclear Information System (INIS)
Choe, Yoon-Jae; Baik, Se-Jin; Jang, Ho-Cheol; Lee, Byung-Jin; Im, In-Young; Ro, Tae-Sun
2003-01-01
In the hot leg pipes of reactor coolant system of the Korean Standard Nuclear Power Plant (KSNP), a non-uniform distribution in temperature has been observed across the cross-section, which is attributed to the non-uniformity of power distribution in the reactor core usually having a peak in the center region, and to the colder coolant bypass flow through the reactor vessel outlet nozzle clearances. As a result, the arithmetic mean temperature of four Resistance Temperature Detectors (RTDs) installed in each hot leg - two in the upper region and two in the lower region around the pipe wall may not correctly represent the actual coolant bulk temperature. It is also believed that there is a skewness in the velocity profile in the hot leg pipe due to the sudden changes in the flow direction and area from the core to the hot leg pipe, through the reactor vessel outlet plenum. These temperature non-uniformity and velocity skewness affect the measurement of the plant parameter such as the reactor coolant flow rate which is calculated by using the bulk temperature of hot leg pipes. A computational analysis has been performed to simulate the temperature and velocity distributions and to evaluate the uncertainty of temperature correction offset in the hot leg pipe. A commercial CFD code, FLUENT, is used for this analysis. The analysis results are compared with the operational data of KSNP and the scaled-down model test data for System 80. From the comparisons, an uncertainty of correction offset is obtained to measure the bulk temperature of hot leg more accurately, which can be also applied to the operating plants, leading to the reduction of temperature measurement uncertainty. Since the uncertainty of temperature in the hot leg pipe is one of major parameters to calculate the uncertainty of the reactor coolant flow rate, the analysis results can contribute to the improvement of the plant performance and safety by reducing the uncertainty of temperature measurement
Developments in Pulsed Neutron Activation for Determination of Water Flow in Pipes
Mattsson, H
2003-01-01
In PNA (pulsed neutron activation) it is important that the measured data can be related to the total mass flow. In this thesis two fundamental problems of the measurement technique and data treatment have been investigated: transport/mixing and background radiation. The principle of PNA is to introduce a radioactive substance into a pipe by bombarding fluid in the pipe with neutron pulses. The fluid in the pipe is activated and subsequently transported and mixed with the flow. Gamma radiation emitted from the activity is measured with one or two detectors downstream from the activation point. The time-resolved signal from the detectors is used to calculate the average velocity of the water flow. Due to the short distance between the neutron generator and the pipe the activity in the pipe becomes highly inhomogeneous. The transport and mixing of the activity were simulated using colour which was injected into the flow. It was found that the inhomogeneous activity distribution must be taken into account if the...
SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK
Directory of Open Access Journals (Sweden)
Tang Van Lam
2017-11-01
Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and
Development of a new concrete pipe molding machine using topology optimization
International Nuclear Information System (INIS)
Park, Hong Seok; Dahal, Prakash; Nguyen, Trung Thanh
2016-01-01
Sulfur polymer concrete (SPC) is a relatively new material used to replace Portland cement for manufacturing sewer pipes. The objective of this work is to develop an efficient molding machine with an inner rotating die to mix, compress and shape the SPC pipe. First, the alternative concepts were generated based on the TRIZ principles to overcome the drawbacks of existing machines. Then, the concept scoring technique was used to identify the best design in terms of machine structure and product quality. Finally, topology optimization was applied with the support of the density method to reduce mass and to displace the inner die. Results showed that the die volume can be reduced by approximately 9% and the displacement can be decreased by approximately 3% when compared with the initial design. This work is expected to improve the manufacturing efficiency of the concrete pipe molding machine
Development of a new concrete pipe molding machine using topology optimization
Energy Technology Data Exchange (ETDEWEB)
Park, Hong Seok; Dahal, Prakash [School of Mechanical and Automotive Engineering, University of Ulsan, Ulsan (Korea, Republic of); Nguyen, Trung Thanh [Faculty of Mechanical Engineering, Le Quy Don Technical University, Hanoi (Viet Nam)
2016-08-15
Sulfur polymer concrete (SPC) is a relatively new material used to replace Portland cement for manufacturing sewer pipes. The objective of this work is to develop an efficient molding machine with an inner rotating die to mix, compress and shape the SPC pipe. First, the alternative concepts were generated based on the TRIZ principles to overcome the drawbacks of existing machines. Then, the concept scoring technique was used to identify the best design in terms of machine structure and product quality. Finally, topology optimization was applied with the support of the density method to reduce mass and to displace the inner die. Results showed that the die volume can be reduced by approximately 9% and the displacement can be decreased by approximately 3% when compared with the initial design. This work is expected to improve the manufacturing efficiency of the concrete pipe molding machine.
Wits, W.W.; Kok, J.B.W.; van Steenhoven, A.A.; van der Meer, T.H.; Stoffels, G.G.M.
2008-01-01
The heat pipe is a two-phase cooling solution, offering very high thermal coefficients, for heat transport. Therefore, it is increasingly used in the design of electronic products. Flat miniature heat pipes are able to effectively remove heat from several hot spots on a Printed Circuit Board (PCB).
Calculation code for erosion-corrosion induced wall thinning in piping systems
International Nuclear Information System (INIS)
Henzel, N.; Kastner, W.; Stellwag, B.; Erve, M.
1988-01-01
There was great material erosion mainly in consequence of an extremely unfavourable geometry at the damaged place in Surry-2. The pipeline sections affected in Trojan were in the area of action of great sources of turbulence, i.e.: less than 10 pipe diameters from junctions, elbows etc. Because of the many parameters which determine the amount of material removal by erosion-corrosion, the analysis of such damage is only possible using a computer program. The main purpose of such a PC code called WATHEC developed by Siemens/KWU is not the subsequent confirmation of damage which has occurred, but its application for preventive diagnosis in pipeline systems. (orig./DG) [de
Development of LBB Piping Evaluation Diagram for APR 1000 Main Steam Line Piping
International Nuclear Information System (INIS)
Yang, J. S.; Jeong, I. L.; Park, C. Y.; Bai, S. Y.
2010-01-01
This paper presents the piping evaluation diagram (PED) to assess the applicability of Leak-Before- Break(LBB) for APR 1000 main steam line piping. LBB-PED of APR 1000 main steam line piping is independent of its piping geometry and has a function of the loads applied in piping system. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the LBB-PED. The LBB-PED, therefore, can be used for quick LBB evaluation of APR 1000 main steam line piping of both design and construction
International Nuclear Information System (INIS)
1988-09-01
This document presents the piping research program plan for the Structural and Seismic Engineering Branch and the Materials Engineering Branch of the Division of Engineering, Office of Nuclear Regulatory Research. The plan describes the research to be performed in the areas of piping design criteria, environmentally assisted cracking, pipe fracture, and leak detection and leak rate estimation. The piping research program addresses the regulatory issues regarding piping design and piping integrity facing the NRC today and in the foreseeable future. The plan discusses the regulatory issues and needs for the research, the objectives, key aspects, and schedule for each research project, or group of projects focussing of a specific topic, and, finally, the integration of the research areas into the regulatory process is described. The plan presents a snap-shot of the piping research program as it exists today. However, the program plan will change as the regulatory issues and needs change. Consequently, this document will be revised on a bi-annual basis to reflect the changes in the piping research program. (author)
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)
Directory of Open Access Journals (Sweden)
Melnyk R. S.
2017-04-01
Full Text Available Aluminium and copper heat pipes with grooved and metal fibrous capillary structure are high effective heat transfer devices. They are used in different cooling systems of electronic equipment like a LED modules, microprocessors, receive-transmit modules and so on. However thus heat pipes have heat transfer limitations. There are few types of this limitations: hydraulic limitation, boiling limitation, liquid entrainment by vapor flow and sonic limitation. There is necessity to know which one of these limitations is determinant for heat pipe due to design process. At a present article calculations of maximum heat transfer ability represented. All these calculations were made for LED cooling by using heat pipes with grooved and metal fibrous capillary structures. Pentane, acetone, isobutane and water were used as a coolants. It was shown that the main operation limit for axial grooved heat pipe, which determinate maximum heat transfer ability due to inclination angle for location of cooling zone higher than evaporation zone case, is entrainment limit for pentane and acetone coolants. Nevertheless, for isobutane coolant the main limitation is a boiling limit. However, for heat pipes with metal fibrous capillary structure the main limitation is a capillary limit. This limitation was a determinant for all calculated coolants: water, pentane and acetone. For high porosity range of capillary structure, capillary limit transfer to sonic limit for heat pipes with water, that means that the vapor velocity increases to sonic velocity and can't grow any more. Due to this, coolant cant in a needed quantity infill condensation zone and the last one drained. For heat pipes with acetone and pentane, capillary limit transfer to boiling limit. All calculations were made for vapor temperature equal to 50°C, and for porosity range from 30% to 90%.
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.)
Optimum Design of FGX-CNT-Reinforced Reddy Pipes Conveying Fluid Subjected to Moving Load
Directory of Open Access Journals (Sweden)
Farid Vakili Tahami
2016-12-01
Full Text Available The harmony search algorithm is applied to the optimum designs of functionally graded (FG-carbon nanotubes (CNTs-reinforced pipes conveying fluid which are subjected to a moving load. The structure is modeled by the Reddy cylindrical shell theory, and the motion equations are derived by Hamilton's principle. The dynamic displacement of the system is derived based on the differential quadrature method (DQM. Moreover, the length, thickness, diameter, velocity, and acceleration of the load, the temperature and velocity of the fluid, and the volume fraction of CNT are considered for the design variables. The results illustrate that the optimum diameter of the pipe is decreased by increasing the volume percentage of CNTs. In addition, by increasing the moving load velocity and acceleration, the FS is decreased.
Characterization of radioactive contamination inside pipes with the Pipe Explorer trademark system
International Nuclear Information System (INIS)
Kendrick, D.T.; Cremer, C.D.; Lowry, W.; Cramer, E.
1995-01-01
The U.S. Department of Energy's nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer trademark system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane. Advantages of this approach include the capability of deploying through constrictions in the pipe, around 90 degrees bends, vertically up and down, and in slippery conditions. Because the detector is transported inside the membrane (which is inexpensive and disposable), it is protected from contamination, which eliminates cross-contamination. Characterization sensors that have been demonstrated with the system thus far include: gamma detectors, beta detectors, video cameras, and pipe locators. Alpha measurement capability is currently under development. A remotely operable Pipe Explorer trademark system has been developed and demonstrated for use in DOE facilities in the decommissioning stage. The system is capable of deployment in pipes as small as 2-inch-diameter and up to 250 feet long. This paper describes the technology and presents measurement results of a field demonstration conducted with the Pipe Explorer trademark system at a DOE site. These measurements identify surface activity levels of U-238 contamination as a function of location in drain lines. Cost savings to the DOE of approximately $1.5 million dollars were realized from this one demonstration
ATHENA calculation model for the ITER-FEAT divertor cooling system. Final report with updates
International Nuclear Information System (INIS)
Eriksson, John; Sjoeberg, A.; Sponton, L.L.
2001-05-01
An ATHENA model of the ITER-FEAT divertor cooling system has been developed for the purpose of calculating and evaluating consequences of different thermal-hydraulic accidents as specified in the Accident Analysis Specifications for the ITER-FEAT Generic Site Safety Report. The model is able to assess situations for a variety of conceivable operational transients from small flow disturbances to more critical conditions such as total blackout caused by a loss of offsite and emergency power. The main objective for analyzing this type of scenarios is to determine margins against jeopardizing the integrity of the divertor cooling system components and pipings. The model of the divertor primary heat transport system encompasses the divertor cassettes, the port limiter systems, the pressurizer, the heat exchanger and all feed and return pipes of these components. The development was pursued according to practices and procedures outlined in the ATHENA code manuals using available modelling components such as volumes, junctions, heat structures and process controls
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
International Nuclear Information System (INIS)
Hara, F.; Seto, K.
1987-01-01
The design value of damping for nuclear piping systems is a vital parameter in ensuring safety in nuclear plants during large earthquakes. Many experiments and on-site tests have been undertaken in nuclear-industry developed countries to determine rational design values. However damping value in nuclear piping systems is so strongly influenced by many piping parameters that it shows a tremendous dispersion in its experimental values. A new trend has recently appeared in designing nuclear pipings, where they attempt to use a device to absorb vibration energy induced by seismic excitation. A typical device is an energy absorbing device, made of a special material having a high capacity of plasticity, which is installed between the piping and the support. This paper deals with the basic study of application of dual vibration absorbers to nuclear piping systems to accomplish high damping value and reduce consequently seismic response at resonance frequencies of a piping system, showing their effectiveness from not only numerical calculation but also experimental evaluation of the vibration responses in a 3D model piping system equipped with dual two vibration absorbers
Evaluation of residual stresses for the multipass welds of 316L stainless steel pipe
International Nuclear Information System (INIS)
Kim, S. H.; Joo, Y. S.; Lee, J. H.
2003-01-01
It is necessary to evaluate the influence of the residual stress and distortion in the design and fabrication of welded structure and the sound welded structure can be maintained by this consideration. Multipass welds of the 316L stainless steel have been widely employed in the pipes of Liquid Metal Reactor. In this study, the residual stresses in the 316L stainless steel pipe welds were calculated by the finite element method using ANSYS code. Also, the residual stresses both on the surface and in the interior of the thickness were measured by HRPD(High Resolution Powder Diffractometer) instrumented in HANARO Reactor. The residual stresses were measured for each 18 points in small(t/d=0.075) and large pipe specimens (t/d=0.034). The experimental and calculated results were compared and the characteristics of the distribution of the residual stress discussed
International Nuclear Information System (INIS)
Koponen, B.L.; Hampel, V.E.
1982-01-01
This compilation contains 688 complete summaries of papers on nuclear criticality safety as presented at meetings of the American Nuclear Society (ANS). The selected papers contain criticality parameters for fissile materials derived from experiments and calculations, as well as criticality safety analyses for fissile material processing, transport, and storage. The compilation was developed as a component of the Nuclear Criticality Information System (NCIS) now under development at the Lawrence Livermore National Laboratory. The compilation is presented in two volumes: Volume 1 contains a directory to the ANS Transaction volume and page number where each summary was originally published, the author concordance, and the subject concordance derived from the keyphrases in titles. Volume 2 contains - in chronological order - the full-text summaries, reproduced here by permission of the American Nuclear Society from their Transactions, volumes 1-41
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.
Piping reliability model development, validation and its applications to light water reactor piping
International Nuclear Information System (INIS)
Woo, H.H.
1983-01-01
A brief description is provided of a three-year effort undertaken by the Lawrence Livermore National Laboratory for the piping reliability project. The ultimate goal of this project is to provide guidance for nuclear piping design so that high-reliability piping systems can be built. Based on the results studied so far, it is concluded that the reliability approach can undoubtedly help in understanding not only how to assess and improve the safety of the piping systems but also how to design more reliable piping systems
Unified pipe network method for simulation of water flow in fractured porous rock
Ren, Feng; Ma, Guowei; Wang, Yang; Li, Tuo; Zhu, Hehua
2017-04-01
Rock masses are often conceptualized as dual-permeability media containing fractures or fracture networks with high permeability and porous matrix that is less permeable. In order to overcome the difficulties in simulating fluid flow in a highly discontinuous dual-permeability medium, an effective unified pipe network method is developed, which discretizes the dual-permeability rock mass into a virtual pipe network system. It includes fracture pipe networks and matrix pipe networks. They are constructed separately based on equivalent flow models in a representative area or volume by taking the advantage of the orthogonality of the mesh partition. Numerical examples of fluid flow in 2-D and 3-D domain including porous media and fractured porous media are presented to demonstrate the accuracy, robustness, and effectiveness of the proposed unified pipe network method. Results show that the developed method has good performance even with highly distorted mesh. Water recharge into the fractured rock mass with complex fracture network is studied. It has been found in this case that the effect of aperture change on the water recharge rate is more significant in the early stage compared to the fracture density change.
International Nuclear Information System (INIS)
1984-12-01
Six topical areas were covered by the Task Group on Other Dynamic Loads and Load Combinations as described below: Event Combinations - dealing with the potential simultaneous occurrence of earthquakes, pipe ruptures, and water hammer events in the piping design basis; Response Combinations - dealing with multiply supported piping with independent inputs, the sequence of combinations between spacial and modal components of response, and the treatment of high frequency modes in combination with low frequency modal responses; Stress Limits/Dynamic Allowables - dealing with inelastic allowables for piping and strain rate effects; Water Hammer Loadings - dealing with code and design specifications for these loadings and procedures for identifying potential water hammer that could affect safety; Relief Valve Opening and Closing Loads - dealing with the adequacy of analytical tools for predicting the effects of these events and, in addition, with estimating effective cycles for fatigue evaluations; and Piping Vibration Loads - dealing with evaluation procedures for estimating other than seismic vibratory loads, the need to consider reciprocating and rotary equipment vibratory loads, and high frequency vibratory loads. NRC staff recommendations or regulatory changes and additional study appear in this report
Results of a bench mark test on the crack opening and leak rate calculation
International Nuclear Information System (INIS)
Grebner, H.
1995-01-01
Results of a bench mark test on the standard problem calculation of crack opening and leak rate in piping components are presented. The bench mark test is based on two experiments performed in phase III of the German HDR safety program. The pipe sections considered in these experiments were a straight pipe with an 80 mm diameter containing a circumferential wall penetrating crack and a pipe branch DN 100/DN 25 with a crack in the weldment between the nozzle and the main pipe. Both test pieces were made of austenitic steel and were loaded by internal pressure and bending moment. For the evaluation of the crack opening either analytical methods or estimation schemes or the finite element method were used, while leak rates were calculated by means of two-phase flow methods. The compilation of the results shows very large scatter bands in general, with deviations between calculated and measured values of up to some one hundred percent. Reasons for this behaviour are uncertainties in the measured data and their evaluation as well as the different methods of calculation and their uncertainties. (author)
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.
Expanding the calculation of activation volumes: Self-diffusion in liquid water
Piskulich, Zeke A.; Mesele, Oluwaseun O.; Thompson, Ward H.
2018-04-01
A general method for calculating the dependence of dynamical time scales on macroscopic thermodynamic variables from a single set of simulations is presented. The approach is applied to the pressure dependence of the self-diffusion coefficient of liquid water as a particularly useful illustration. It is shown how the activation volume associated with diffusion can be obtained directly from simulations at a single pressure, avoiding approximations that are typically invoked.
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)
Calculation of fluid (steam) hammer loading to piping systems by the response spectrum method
International Nuclear Information System (INIS)
Krause, G.; Schrader, W.; Leimbach, K.R.
1983-01-01
Today computations of fluid and steam hammer loading to piping systems are usually performed as a time-history analysis in which the transient pressure forces act as external excitations. For practical purposes it is desirable to be able to treat fluid hammer loading using the response spectrum method similarily as loads from external events. Two advantages arise from the use of spectra in the analysis of piping systems subjected to dynamic force excitations. Firstly, the response spectrum method is much less sensitive to model idealization than the time-history method. Secondly, computational efforts are reduced. In this paper the algorithm for the treatment of force excitations through the modal response spectrum method is briefly presented. The effect of the residuum accounting for higher modes which are not part of the modal decomposition is considered. In particular various methods of superposition of the responses of the dynamic forces and of the modes are investigated. Results and comparisons are presented of several response spectrum analyses and time-history analyses. (orig.)
Evaluation of wall thinning profile by flow accelerated corrosion in separation and union pipe
International Nuclear Information System (INIS)
Watanabe, Shun; Yoneda, Kimitoshi
2013-01-01
Flow Accelerated Corrosion (FAC) is a pipe wall thinning phenomena to be monitored and managed in power plants with high priority. At present, its management has been conducted with conservative evaluation of thinning rate and residual lifetime of the piping based on wall thickness measurements. However, noticeable case of wall thinning was occurred at separation and union pipe. In such pipe system, it is a problem to manage section beneath reinforcing plate of T-tube pipe and 'crotch' of T-joint pipe; the region where wall thickness measurement is difficult to conduct with ordinary ultrasonic testing device. In this study, numerical analysis for separation and union part of T-tube and T-joint pipe was conducted, and wall thinning profile by Flow Accelerated Corrosion was evaluated by calculating mass transfer coefficient and geometry factor. Based on these results, we considered applicable wall thinning management for T-tube and T-joint pipe. In the case of union flow from main and branch pipe, the wall thinning profile of T-tube showed the tendency of increase at main pipe like semielliptical region. On the other hand, noticeable profile appeared at 'crotch' in T-joint. Although it was found that geometry factor of T-joint in this case was half the value of T-tube, an alternative evaluation method to previous one might be needed for the profiles of 'semielliptical region' and 'crotch'. (author)
International Nuclear Information System (INIS)
Kirisits, Christian; Wexberg, Paul; Gottsauner-Wolf, Michael; Pokrajac, Boris; Ortmann, Elisabeth; Aiginger, Hannes; Glogar, Dietmar; Poetter, Richard
2001-01-01
Background and purpose: Radioactive stents are under investigation for reduction of coronary restenosis. However, the actual dose delivered to specific parts of the coronary artery wall based on the individual vessel anatomy has not been determined so far. Dose-volume histograms (DVHs) permit an estimation of the actual dose absorbed by the target volume. We present a method to calculate DVHs based on intravascular ultrasound (IVUS) measurements to determine the dose distribution within the vessel wall. Materials and methods: Ten patients were studied by intravascular ultrasound after radioactive stenting (BX Stent, P-32, 15-mm length) to obtain tomographic cross-sections of the treated segments. We developed a computer algorithm using the actual dose distribution of the stent to calculate differential and cumulative DVHs. The minimal target dose, the mean target dose, the minimal doses delivered to 10 and 90% of the adventitia (DV10, DV90), and the percentage of volume receiving a reference dose at 0.5 mm from the stent surface cumulated over 28 days were derived from the DVH plots. Results were expressed as mean±SD. Results: The mean activity of the stents was 438±140 kBq at implantation. The mean reference dose was 111±35 Gy, whereas the calculated mean target dose within the adventitia along the stent was 68±20 Gy. On average, DV90 and DV10 were 33±9 Gy and 117±41 Gy, respectively. Expanding the target volume to include 2.5-mm-long segments at the proximal and distal ends of the stent, the calculated mean target dose decreased to 55±17 Gy, and DV 90 and DV 10 were 6.4±2.4 Gy and 107±36 Gy, respectively. Conclusions: The assessment of DVHs seems in principle to be a valuable tool for both prospective and retrospective analysis of dose-distribution of radioactive stents. It may provide the basis to adapt treatment planning in coronary brachytherapy to the common standards of radiotherapy
Energy Technology Data Exchange (ETDEWEB)
Chen, Y.L.; Wang, G.Z., E-mail: gzwang@ecust.edu.cn; Xuan, F.Z.; Tu, S.T.
2015-04-15
Highlights: • Solution of constraint parameter τ* for through-wall cracked pipes has been obtained. • Constraint increases with increasing crack length and radius–thickness ratio of pipes. • Constraint-dependent LBB curve for through-wall cracked pipes has been constructed. • For increasing accuracy of LBB assessments, constraint effect should be considered. - Abstract: The leak-before-break (LBB) concept has been widely applied in the structural integrity assessments of pressured pipes in nuclear power plants. However, the crack-tip constraint effects in LBB analyses and designs cannot be incorporated. In this paper, by using three-dimensional finite element calculations, the modified load-independent T-stress constraint parameter τ* for circumferential through-wall cracked pipes with different geometries and crack sizes has been analyzed under different loading conditions, and the solutions of the crack-tip constraint parameter τ* have been obtained. Based on the τ* solutions and constraint-dependent J–R curves of a steel, the constraint-dependent LBB (leak-before-break) curves have been constructed. The results show that the constraint τ* increases with increasing crack length θ, mean radius R{sub m} and radius–thickness ratio R{sub m}/t of the pipes. In LBB analyses, the critical crack length calculated by the J–R curve of the standard high constraint specimen for pipes with shorter cracks is over-conservative, and the degree of conservatism increases with decreasing crack length θ, R{sub m} and R{sub m}/t. Therefore, the constraint-dependent LBB curves should be constructed to modify the over-conservatism and increase accuracy of LBB assessments.
Temperature control with high performance gravity-assist heat pipes
International Nuclear Information System (INIS)
Kemme, J.E.; Deverall, J.E.; Keddy, E.S.; Phillips, J.R.; Ranken, W.A.
1975-01-01
The development of high performance heat pipes for controlling the temperature of irradiation experiments in the Experimental Breeder Reactor (EBR-II) is described. Because this application involves vertical operation in a gravity-assist mode with the evaporator down, several tests were made with sodium and potassium heat pipes in this position to establish their performance limits as a function of operating temperature. Best performance was achieved with a new wick structure consisting of a fine porous liner next to the heat-pipe wall and four helical channels next to the vapor passage. Also, a new modification of heat-pipe theory was discovered for determining performance limits for this type of wick. In its most rudimentary form, this modification says that the dynamic pressure gradient in the vapor stream cannot exceed the gravity gradient causing return of liquid. Once this modification was expressed in the form of a limiting equation, and a term was added to account for the slight capillary force developed in the channels, good agreement was obtained between calculated limits and those measured in several tests with both sodium and potassium. These tests showed rather conclusively that only half of the liquid head in the evaporator section was causing return of condensate, whereas existing theory predicts that the full head of liquid in the heat pipe is available for condensate return. (U.S.)
Flow through a cylindrical pipe with a periodic array of fractal orifices
van Melick, P.A.J.; Geurts, Bernardus J.
2013-01-01
We apply direct numerical simulation (DNS) of the incompressible Navier–Stokes equations to predict flow through a cylindrical pipe in which a periodic array of orifice plates with a fractal perimeter is mounted. The flow is simulated using a volume penalization immersed boundary method with which
Flow through a cylindrical pipe with a periodic array of fractal orifices
van Melick, P.A.J.; Geurts, B.J.
2013-01-01
We apply direct numerical simulation (DNS) of the incompressible Navier-Stokes equations to predict flow through a cylindrical pipe in which a periodic array of orifice plates with a fractal perimeter is mounted. The flow is simulated using a volume penalization immersed boundary method with which
Energy Technology Data Exchange (ETDEWEB)
Gomez-Tenedor Alonso, S.; Rincon Perez, M.; Penedo Cobos, J. M.; Garcia Castejon, M. A.
2011-07-01
The discrepancies in the calculation of the same volume between different planning systems impact on dose-volume histograms and therefore clinical assessment of dosimetry for patients. The transfer, by a local network, tomographic study (CT) and contours of critical organs of patients, between our two planning systems allows us to evaluate the calculation of identical volumes.
Nordström, Jonny; Kero, Tanja; Harms, Hendrik Johannes; Widström, Charles; Flachskampf, Frank A; Sörensen, Jens; Lubberink, Mark
2017-11-14
Quantitative measurement of myocardial blood flow (MBF) is of increasing interest in the clinical assessment of patients with suspected coronary artery disease (CAD). 15 O-water positron emission tomography (PET) is considered the gold standard for non-invasive MBF measurements. However, calculation of left ventricular (LV) volumes and ejection fraction (EF) is not possible from standard 15 O-water uptake images. The purpose of the present work was to investigate the possibility of calculating LV volumes and LVEF from cardiac-gated parametric blood volume (V B ) 15 O-water images and from first pass (FP) images. Sixteen patients with mitral or aortic regurgitation underwent an eight-gate dynamic cardiac-gated 15 O-water PET/CT scan and cardiac MRI. V B and FP images were generated for each gate. Calculations of end-systolic volume (ESV), end-diastolic volume (EDV), stroke volume (SV) and LVEF were performed with automatic segmentation of V B and FP images, using commercially available software. LV volumes and LVEF were calculated with surface-, count-, and volume-based methods, and the results were compared with gold standard MRI. Using V B images, high correlations between PET and MRI ESV (r = 0.89, p 0.86, p dynamic 15 O-water PET is feasible and shows good correlation with MRI. However, the analysis method is laborious, and future work is needed for more automation to make the method more easily applicable in a clinical setting.
Ma, Hongbin
2015-01-01
This book presents the fundamental fluid flow and heat transfer principles occurring in oscillating heat pipes and also provides updated developments and recent innovations in research and applications of heat pipes. Starting with fundamental presentation of heat pipes, the focus is on oscillating motions and its heat transfer enhancement in a two-phase heat transfer system. The book covers thermodynamic analysis, interfacial phenomenon, thin film evaporation, theoretical models of oscillating motion and heat transfer of single phase and two-phase flows, primary factors affecting oscillating motions and heat transfer, neutron imaging study of oscillating motions in an oscillating heat pipes, and nanofluid’s effect on the heat transfer performance in oscillating heat pipes. The importance of thermally-excited oscillating motion combined with phase change heat transfer to a wide variety of applications is emphasized. This book is an essential resource and learning tool for senior undergraduate, gradua...
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
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
J simplified assessment for cracked pipes and elbows in the RSE-M code
International Nuclear Information System (INIS)
Delliou, P.L.; Sermage, J.-P.; Gilles, P.; Marie, S.; Kayser, Y.; Barthelet, B.
2005-01-01
RSE-M Code provides rules and requirements for in-service inspection of French Pressurized Water Reactor power plant components. Non mandatory guidance is given in the Code for defect assessment in a wide range of configurations: surface cracked pipes and elbows under pressure, moment and thermal loading. The Code provides influence coefficients to calculate stress intensity factors in pipes and elbows containing semi-elliptical surface defects (circumferential or longitudinal). The J assessment method is based on the reference stress concept with two options for reference loads evaluation: 'CEP elastic plastic stress' and 'CLC modified limit load'. This paper presents an overview of all the formulations and namely the case of pipe-to elbow junctions. The paper provides also a description of the very large data base of 2D and 3D J elastic-plastic finite element calculations performed to establish and validate the formulations. Finally an applicability domain of the methods is given ensuring a conservative prediction of J. (authors)
An in-pipe mobile micromachine using fluid power. A mechanism adaptable to pipe diameters
International Nuclear Information System (INIS)
Yoshida, Kazuhiro; Yokota, Shinichi; Takahashi, Ken
2000-01-01
To realize micro maintenance robots for small diameter pipes of nuclear reactors and so on, high power in-pipe mobile micromachines have been required. The authors have proposed the bellows microactuator using fluid power and have tried to apply the actuators to in-pipe mobile micromachines. In the previous papers, some inchworm mobile machine prototypes with 25 mm in diameter are fabricated and the traveling performances are experimentally investigated. In this paper, to miniaturize the in-pipe mobile machine and to make it adaptable to pipe diameters, firstly, a simple rubber-tube actuator constrained with a coil-spring is proposed and the static characteristics are investigated. Secondly, a supporting mechanism which utilizes a toggle mechanism and is adaptable to pipe diameters is proposed and the supporting forces are investigated. Finally, an in-pipe mobile micromachine for pipe with 4 - 5 mm in diameter is fabricated and the maximum traveling velocity of 7 mm/s in both ahead and astern movements is experimentally verified. (author)
International Nuclear Information System (INIS)
Koyanagi, Ryoichi
1984-01-01
Many piping systems are supported by flexible structures or attached to thin shell walls so it is very important to consider the dynamic coupling effects between these systems in dynamic analysis. This paper presents a practical method of dynamic analysis of an individual system considering the dynamic coupling effects of coupled equipment-piping systems. In this method, dynamic responses are calculated by using the modal information which is obtained from the other analysis for associative structure. Analytical results for the complete model and of this method for an individual system are presented in the piping-supporting structure system and a piping-shell system. From the comparison of these results, it shows that this method is accurate, useful and economically applicable to the dynamic analysis of large model. (author)
Fracture mechanics assessment of thermal aged nuclear piping based on the Leak-Before-Break concept
Energy Technology Data Exchange (ETDEWEB)
Chen, Mingya, E-mail: chenmingya@cgnpc.com.cn [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Yu, Weiwei [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China); Qian, Guian [Paul Scherrer Institute, Nuclear Energy and Safety Department, Villigen PSI (Switzerland); Wang, Rongshan; Lu, Feng; Zhang, Guodong; Xue, Fei; Chen, Zhilin [Suzhou Nuclear Power Research Institute, Suzhou, Jiangsu Province (China)
2016-05-15
Highlights: • The effects of thermal aging on crack unstable tearing are studied. • The critical size of crack unstable tearing is calculated by different methods. • The critical failure models are compared. • The conservatism of J–T diagram is shown. - Abstract: The Leak-Before-Break (LBB) concept has been accepted to design the primary piping system of the pressurized water reactor (PWR). Due to thermal aging of long term operation, the cast stainless steels (CSSs) which are used for the primary piping of PWR, suffer a significant loss of fracture toughness, and as a consequence the safety margin of the thermal aged pipe decreases. Therefore, the aged piping should be analyzed and validated by the LBB concept. In this paper, elastic–plastic fracture mechanics (EPFM) assessments of the thermal aged piping are presented according to the LBB concept. The critical break size of crack unstable tearing is calculated by the EPFM method. The crack driving force diagram (J–a diagram), the stability assessment diagram (J–T diagram) and a numerical method are applied to calculate the critical crack size of crack break. The effects of thermal aging on the plastic limit load, J–T diagram, critical crack size of the EPFM and the critical failure mode are studied. The results show that the thermal aging effect decreases the maximum allowed J-integral at a certain ductile tearing modulus by more than 50% and it increases the flow stress and plastic limit load by 11.78%. The results based on the J–T diagram are about 40% conservative than those based on the direct numerical method for the high loading case. For the thermal aged piping, it is important to consider the competition failure modes between plastic collapse and unstable ductile tearing.
International Nuclear Information System (INIS)
Wright, Graham A.; McDade, Mark; Martin, William; Hutton, William
2002-01-01
Gated SPECT (GSPECT) offers the possibility of obtaining additional functional information from perfusion studies, including calculation of left ventricular ejection fraction (LVEF). The calculation of LVEF relies upon the identification of the endocardial surface, which will be affected by the spatial resolution and statistical noise in the reconstructed images. The aim of this study was to compare LVEFs and ventricular volumes calculated from GSPECT using six reconstruction filters. GSPECT and radionuclide ventriculography (RNVG) were performed on 40 patients; filtered back projection was used to reconstruct the datasets with each filter. LVEFs and volumes were calculated using the Cedars-Sinai QGS package. The correlation coefficient between RNVG and GSPECT ranged from 0.81 to 0.86 with higher correlations for smoother filters. The narrowest prediction interval was 11±2%. There was a trend towards higher LVEF values with smoother filters, the ramp filter yielding LVEFs 2.55±3.10% (p<0.001) lower than the Hann filter. There was an overall fall in ventricular volumes with smoother filters with a mean difference of 13.98±10.15 ml (p<0.001) in EDV between the Butterworth-0.5 and Butterworth-0.3 filters. In conclusion, smoother reconstruction filters lead to lower volumes and higher ejection fractions with the QGS algorithm, with the Butterworth-0.4 filter giving the highest correlation with LVEFs from RNVG. Even if the optimal filter is chosen the uncertainty in the measured ejection fractions is still too great to be clinically acceptable. (author)
Heat pipes and heat pipe exchangers for heat recovery systems
Energy Technology Data Exchange (ETDEWEB)
Vasiliev, L L; Grakovich, L P; Kiselev, V G; Kurustalev, D K; Matveev, Yu
1984-01-01
Heat pipes and heat pipe exchangers are of great importance in power engineering as a means of recovering waste heat of industrial enterprises, solar energy, geothermal waters and deep soil. Heat pipes are highly effective heat transfer units for transferring thermal energy over large distance (tens of meters) with low temperature drops. Their heat transfer characteristics and reliable working for more than 10-15 yr permit the design of new systems with higher heat engineering parameters.
Pipe restraints for nuclear power plants
International Nuclear Information System (INIS)
Keever, R.E.; Broman, R.; Shevekov, S.
1976-01-01
A pipe restraint for nuclear power plants in which a support member is anchored on supporting surface is described. Formed in the support member is a semicylindrical wall. Seated on the semicylindrical wall is a ring-shaped pipe restrainer that has an inner cylindrical wall. The inner cylindrical wall of the pipe restrainer encircles the pressurized pipe. In a modification of the pipe restraint, an arched-shaped pipe restrainer is disposed to overlie a pressurized pipe. The ends of the arch-shaped pipe restrainer are fixed to support members, which are anchored in concrete or to a supporting surface. A strap depends from the arch-shaped pipe restrainer. The pressurized pipe is supported by the depending strap
Energy Technology Data Exchange (ETDEWEB)
Tanaka, Y.; Kataoka, T.; Kokusho, T.; Yoshida, Y. (Central Research Institute of Electric Power Industry, Tokyo (Japan))
1994-03-21
In order to establish a practical method to evaluate safety of buried pipes, which were generally used as cable-protection pipes, at shallow depth and under vehicle loads, a series of experiments and analyses were performed. Based on the results, a simplified method to evaluate stress and deformation of the buried pipes in pavement were proposed. In the experiments, hard PVC pipes, light steel conduit pipes, and corrugated hard PE pipes in nominal diameters from 75 to 200 mm were applied as specimens for representative flexible pipes, and field load tests in actual size as well as laboratory load tests using soil vessels were carried out. The calculated results by the proposed method were compared to the experimental results, finding that the calculated values gave a satisfactory agreement with measured values which were obtained by the field tests. As a result, it was confirmed that the practical method to evaluate circumferential stresses at the top of pipes and flat-deflection ratios of small-sized buried Pipes induced by wheel loads under various conditions were presented. 20 refs., 17 figs., 6 tabs.
Energy Technology Data Exchange (ETDEWEB)
Cremer, C.D.; Lowry, W.; Cramer, E. [Science and Engineering Associates, Inc., Albuquerque, NM (United States)] [and others
1995-10-01
The U.S. Department of Energy`s nuclear facility decommissioning program needs to characterize radiological contamination inside piping systems before the pipe can be recycled, remediated, or disposed. Historically, this has been attempted using hand held survey instrumentation, surveying only the accessible exterior portions of pipe systems. Difficulty, or inability of measuring threshold surface contamination values, worker exposure, and physical access constraints have limited the effectiveness of this approach. Science and Engineering associates, Inc. under contract with the DOE Morgantown Energy Technology Center has developed and demonstrated the Pipe Explorer{trademark} system, which uses an inverting membrane to transport various characterization sensors into pipes. The basic process involves inverting (turning inside out) a tubular impermeable membrane under air pressure. A characterization sensor is towed down the interior of the pipe by the membrane.
Comparison of LMFBR piping response obtained using response spectrum and time history methods
International Nuclear Information System (INIS)
Hulbert, G.M.
1981-04-01
The dynamic response to a seismic event is calculated for a piping system using a response spectrum analysis method and two time history analysis methods. The results from the analytical methods are compared to identify causes for the differences between the sets of analytical results. Comparative methods are also presented which help to gain confidence in the accuracy of the analytical methods in predicting piping system structure response during seismic events
Baseline design of an OTEC pilot plantship. Volume C. Specifications
Energy Technology Data Exchange (ETDEWEB)
Glosten, L. R.; Bringloe, Thomas; Soracco, Dave; Fenstermacher, Earl; Magura, Donald; Sander, Olof; Richards, Dennis; Seward, Jerry
1979-05-01
Volume C is part of a three-volume report that presents a baseline engineering design of an Ocean Thermal Energy Conversion (OTEC) plantship. This volume provides the specifications for the hull, cold-water pipe, ship outfitting and machinery, OTEC power system, electrical system, and folded-tube heat exchangers.
Frictional pressure drop of gas liquid two-phase flow in pipes
International Nuclear Information System (INIS)
Shannak, Benbella A.
2008-01-01
Experiments of air water two-phase flow frictional pressure drop of vertical and horizontal smooth and relatively rough pipes were conducted, respectively. The result demonstrated that the frictional pressure drop increases with increasing relative roughness of the pipe. However, the influence of the relative roughness becomes more evident at higher vapour quality and higher mass flux. A new prediction model for frictional pressure drop of two-phase flow in pipes is proposed. The model includes a new definition of the Reynolds number and the friction factor of two-phase flow. The proposed model fits the presented experimental data very well, for vertical, horizontal, smooth and rough pipes. Therefore, the reproductive accuracy of the model is tested on the experimental data existing in the open literature and compared with the most common models. The statistical comparison, based on the Friedel's Data-Bank containing of about 16,000 measured data, demonstrated that the proposed model is the best overall agreement with the data. The model was tested for a wide range of flow types, fluid systems, physical properties and geometrical parameters, typically encountered in industrial piping systems. Hence, calculating based on the new approach is sufficiently accurate for engineering purposes
International Nuclear Information System (INIS)
Motohashi, Kazuhiko
2009-01-01
After an integration with ADLPipe, AutoPIPE V8i (ver.9.1) became the structural analysis program of plant piping system featured with analysis capability for the ASME NB Class 1 and JSME PPC-Class 2 piping codes including ASME NC Class 2 and ASME ND Class 3. This article described analysis capability for the JSME PPC-Class 2 piping code as well as new general features such as static analysis up to 100 thermal, 10 seismic and 10 wind load cases including different loading scenarios and pipe segment edit function: join, split, reverse and re-order segments. (T. Tanaka)
Surface Thermal Insulation and Pipe Cooling of Spillways during Concrete Construction Period
Directory of Open Access Journals (Sweden)
Wang Zhenhong
2014-01-01
Full Text Available Given that spillways adopt a hydraulic thin concrete plate structure, this structure is difficult to protect from cracks. The mechanism of the cracks in spillways shows that temperature stress is the major reason for cracks. Therefore, an effective way of preventing cracks is a timely and reasonable temperature-control program. Studies show that one effective prevention method is surface thermal insulation combined with internal pipe cooling. The major factors influencing temperature control effects are the time of performing thermal insulation and the ways of internal pipe cooling. To solve this problem, a spillway is taken as an example and a three-dimensional finite element program and pipe cooling calculation method are adopted to conduct simulation calculation and analysis on the temperature fields and stress fields of concretes subject to different temperature-control programs. The temperature-control effects are then compared. Optimization results show that timely and reasonable surface thermal insulation and water-flowing mode can ensure good temperature-control and anticrack effects. The method has reference value for similar projects.
A new method to butt weld pipes with laser at different angles
International Nuclear Information System (INIS)
Gualini, M.M.S.
1999-01-01
Laser butt welding of pipes at different angles may be cumbersome and may require very expensive tooling. The pipe size may not allow using the laser for large volume throughputs. We propose a rotary optical head composed by an adjustable focus lens system and two reflecting mirrors. The laser beam is bent at 90 deg. C. so that weld can be performed inwards outwards. The optic head design compensates the rotary backlash and vibrations, like a penta prism thus ensuring a perfect follow up of the weld track. The optic head can be inclined at 45 deg. C. to laser butt weld pipe each other at 90 deg. C. In this case the laser beam focus position is computer controlled in order to keep the focus point always on the elliptical weld profile. The paper covers theoretical and practical aspects of the proposed device. (author)
International Nuclear Information System (INIS)
Guangming, Xiao; Yanxia, Du; Yewei, Gui; Lei, Liu; Xiaofeng, Yang; Dong, Wei
2014-01-01
The theories of heat transfer, thermodynamics and fluid dynamics are employed to develop the coupled heat transfer analytical methods for the heat-pipe-cooled thermal protection structure (HPC TPS), and a three-dimensional numerical method considering the sonic limit of heat pipe is proposed. To verify the calculation correctness, computations are carried out for a typical heat pipe and the results agree well with experimental data. Then, the heat transfer characteristics and limitations of HPC TPS are mainly studied. The studies indicate that the use of heat pipe can reduce the temperature at high heat flux region of structure efficiently. However, there is a frozen startup period before the heat pipe reaching a steady operating state, and the sonic limit will be a restriction on the heat transfer capability. Thus, the effects of frozen startup must be considered for the design of HPC TPS. The simulation model and numerical method proposed in this paper can predict the heat transfer characteristics of HPC TPS quickly and exactly, and the results will provide important references for the design or performance evaluation of HPC TPS. - Highlights: • Numerical methods for the heat-pipe-cooled thermal protection structure are studied. • Three-dimensional simulation model considering sonic limit of heat pipe is proposed. • The frozen startup process of the embedded heat pipe can be predicted exactly. • Heat transfer characteristics of TPS and limitations of heat pipe are discussed
Development of piping evaluation diagram for LBB application to KNGR surge line
International Nuclear Information System (INIS)
Yoon, K. S.; Park, W. B.; Kim, J. M.; Choi, T. S.; Yang, J. S.; Park, C. Y.
1998-01-01
Plant specific data, such as pipe geometry, material properties and pipe loads, are required in order to evaluate Leak-Before-Break (LBB) applicability to piping systems in nuclear power plant under the construction. However, the existing method of LBB evaluation for KSNP's can not be used for newly developed nuclear plants such as Korean Next Generation Reactor (KNGR) which material properties is not available and LBB evaluation is required during design process. In order to solve this problem during developing process for KNGR surge line LBB Piping Evaluation Diagram (PED), which is independent of piping geometry and has a function of the loads applied in piping system, is developed in this paper. Also, in order to evaluate LBB applicability during construction process with only the comparative evaluation of material properties between actually used and expected, the expected changes of material properties are considered in the PED. The PED, therefore, can be used for quick LBB evaluation of KNGR surge line in the process of both design and construction. The benefit obtained by using the PED is : 1) to be able to very quickly confirm LBB applicability without calculating any leakage crack length for all concerned piping locations in the process of both iterative design for optimal routing and construction and 2) to save significantly a lot of computing times required for the corresponding LBB analyses
Directory of Open Access Journals (Sweden)
Nelson H. T. Lemes
2010-01-01
Full Text Available Analytical solutions of a cubic equation with real coefficients are established using the Cardano method. The method is first applied to simple third order equation. Calculation of volume in the van der Waals equation of state is afterwards established. These results are exemplified to calculate the volumes below and above critical temperatures. Analytical and numerical values for the compressibility factor are presented as a function of the pressure. As a final example, coexistence volumes in the liquid-vapor equilibrium are calculated. The Cardano approach is very simple to apply, requiring only elementary operations, indicating an attractive method to be used in teaching elementary thermodynamics.
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.
Two-phase flow structure in large diameter pipes
International Nuclear Information System (INIS)
Smith, T.R.; Schlegel, J.P.; Hibiki, T.; Ishii, M.
2012-01-01
Highlights: ► Local profiles of various quantities measured in large diameter pipe. ► Database for interfacial area in large pipes extended to churn-turbulent flow. ► Flow regime map confirms previous models for flow regime transitions. ► Data will be useful in developing interfacial area transport models for large pipes. - Abstract: Flow in large pipes is important in a wide variety of applications. In the nuclear industry in particular, understanding of flow in large diameter pipes is essential in predicting the behavior of reactor systems. This is especially true of natural circulation Boiling Water Reactor (BWR) designs, where a large-diameter chimney above the core provides the gravity head to drive circulation of the coolant through the reactor. The behavior of such reactors during transients and during normal operation will be predicted using advanced thermal–hydraulics analysis codes utilizing the two-fluid model. Essential to accurate two-fluid model calculations is reliable and accurate computation of the interfacial transfer terms. These interfacial transfer terms can be expressed as the product of one term describing the potential driving the transfer and a second term describing the available surface area for transfer, or interfacial area concentration. Currently, the interfacial area is predicted using flow regime dependent empirical correlations; however the interfacial area concentration is best computed through the use of the one-dimensional interfacial area transport equation (IATE). To facilitate the development of IATE source and sink term models in large-diameter pipes a fundamental understanding of the structure of the two-phase flow is essential. This understanding is improved through measurement of the local void fraction, interfacial area concentration and gas velocity profiles in pipes with diameters of 0.102 m and 0.152 m under a wide variety of flow conditions. Additionally, flow regime identification has been performed to
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.)
Measurement of transitional flow in pipes using ultrasonic flowmeters
Energy Technology Data Exchange (ETDEWEB)
Zheng-Gang, Liu; Guang-Sheng, Du; Zhu-Feng, Shao; Qian-Ran, He; Chun-Li, Zhou, E-mail: lzhenggang@sdu.edu.cn [School of Energy and Power Engineering, Qian-Fo-shan campus, Shandong University, Jinan City 250061, Shandong Province (China)
2014-10-01
The accuracy of an ultrasonic flowmeter depends on the ratio k of average profile velocity of pipe and average velocity of an ultrasonic propagation path. But there is no appropriate method of calculating k for transition flow. In this paper, the velocity field of the transition flow in a pipe is measured by particle image velocimetry. On this basis, the k of U-shaped and V-shaped ultrasonic flowmeter is obtained when Reynolds number is between 2000 and 20 000. It is shown that the k is constant when the Reynolds number is in the range of 2000–2400 and 5400–20 000, and the k decreases with the increasing of Re when the Reynolds number is 2400–5400. The results of study can be used to improve the measurement accuracy of ultrasonic flowmeters when flow is transition flow and can provide help for the study of pipe flow. (paper)
International Nuclear Information System (INIS)
Petruschke, W.; Strunk, G.
1987-01-01
The investigations according to the system identification show that the piping model using beam theory and flexibility factors according to the Karman theory are adequate for evaluating natural frequencies, mode shapes, static displacements and stresses. The same accuracy can be seen by comparing the piping response due to blowdown within the elastic range. The simplified elastic-plastic analysis in general overestimates the maximum amplitudes while the frequency content is not simulated very well. For practical purposes, it can be an adequate tool in many cases. The elastic-plastic analysis is the most expensive procedure but gives also the best results. The use of beam elements with multilinear moment-curvature relationships results in a good approximation for the global behaviour (displacements). The strains according to this theory only include the beam deformation modes
Directory of Open Access Journals (Sweden)
Bo Sun
2018-03-01
Full Text Available In the degradation process, the randomness and multiplicity of variables are difficult to describe by mathematical models. However, they are common in engineering and cannot be neglected, so it is necessary to study this issue in depth. In this paper, the copper bending pipe in seawater piping systems is taken as the analysis object, and the time-variant reliability is calculated by solving the interference of limit strength and maximum stress. We did degradation experiments and tensile experiments on copper material, and obtained the limit strength at each time. In addition, degradation experiments on copper bending pipe were done and the thickness at each time has been obtained, then the response of maximum stress was calculated by simulation. Further, with the help of one kind of Monte Carlo method we propose, the time-variant reliability of copper bending pipe was calculated based on the stochastic degradation process and interference theory. Compared with traditional methods and verified by maintenance records, the results show that the time-variant reliability model based on the stochastic degradation process proposed in this paper has better applicability in the reliability analysis, and it can be more convenient and accurate to predict the replacement cycle of copper bending pipe under seawater-active corrosion.
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.
The FTIR study of uranium oxides by the method of light pipe reflection spectroscopy
International Nuclear Information System (INIS)
Bao Zhu Yu; Hansen, W.N.
1988-01-01
Light pipe infrared reflection spectra of UO 2 , UO 3 , U 3 O 8 have been studied by using an FTIR spectrometer. The uranium oxide powders were ground to ensure fine particle size and distributed on the inner surface of a straight glass pipe with gold coating. The infrared beam from the inter-ferometer was focused into one end of the pipe at 45 0 incidence and then the transmitted beam was refocused by a pair of Cassegrainian type mirrors. The resultant spectra show the infrared characteristics of the ...-U-O-U-O-..., uranyl ion UO 2 2+ bond vibration and the active lattice vibrations predicted by group theory calculations. In comparison to the transmission spectra measured by authors or reported in literature, this 45 0 incident light pipe method as well as the previous light pipe method offer advantages of sensitivity, ease of acquisition and interpretation, and require a very small sample. It confirms the power of the light pipe method for studying powders and its special utility for the infrared studies of hazardous materials. (Author)
International Nuclear Information System (INIS)
Lim, E.Y.
1981-08-01
The PRAISE (Piping Reliability Analysis Including Seismic Events) computer code estimates the influence of earthquakes on the probability of failure at a weld joint in the primary coolant system of a pressurized water reactor. Failure, either a through-wall defect (leak) or a complete pipe severance (a large-LOCA), is assumed to be caused by fatigue crack growth of an as-fabricated interior surface circumferential defect. These defects are assumed to be two-dimensional and semi-elliptical in shape. The distribution of initial crack sizes is a function of crack depth and aspect ratio. Crack propagation rates are governed by a Paris-type relationship with separate RMS cyclic stress intensity factors for the depth and length. Both uniform through the wall and radial gradient thermal stresses are included in the calculation of the stress intensity factors. The failure probabilities are estimated by applying Monte Carlo methods to simulate the life histories of the selected weld joint. In order to maximize computational efficiency, a stratified sampling procedure is used to select the initial crack size. Hydrostatic proof test, pre-service inspection, and in-service inspection can be simulated. PRAISE treats the inter-arrival times of operating transients either as a constant or exponentially distributed according to observed or postulated rates. Leak rate and leak detection models are also included. The criterion for complete pipe severance is exceedance of a net section critical stress. Earthquakes of various intensity and arbitrary occurrence times can be modeled. PRAISE presently assumes that exactly one initial defect exists in the weld and that the earthquake of interest is the first earthquake experienced at the reactor
Directory of Open Access Journals (Sweden)
Kiswanto Gandjar
2017-01-01
Full Text Available The increase in the volume of rough machining on the CBV area is one of the indicators of increased efficiencyof machining process. Normally, this area is not subject to the rough machining process, so that the volume of the rest of the material is still big. With the addition of CC point and tool orientation to CBV area on a complex surface, the finishing will be faster because the volume of the excess material on this process will be reduced. This paper presents a method for volume calculation of the parts which do not allow further occurrence of the machining process, particulary for rough machining on a complex object. By comparing the total volume of raw materials and machining area volume, the volume of residual material,on which machining process cannot be done,can be determined. The volume of the total machining area has been taken into account for machiningof the CBV and non CBV areas. By using delaunay triangulation for the triangle which includes the machining and CBV areas. The volume will be calculated using Divergence(Gaussian theorem by focusing on the direction of the normal vector on each triangle. This method can be used as an alternative to selecting tothe rough machining methods which select minimum value of nonmachinable volume so that effectiveness can be achieved in the machining process.
Simulation of boiling flow in evaporator of separate type heat pipe with low heat flux
International Nuclear Information System (INIS)
Kuang, Y.W.; Wang, Wen; Zhuan, Rui; Yi, C.C.
2015-01-01
Highlights: • A boiling flow model in a separate type heat pipe with 65 mm diameter tube. • Nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux. • The two-phase heat transfer coefficient is less sensitive to the total mass flux. - Abstract: The separate type heat pipe heat exchanger is considered to be a potential selection for developing passive cooling spent fuel pool – for the passive pressurized water reactor. This paper simulates the boiling flow behavior in the evaporator of separate type heat pipe, consisting of a bundle of tubes of inner diameter 65 mm. It displays two-phase characteristic in the evaporation section of the heat pipe working in low heat flux. In this study, the two-phase flow model in the evaporation section of the separate type heat pipe is presented. The volume of fluid (VOF) model is used to consider the interaction between the ammonia gas and liquid. The flow patterns and flow behaviors are studied and the agitated bubbly flow, churn bubbly flow are obtained, the slug bubble is likely to break into churn slug or churn froth flow. In addition, study on the heat transfer coefficients indicates that the nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux, with the heat transfer coefficient being less sensitive to the total mass flux
Operation and management of United Central Piping LPG supply stations in Shenzhen
Energy Technology Data Exchange (ETDEWEB)
Lai Yankai
1997-11-01
Shenzhen has based its city gas development project on the eventual conversion to natural gas supply by way of central piping LPG supply stations. To fully exploit the potential gas supply capability of every central piping station and cut down the total running cost, we have been connecting the existing supply stations and their piping system into a network, which not only provided a more reliable gas supply performance, but can greatly simplify the evacuation of gas stations from the ever-expanding downtown areas to suburbs. Through this way, the periodic gas stock held by individual stations can be transferred to storage terminal or stations of enough holding capability; the supplying distance has been much lengthened and the gas volume held in the piping system increased; gas supply covered by small stations has been shifted to new and large stations. By linking these stations, we are able to provide pipeline LP gas supply for a large area, and in the same time lay down the pipeline infrastructure for the upcoming LNG supply so that an easy conversion to LNG supply can be secured as soon as the projected LNG terminal is put to service. (au)
Valdman, V. V.; Gridnev, S. O.
2017-10-01
The article examines into the vital issues of measuring and calculating the raw stock volumes in covered storehouses at mining and processing plants. The authors bring out two state-of-the-art high-technology solutions: 1 - to use the ground-based laser scanning system (the method is reasonably accurate and dependable, but costly and time consuming; it also requires the stoppage of works in the storehouse); 2 - to use the fundamentally new computerized stocktaking system in mine surveying for the ore mineral volume calculation, based on the profile digital images. These images are obtained via vertical projection of the laser plane onto the surface of the stored raw materials.
International Nuclear Information System (INIS)
Lucca, G.
1999-01-01
The paper presents an algorithm for the calculation, in the frequency domain, of the induced voltages and currents on a generic metallic pipe network exposed to the electromagnetic interference from a power line or an electrified railway line. By assuming as known the voltages and the currents on the inducing line, the algorithm may be subdivided into the following main steps: a) determination of the ideal electromotive force and current generators to be applied to the induced structure in order to represent the electromagnetic influence from the inducing line; b) modelling of the pipe network by means of a suitable equivalent electric network; c) calculation of voltages and currents on the induced network [it
International Nuclear Information System (INIS)
Guyette, M.; De Smet, M.
1995-01-01
In this paper we outline a methodology to assess the fatigue induced in piping systems submitted to thermal stratification. More specifically, the transformation from the measured outer wall temperature time histories to stress time histories in any point of the line is treated.By means of inverse transfer functions, the fluid temperature distribution is calculated from the outside wall temperatures measured in a limited number of temperature sections. Using direct transfer functions, the local stresses due to stratification may be determined as well as the pipe free curvatures and the pipe free axial strains. Using a finite beam element model of the line, the global response of the line (in terms of displacements or stresses) due to the applied curvatures, axial strains, end point displacements, internal pressure and possible contacts with the pipe environment may be determined.The method is illustrated for the surge lines of the Doel 2 and Doel 4 nuclear power plants. An excellent correlation is found between measured and calculated displacements. Typical stress time histories are shown for a plant cool down. ((orig.))
Energy Technology Data Exchange (ETDEWEB)
Weber, Jochen [BHR Hochdruck-Rohrleitungsbau GmbH, Essen (Germany); Lange, Heinz-Wilhelm [LISEGA AG, Zeven (Germany)
2009-05-15
This article reports on the current status of the EN 13480-3. Information was also provided on this subject at the 24th FDBR Conference on Pipe Technology in March 2009 in Magdeburg, Germany. The code was published in the year 2002 and was at first applied with some hesitation, but is now widely followed. To create a common European code 'Metallic industrial piping systems - Design and calculation', which has grown out of a variety of national guidelines, was an impressive feat by all those involved, which will not be depreciated or placed in question here. However, as is to be expected with such a new and complex set of rules, some weak points and errors became evident in its implementation. The FDBR working group 'Strength Calculation / Technical Codes' has identified these errors and elaborated corrective suggestions. On the occasion of the 22nd Plenary Session of the CEN/TC267 in November 2008 it was agreed that this year, 2009, an updated working version of the norm will first be created. This forms the basis of the planned publication of a revised edition of the EN 13480-3 in the first quarter of 2010. Unfortunately, at the moment very few European countries are participating in the revision of the code. (orig.)
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
Simplified piping analysis methods with inelastic supports
International Nuclear Information System (INIS)
Lin, C.W.; Romanko, A.D.
1986-01-01
Energy absorbing supports (EAS) which contain x-shaped plates or dampers with heavy viscous fluid can absorb a large amount of energy during vibratory motions. The response of piping systems supported by these types of energy absorbing devices can be markedly reduced as compared with ordinary supports using rigid rods, hangers or snubbers. In this paper, a simple multiple support response spectrum technique is presented, which would allow the energy dissipation nature of the EAS be factored in the piping response calculation. In the meantime, the effect of lower system frequencies due to the reduced support stiffness from local yielding is also included in the analysis. Numerical results obtained show that this technique is more conservative than the time history solution by an acceptable and realistic margin; and it has less than 10 percent of the computation cost
Evaluation of vibration and vibration fatigue life for small bore pipe in nuclear power plants
International Nuclear Information System (INIS)
Wang Zhaoxi; Xue Fei; Gong Mingxiang; Ti Wenxin; Lin Lei; Liu Peng
2011-01-01
The assessment method of the steady state vibration and vibration fatigue life of the small bore pipe in the supporting system of the nuclear power plants is proposed according to the ASME-OM3 and EDF evaluation methods. The GGR supporting pipe system vibration is evaluated with this method. The evaluation process includes the filtration of inborn sensitivity, visual inspection, vibration tests, allowable vibration effective velocity calculation and vibration stress calculation. With the allowable vibration effective velocity calculated and the vibration velocity calculated according to the acceleration data tested, the filtrations are performed. The vibration stress at the welding coat is calculated with the spectrum method and compared with the allowable value. The response of the stress is calculated with the transient dynamic method, with which the fatigue life is evaluated with the Miners linear accumulation model. The vibration stress calculated with the spectrum method exceeds the allowable value, while the fatigue life calculated from the transient dynamic method is larger than the designed life with a big safety margin. (authors)
Adjustment of pipe flow explicit friction factor equations for application to tube bundles
International Nuclear Information System (INIS)
Wiltz, Christopher L.; Bowen, Mike D.; Von Olnhausen, Wayne A.
2005-01-01
Full text of publication follows: The accurate determination of single phase friction losses or friction pressure drop in tube bundles is essential in the thermal-hydraulic analyses of components such as nuclear fuel assemblies, heat exchangers and steam generators. Such friction losses are normally calculated using a friction factor, f, along with the experimental observation that the friction pressure drop in a pipe is proportional to the dynamic pressure (1/2 ρV 2 ) of the flow: ΔP = 1/2 ρV 2 (fL/D). In this equation L is the pipe or tube bundle length and D is the hydraulic diameter of the pipe or tube bundle. The friction factor is normally calculated using one of a number of explicit friction factor equations. A significant amount of work has been accomplished in developing explicit friction factor equations. These explicit equations range from approximations, which were developed for ease of numerical evaluation, to those which are mathematically complex but yield very good fits to the test data. These explicit friction factor equations are based on a large experimental data base, nearly all of which comes from pipe flow geometry information, and have been historically applied to tube bundles. This paper presents an adjustment method which may be applied to various explicit friction factor equations developed for pipe flow to accurately predict the friction factor for tube bundles. The characteristic of the adjustment is based on experimental friction pressure loss data obtained by Framatome ANP through flow testing of a nuclear fuel assembly (tube bundle) at its Richland Test Facility (RTF). Through adjustment of previously developed explicit friction factor equations for pipe flow, the vast amount of historical development and experimentation in the area of single phase pipe flow friction loss may be incorporated into the evaluation of single phase friction losses within tube bundles. Comparisons of the application of one or more of the previously
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
Directory of Open Access Journals (Sweden)
S. Sugiharto
2014-08-01
Full Text Available Measurement of vapor flow in geothermal pipe faces great challenges due to fast fluids flow in high-temperature and high-pressure environment. In present study the flow rate measurement has been performed to characterization the geothermal vapor flow in a pipe. The experiment was carried out in a pipe which is connected to a geothermal production well, KMJ-14. The pipe has a 10” outside diameter and contains dry vapor at a pressure of 8 kg/cm2 and a temperature of 170 oC. Krypton-85 gas isotope (85Kr has been injected into the pipe. Three collimated radiation detectors positioned respectively at 127, 177 and 227m from injection point were used to obtain experimental data which represent radiotracer residence time distribution (RTD in the pipe. The last detector at the position of 227 m did not respond, which might be due to problems in cable connections. Flow properties calculated using mean residence time (MRT shows that the flow rate of the vapor in pipe is 10.98 m/s, much faster than fluid flow commonly found in various industrial process plants. Best fitting evaluated using dedicated software developed by IAEA expert obtained the Péclet number Pe as 223. This means that the flow of vapor of geothermal fluids in pipe is plug flow in character. The molecular diffusion coefficient is 0.45 m2/s, calculated from the axial dispersion model.
International Nuclear Information System (INIS)
Mochizuki, Masahito; Enomoto, Kunio; Okamoto, Noriaki; Saitoh, Hideyo; Hayashi, Eisaku.
1994-01-01
This paper studies welding residual stresses at the intersection of a small diameter pipe penetrating a thick plate. The pipe is welded to the plate, and Tungsten Innert Gas (TIG) cladding is melted on the inner surface of the pipe to protect it from stress corrosion cracking due to long operation in nuclear power plants. Stresses are calculated by heat conduction analysis and thermal elasto-plastic analysis, and also measured by strain gauges. Welding residual stresses are shown to have no corrosive influence on the inner pipe surface, and the stresses are compressed enough to protect the pipe against stress corrosion cracking on the outer surface. It was also studied to make clear the production mechanism of the residual stresses which were generated by welding processes at the pipe. (author)
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.
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.
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
A methodology to evaluate the fatigue life of flexible pipes
Energy Technology Data Exchange (ETDEWEB)
Sousa, Fernando J.M. de; Sousa, Jose Renato M. de; Siqueira, Marcos Q. de; Sagrilo, Luis V.S. [Coordenacao dos Programas de Pos-graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Lemos, Carlos Alberto D. de [Petroleo Brasileiro S.A. (PETROBRAS), Rio de Janeiro, RJ (Brazil)
2009-07-01
This paper focus on a methodology to perform the fatigue analysis of flexible pipes. This methodology employs functions that convert forces and moments obtained in global analyses into stresses. The stresses are then processed by well-known cycle counting methods, and S-N curves evaluate the damage at several points in the pipe cross-section. Palmgren-Miner linear damage hypothesis is assumed in order to calculate the accumulated fatigue damage. A parametric study on the fatigue life of a flexible pipe employing this methodology is presented. The main points addressed in the study are the influence of friction between layers in the results, the importance of evaluating the fatigue life in various points of the pipe cross-section and the effect of different mean stress levels. The obtained results suggest that the consideration of friction effects strongly influences the fatigue life of flexible risers and these effects have to be accounted both in the global and local analyses of the riser. Moreover, mean stress effects are also significant and at least 8 equally spaced wires in each analyzed section of the riser must be considered in fatigue analyses. (author)
Instationary discharge rates and shear factors in pipe ruptures
International Nuclear Information System (INIS)
Pana, P.
1976-01-01
The loads observed in ruptures of steam- or water-conducting pipes may occur as reactive forces on the pipes themselves or as jet forces on the structural components adjacent to the point of rupture. The present paper deals with the instationary acceleration phase directly after rupture. The general laws of conservation (mass, energy, momentum) may be used, but in their instationary form. This results in a system of partial differential equations which does not provide a comprehensive mathematical solution. However, since efficient electronic computer systems are available, difference methods are increasingly often used. Such calculations were carried out for water-steam as an ideal gas and under simplifying assumptions. (orig./AK) [de
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)
Small bore pipe acceptance criteria for watts bar nuclear plant Tennessee Valley Authority
International Nuclear Information System (INIS)
Sun, W.S.; Lee, R.L.; Kalyanan, N.
1991-01-01
Small bore pipe (≤2 inches NPS) is traditionally analyzed by simplified techniques using Cook Book approach, which yield conservative results. However, reconciliation of these systems for as-built condition where the original criteria is observed to have been exceeded (or due to additions etc.) generally becomes a time consuming and expensive operation since a rigorous computer aided analysis or a detailed hand calculation becomes necessary. The acceptance criteria in this paper can be effectively used in such cases. The approach involves utilizing basic engineering principles and plant specific parameters (such as earthquake spectra) to estimate the system response such as pipe stress due to various loading conditions, piping frequency, support and anchor loads, valve acceleration etc
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)
Directory of Open Access Journals (Sweden)
I E. Lobanov
2017-01-01
Full Text Available Objectives. The aim of present work was to carry out mathematical modelling of heat transfer with symmetrical heating in flat channels and round pipes with rough walls.Methods. The calculation was carried out using the L'Hôpital-Bernoulli's method. The solution of the problem of intensified heat transfer in a round tube with rough walls was obtained using the Lyon's integral.Results. Different from existing theories, a methodology of theoretical computational heat transfer determination for flat rough channels and round pipes with rough walls is developed on the basis of the principle of full viscosity superposition in a turbulent boundary layer. The analysis of the calculated heat transfer and hydroresistivity values for flat rough channels and round rough pipes shows that the increase in heat transfer is always less than the corresponding increase in hydraulic resistance, which is a disadvantage as compared to channels with turbulators, with all else being equal. The results of calculating the heat transfer for channels with rough walls in an extended range of determinant parameters, which differ significantly from the corresponding data for the channels with turbulators, determine the level of heat exchange intensification.Conclusion. An increase in the calculated values of the relative average heat transfer Nu/NuGL for flat rough channels and rough pipes with very high values of the relative roughness is significantly contributed by both an increase in the relative roughness height and an increase in the Reynolds number Re. In comparison with empirical dependencies, the main advantage of solutions for averaged heat transfer in rough flat channels and round pipes under symmetrical thermal load obtained according to the developed theory is that they allow the calculation of heat exchange in rough pipes to be made in the case of large and very large relative heights of roughness protrusions, including large Reynolds numbers, typical for pipes
Fully plastic crack opening analyses of complex-cracked pipes for Ramberg-Osgood materials
International Nuclear Information System (INIS)
Jeong, Jae Uk; Choi, Jae Boong; Huh, Nam Su; Kim, Yun Jae
2016-01-01
The plastic influence functions for calculating fully plastic Crack opening displacement (COD) of complex-cracked pipes were newly proposed based on systematic 3-dimensional (3-D) elastic-plastic Finite element (FE) analyses using Ramberg-Osgood (R-O) relation, where global bending moment, axial tension and internal pressure are considered separately as a loading condition. Then, crack opening analyses were performed based on GE/EPRI concept by using the new plastic influence functions for complex-cracked pipes made of SA376 TP304 stainless steel, and the predicted CODs were compared with FE results based on deformation plasticity theory of tensile material behavior. From the comparison, the confidence of the proposed fully plastic crack opening solutions for complex-cracked pipes was gained. Therefore, the proposed engineering scheme for COD estimation using the new plastic influence functions can be utilized to estimate leak rate of a complex-cracked pipe for R-O material.
A study on the heat transfer characteristics of a self-oscillating heat pipe
International Nuclear Information System (INIS)
Yoon, Seok Hun; Oh, Cheol; Choi, Jae Hyuk
2002-01-01
In this paper, the heat transfer characteristics of a self-oscillating heat pipe are experimentally investigated for the effect of various working fluid fill charge ratios and heat loads. The characteristics of temperature oscillations of the working fluid are also analysed based on chaotic dynamics. The heat pipe is composed of a heating section, a cooling section and an adiabatic section, and has a 0.002m internal diameter, a 0.34m length in each turn and consists of 19 turns. The heating and the cooling portion of each turn has a length of 70mm. A series of experiments was carried out to measure the temperature distributions and the pressure variations of the heat pipe. Furthermore, heat transfer performance, effective thermal conductivity, boiling heat transfer and condensation heat transfer coefficients are calculated for various operating conditions. Experimental results show the efficacy of this type of heat pipe
International Nuclear Information System (INIS)
Tiari, Saeed; Qiu, Songgang; Mahdavi, Mahboobe
2015-01-01
Highlights: • A finned heat pipe-assisted latent heat thermal energy storage system is studied. • The effects of heat pipes spacing and fins geometrical features are investigated. • Smaller heat pipes spacing and longer fins improve the melting rate. • The optimal heat pipe and fin arrangements are determined. - Abstract: In the present study, the thermal characteristics of a finned heat pipe-assisted latent heat thermal energy storage system are investigated numerically. A transient two-dimensional finite volume based model employing enthalpy-porosity technique is implemented to analyze the performance of a thermal energy storage unit with square container and high melting temperature phase change material. The effects of heat pipe spacing, fin length and numbers and the influence of natural convection on the thermal response of the thermal energy storage unit have been studied. The obtained results reveal that the natural convection has considerable effect on the melting process of the phase change material. Increasing the number of heat pipes (decreasing the heat pipe spacing) leads to the increase of melting rate and the decrease of base wall temperature. Also, the increase of fin length results in the decrease of temperature difference within the phase change material in the container, providing more uniform temperature distribution. It was also shown that number of the fins does not have a significant effect on the performance of the system
Pipe support program at Pickering
International Nuclear Information System (INIS)
Sahazizian, L.A.; Jazic, Z.
1997-01-01
This paper describes the pipe support program at Pickering. The program addresses the highest priority in operating nuclear generating stations, safety. We present the need: safety, the process: managed and strategic, and the result: assurance of critical piping integrity. In the past, surveillance programs periodically inspected some systems, equipment, and individual components. This comprehensive program is based on a managed process that assesses risk to identify critical piping systems and supports and to develop a strategy for surveillance and maintenance. The strategy addresses all critical piping supports. Successful implementation of the program has provided assurance of critical piping and support integrity and has contributed to decreasing probability of pipe failure, reducing risk to worker and public safety, improving configuration management, and reducing probability of production losses. (author)
Energy Technology Data Exchange (ETDEWEB)
Wang, Zhong-Min, E-mail: wangzhongm@xaut.edu.cn; Liu, Yan-Zhuang
2016-03-15
Highlights: • We investigate the transverse vibration of FGM pipe conveying fluid. • The FGM pipe conveying fluid can be classified into two cases. • The variations between the frequency and the power law exponent are obtained. • “Case 1” is relatively more reasonable than “case 2”. - Abstract: Problems related to the transverse vibration of pipe conveying fluid made of functionally graded material (FGM) are addressed. Based on inside and outside surface material compositions of the pipe, FGM pipe conveying fluid can be classified into two cases. It is hypothesized that the physical parameters of the material along the direction of the pipe wall thickness change in the simple power law. A differential equation of motion expressed in non-dimensional quantities is derived by using Hamilton's principle for systems of changing mass. Using the assuming modal method, the pipe deflection function is expanded into a series, in which each term is expressed to admissible function multiplied by generalized coordinate. Then, the differential equation of motion is discretized into the two order differential equations expressed in the generalized coordinates. Based on symplectic elastic theory and the introduction of dual system and dual variable, Hamilton's dual equations are derived, and the original problem is reduced to eigenvalue and eigenvector problem in the symplectic space. Finally, a symplectic method is employed to analyze the vibration and stability of FGM pipe conveying fluid. For a clamped–clamped FGM pipe conveying fluid in “case 1” and “case 2”, the dimensionless critical flow velocity for first-mode divergence and the critical coupled-mode flutter flow velocity are obtained, and the variations between the real part and imaginary part of dimensionless complex frequency and fluid velocity, mass ratio and the power law exponent (or graded index, volume fraction) for FGM pipe conveying fluid are analyzed.
International Nuclear Information System (INIS)
Wang, Zhong-Min; Liu, Yan-Zhuang
2016-01-01
Highlights: • We investigate the transverse vibration of FGM pipe conveying fluid. • The FGM pipe conveying fluid can be classified into two cases. • The variations between the frequency and the power law exponent are obtained. • “Case 1” is relatively more reasonable than “case 2”. - Abstract: Problems related to the transverse vibration of pipe conveying fluid made of functionally graded material (FGM) are addressed. Based on inside and outside surface material compositions of the pipe, FGM pipe conveying fluid can be classified into two cases. It is hypothesized that the physical parameters of the material along the direction of the pipe wall thickness change in the simple power law. A differential equation of motion expressed in non-dimensional quantities is derived by using Hamilton's principle for systems of changing mass. Using the assuming modal method, the pipe deflection function is expanded into a series, in which each term is expressed to admissible function multiplied by generalized coordinate. Then, the differential equation of motion is discretized into the two order differential equations expressed in the generalized coordinates. Based on symplectic elastic theory and the introduction of dual system and dual variable, Hamilton's dual equations are derived, and the original problem is reduced to eigenvalue and eigenvector problem in the symplectic space. Finally, a symplectic method is employed to analyze the vibration and stability of FGM pipe conveying fluid. For a clamped–clamped FGM pipe conveying fluid in “case 1” and “case 2”, the dimensionless critical flow velocity for first-mode divergence and the critical coupled-mode flutter flow velocity are obtained, and the variations between the real part and imaginary part of dimensionless complex frequency and fluid velocity, mass ratio and the power law exponent (or graded index, volume fraction) for FGM pipe conveying fluid are analyzed.
Heat pipe applications workshop report
International Nuclear Information System (INIS)
Ranken, W.A.
1978-04-01
The proceedings of the Heat Pipe Applications Workshop, held at the Los Alamos Scientific Laboratory October 20-21, 1977, are reported. This workshop, which brought together representatives of the Department of Energy and of a dozen industrial organizations actively engaged in the development and marketing of heat pipe equipment, was convened for the purpose of defining ways of accelerating the development and application of heat pipe technology. Recommendations from the three study groups formed by the participants are presented. These deal with such subjects as: (1) the problem encountered in obtaining support for the development of broadly applicable technologies, (2) the need for applications studies, (3) the establishment of a heat pipe technology center of excellence, (4) the role the Department of Energy might take with regard to heat pipe development and application, and (5) coordination of heat pipe industry efforts to raise the general level of understanding and acceptance of heat pipe solutions to heat control and transfer problems
International Nuclear Information System (INIS)
Jutas, A.; Kersiene, N.; Leisis, V.; Ziliukas, A.
2003-01-01
The use of different standards for the pressure vessels design educes some phenomena what by different methodologies are estimated. In Lithuania the newly found interest is driven by the desire to develop the design methodologies that are fundamentally based on principles of pressure vessels design and resource evaluation. In this article the differences of the Eurocode prEN13445-3 and Russian SNiP 2.05.06-85 and stress analysis of the pipes is shown too. These standards designed for calculation and assessment of exploitation loads of new pressure vessels. Usually corroded pipes are replaced then new and old pipe together welded. These pipelines for the pressure tests are very sensitive and re-calculation of limited pressure with assessment of geometry' changes of the corroded pipes is needful. (author)
International Nuclear Information System (INIS)
Ari Satmoko; Edi Karyanta; Dedy Haryanto; Abdul Hafid; Sudarno; Kussigit Santosa; Pinitoyo, A.; Demon Handoyo
2003-01-01
One of several steps in industrial plant construction is preparing piping layout drawing. In this drawing, pipe and all other pieces such as instrumentation, equipment, structure should be modeled A software called CAPD was developed to replace and to behave as piping drafter or designer. CAPD was successfully developed by adding both subprogram CHKUPIPE and CHKMANV. The first subprogram can check and gives warning if there is piping pocket in the piping system. The second can identify valve position and then check whether valve can be handled by operator hand The main program CAPD was also successfully modified in order to be capable in limiting the maximum length of straight pipe. By limiting the length, piping flexibility can be increased. (author)
Filtering algorithm for radial displacement measurements of a dented pipe
International Nuclear Information System (INIS)
Hojjati, M.H.; Lukasiewicz, S.A.
2008-01-01
Experimental measurements are always affected by some noise and errors caused by inherent inaccuracies and deficiencies of the experimental techniques and measuring devices used. In some fields, such as strain calculations in a dented pipe, the results are very sensitive to the errors. This paper presents a filtering algorithm to remove noise and errors from experimental measurements of radial displacements of a dented pipe. The proposed filter eliminates the errors without harming the measured data. The filtered data can then be used to estimate membrane and bending strains. The method is very effective and easy to use and provides a helpful practical measure for inspection purposes
Piping engineering and operation
International Nuclear Information System (INIS)
1993-01-01
The conference 'Piping Engineering and Operation' was organized by the Institution of Mechanical Engineers in November/December 1993 to follow on from similar successful events of 1985 and 1989, which were attended by representatives from all sectors of the piping industry. Development of engineering and operation of piping systems in all aspects, including non-metallic materials, are highlighted. The range of issues covered represents a balance between current practices and implementation of future international standards. Twenty papers are printed. Two, which are concerned with pressurized pipes or steam lines in the nuclear industry, are indexed separately. (Author)
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.
Wijdeven, S.M.J.; Vaessen, O.H.B.; Hees, van A.F.M.; Olsthoorn, A.F.M.
2005-01-01
Dead wood is recognized as one of the key indicators for sustainable forest management and biodiversity. Accurate assessments of dead wood volume are thus necessary. In this study New volume models were designed based on actual volume measurements of coarse woody debris. The New generic model
Conservatism inherent to simplified qualification techniques used for piping steady state vibration
International Nuclear Information System (INIS)
Olson, D.E.; Smetters, J.L.
1983-01-01
This paper examines some of the qualification techniques currently used by the power industry, including the techniques specified in a recently issued standard related to this subject (ANSI/ASME OM-3, Requirements for Preoperational and Initial Startup Vibration Testing of Nuclear Power Plant Piping Systems). Several methods are used to demonstrate the amount of conservatism inherent in these techniques. Allowable limits calculated by the use of simplified techniques are compared to limits calculated by more detailed computer analysis. A portion of a reactor feedwater piping system along with the results of a piping vibration monitoring program recently completed in a nuclear power plant are used as case studies. The limits determined by the use of simplified criteria are also compared to limits determined empirically through the use of strain gauges. The simple beam analogies that use vibrational displacement as acceptance criteria were found to be conservative for all the examples studied. However, when velocity was used as a criterion, it was not always conservative. Simplified techniques that result in displacement allowables appear to be the most viable method of qualifying piping vibrations. Quantities referred to in the paper are cited in British units throughout. These may be converted to the International System of Units (SI) as follows: 1 foot=0.3048 meter; 1 inch=0.0254 meter=1,000 mils; 1 psi=6,894 pascals; and 1 inch/second=0.0254 meter/second. (orig.)
Li, Chang; Wang, Qing; Shi, Wenzhong; Zhao, Sisi
2018-05-01
The accuracy of earthwork calculations that compute terrain volume is critical to digital terrain analysis (DTA). The uncertainties in volume calculations (VCs) based on a DEM are primarily related to three factors: 1) model error (ME), which is caused by an adopted algorithm for a VC model, 2) discrete error (DE), which is usually caused by DEM resolution and terrain complexity, and 3) propagation error (PE), which is caused by the variables' error. Based on these factors, the uncertainty modelling and analysis of VCs based on a regular grid DEM are investigated in this paper. Especially, how to quantify the uncertainty of VCs is proposed by a confidence interval based on truncation error (TE). In the experiments, the trapezoidal double rule (TDR) and Simpson's double rule (SDR) were used to calculate volume, where the TE is the major ME, and six simulated regular grid DEMs with different terrain complexity and resolution (i.e. DE) were generated by a Gauss synthetic surface to easily obtain the theoretical true value and eliminate the interference of data errors. For PE, Monte-Carlo simulation techniques and spatial autocorrelation were used to represent DEM uncertainty. This study can enrich uncertainty modelling and analysis-related theories of geographic information science.
Synchrotron radiation leakage from the B-factory beam pipe
International Nuclear Information System (INIS)
Jenkins, T.M.; Nelson, W.R.; Ipe, N.
1990-01-01
The high-energy ring (HER) of the B-Factory, running at an energy of 9 GeV, generates the synchrotron spectrum when applied to a ring with the PEP bending radius. The B-Factory HER may also run at 12 GeV, producing the harder spectrum. Depending upon beam-pipe material and thickness, some of this radiation may escape and deposit energy in the surrounding material. This was originally pointed out in PEP-109 during the initial design of PEP, and subsequently verified by measurements at both PEP and PETRA at DESY. Of concern to the B Factory is magnet insulation, though other adjacent materials such as wire insulation and cooling water hoses are even more radiosensitive. Radiation damage to magnets is a function of the type of material used in the potting compound. The PEP magnets, which hopefully can be used for the high-energy ring of the B-Factory, are insulated with an epoxy composed of DER-332, DER-732, NMA and aluminum oxide. It is estimated that this epoxy compound should tolerate doses near the 10 10 rad range. To summarize the results of these calculations, 0.87 cm of copper is needed at the point of maximum dose from 12-GeV operation in order to reach the dose criterion if a rectangular beam pipe is used. The copper needs to be only 0.7-cm thick for an octagonal beam pipe and the same energy. For 9-GeV operation, an octagonal copper pipe needs only to be 0.25-cm thick. An octagonal aluminum pipe that is 0.5-cm thick also needs 0.3 cm of lead on the outside to reach the same criterion for 12-GeV operation. For 9-GeV operation, the aluminum pipe still requires a lead liner
Seismic analysis response factors and design margins of piping systems
International Nuclear Information System (INIS)
Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.
1985-01-01
The objective of the simplified methods project of the Seismic Safety Margins Research Program is to develop a simplified seismic risk methodology for general use. The goal is to reduce seismic PRA costs to roughly 60 man-months over a 6 to 8 month period, without compromising the quality of the product. To achieve the goal, it is necessary to simplify the calculational procedure of the seismic response. The response factor approach serves this purpose. The response factor relates the median level response to the design data. Through a literature survey, we identified the various seismic analysis methods adopted in the U.S. nuclear industry for the piping system. A series of seismic response calculations was performed. The response factors and their variabilities for each method of analysis were computed. A sensitivity study of the effect of piping damping, in-structure response spectra envelop method, and analysis method was conducted. In addition, design margins, which relate the best-estimate response to the design data, are also presented
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
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)
Characterization of pipes, drain lines, and ducts using the pipe explorer system
International Nuclear Information System (INIS)
Cremer, C.D.; Kendrick, D.T.; Cramer, E.
1997-01-01
As DOE dismantles its nuclear processing facilities, site managers must employ the best means of disposing or remediating hundreds of miles of potentially contaminated piping and duct work. Their interiors are difficult to access, and in many cases even the exteriors are inaccessible. Without adequate characterization, it must be assumed that the piping is contaminated, and the disposal cost of buried drain lines can be on the order of $1,200/ft and is often unnecessary as residual contamination levels often are below free release criteria. This paper describes the program to develop a solution to the problem of characterizing radioactive contamination in pipes. The technical approach and results of using the Pipe Explorer trademark system are presented. The heart of the system is SEA's pressurized inverting membrane adapted to transport radiation detectors and other tools into pipes. It offers many benefits over other pipe inspection approaches. It has video and beta/gamma detection capabilities, and the need for alpha detection has been addressed through the development of the Alpha Explorer trademark. These systems have been used during various stages of decontamination and decommissioning of DOE sites, including the ANL CP-5 reactor D ampersand D. Future improvements and extensions of their capabilities are discussed
Application of a drift-flux model to flashing in straight pipes
International Nuclear Information System (INIS)
Hirt, C.W.; Romero, N.C.
1975-06-01
A new computer program, SOLA-OF, has been written to solve the unsteady, two-dimensional equations of motion for a two-phase mixture. The equations solved are based on the drift-flux approximation and include a phase transition model and a general drift velocity calculation. The SOLA-DF code is used for a study of the blowdown of straight pipes initially filled with water at high temperature and pressure. Computed results are presented that show the relative importance of phase transition rates, pipe friction, drift velocity magnitude, and other model variations. The computed results are also compared with experimental data. 7 references. (auth)
Crack resistance of austenitic pipes with circumferential through-wall cracks
International Nuclear Information System (INIS)
Foerster, K.; Grueter, L.; Setz, W.; Bhandari, S.; Debaene, J.P.; Faidy, C.; Schwalbe, K.H.
1993-01-01
For monotonously increasing load the correct evaluation of the crack resistance properties of a structure is essential for safety analyses. Considerable attention has been given to the through-wall case, since this is generally believed to be the controlling case with regard to complete pipe failure. The maximum load conditions for circumferential crack growth in pipes under displacement-controlled loadings has been determined. The need for crack resistance curves, measured on circumferentially through-wall cracked straight pipes of austenitic stainless steel 316L under bending, is emphasized by the limitation in the data range on small specimens and by the differences in the procedures. To answer open questions and to improve calculational methods a joint fracture mechanics program is being performed by Electricite de France, Novatome and Siemens-Interatom. The working program contains experimental and theoretical investigations on the applicability of small-specimen data to real structures. 10 refs., 10 figs., 4 tabs
Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems
International Nuclear Information System (INIS)
Strong, B.R. Jr.; Baschiere, R.J.
1978-01-01
The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)
International Nuclear Information System (INIS)
Wibisono; Sugiharto
2004-01-01
The computer program, namely scaling pipe has been constructed to support the work of radiation monitoring either utilizing of sealed or unsealed source as well. The paper describes the performance of that computer program which is able to display numerical data coming from rate-meter to become graphical data and its ability to store data in form of software file with text format. The stored file can be performed by other software therefore the data can be calculated and analyzed. This computer program has been applied to measure the thickness of steel pipe with diameter of 3 inch which is machined with the different of 0.1 mm for each 50 mm length of the pipe. In this research the gamma radiation originated from 200 mCi of 241 Am source has been passed through the center of the pipe and the dose rate at the opposite of the source is measured. The attenuation coefficient calculated from that measurement is 0.0944 mm -1 with the deviation factor of 44 cps/mm. (author)
Flexible mobile robot system for smart optical pipe inspection
Kampfer, Wolfram; Bartzke, Ralf; Ziehl, Wolfgang
1998-03-01
Damages of pipes can be inspected and graded by TV technology available on the market. Remotely controlled vehicles carry a TV-camera through pipes. Thus, depending on the experience and the capability of the operator, diagnosis failures can not be avoided. The classification of damages requires the knowledge of the exact geometrical dimensions of the damages such as width and depth of cracks, fractures and defect connections. Within the framework of a joint R&D project a sensor based pipe inspection system named RODIAS has been developed with two partners from industry and research institute. It consists of a remotely controlled mobile robot which carries intelligent sensors for on-line sewerage inspection purpose. The sensor is based on a 3D-optical sensor and a laser distance sensor. The laser distance sensor is integrated in the optical system of the camera and can measure the distance between camera and object. The angle of view can be determined from the position of the pan and tilt unit. With coordinate transformations it is possible to calculate the spatial coordinates for every point of the video image. So the geometry of an object can be described exactly. The company Optimess has developed TriScan32, a special software for pipe condition classification. The user can start complex measurements of profiles, pipe displacements or crack widths simply by pressing a push-button. The measuring results are stored together with other data like verbal damage descriptions and digitized images in a data base.
International Nuclear Information System (INIS)
Wang, L.W.; Wang, R.Z.; Lu, Z.S.; Chen, C.J.
2006-01-01
The split heat pipe type compound adsorption ice maker for fishing boats not only has the advantage of large volume cooling density but also has the advantage of less power consumption and high heat transfer performance. The available heat pipe media for the split heat pipe type compound adsorption ice maker, which are methanol, acetone and water are studied and compared in this paper, and the heat pipe medium of water shows the better performance for the reason of its stable heating and cooling process and high heat transfer performance. Considering the waste heat recovered from the diesel engine on fishing boats varies when the velocity of the fishing boat changes, the refrigeration performances at the condition of different values of heating power are studied while water is used as the heat pipe medium. Results show that the cooling power, as while as COP and SCP decrease when the heating power decreases. The highest COP and SCP are 0.41 and 731 W/kg, respectively, at the highest heating power of 4.2 kW, and the values decrease by 22% and 33%, respectively, when the heating power decreases by 15%. The values decrease by 32% and 51%, respectively, when the heating power decreases by 30%. The performance of the adsorption ice maker for the fishing boat with the 6160A type diesel engine is estimated, and the results show that the cooling power and ice productivity are as high as 5.44 kW and 1032 kg ice per day, respectively, even if the recovered waste heat decreases by 30% compared with the normal value. It can satisfy the ice requirements of such a fishing boat
Thermal fatigue crack growth in mixing tees nuclear piping - An analytical approach
International Nuclear Information System (INIS)
Radu, V.
2009-01-01
The assessment of fatigue crack growth due to cyclic thermal loads arising from turbulent mixing presents significant challenges, principally due to the difficulty of establishing the actual loading spectrum. So-called sinusoidal methods represent a simplified approach in which the entire spectrum is replaced by a sine-wave variation of the temperature at the inner pipe surface. The need for multiple calculations in this process has lead to the development of analytical solutions for thermal stresses in a pipe subject to sinusoidal thermal loading, described in previous work performed at JRC IE Petten, The Netherlands, during the author's stage as seconded national expert. Based on these stress distributions solutions, the paper presents a methodology for assessment of thermal fatigue crack growth life in mixing tees nuclear piping. (author)
FLANGE-ORNL, Flanged Pipe Joint Stress Analysis, Internal Pressure, Moment Loads, Temperature
International Nuclear Information System (INIS)
Rodabaugh, E.C.; Moore, S.E.
1979-01-01
1 - Description of problem or function: FLANGE-ORNL calculates appropriate loads, stresses, and displacements for the flanges, bolts, and gaskets that comprise a flanged piping joint for internal pressure or moment loading on the pipe, temperature difference between the flange hub and ring, and variations in bolt load that result from pressure, hub-ring temperature gradient and/or bolt-ring temperature differences. Flanges considered may be tapered-hub, straight or blind. 2 - Method of solution: The solution is based on discontinuity analysis and the theory of plates and shells
Leak-before-break behaviour of nuclear piping systems
International Nuclear Information System (INIS)
Bartholome, G.; Wellein, R.
1992-01-01
The general concept for break preclusion of nuclear piping systems in the FRG consists of two main prerequisites: Basic safety; independent redundancies. The leak-before-break behaviour is open of these redundancies and will be verified by fracture mechanics. The following items have to be evaluated: The growth of detected and postulated defects must be negligible in one life time of the plant; the growth behaviour beyond design (i.e. multiple load collectives are taken into account) leads to a stable leak; This leakage of the piping must be detected by an adequate leak detection system long before the critical defect size is reached. The fracture mechanics calculations concerning growth and instability of the relevant defects and corresponding leakage areas are described in more detail. The leak-before-break behaviour is shown for two examples of nuclear piping systems in pressurized water reactors: main coolant line of SIEMENS-PWR 1300 MW (ferritic material, diameter 800 mm); surge line of Russian WWER 440 (austenitic material, diameter 250 mm). The main results are given taking into account the relevant leak detection possibilities. (author). 9 refs, 9 figs
A new approach to solve elastoplastic dynamic piping problems
International Nuclear Information System (INIS)
Leite de Andrade, J.E.; Guerreiro Ribeiro, S.V.
1981-01-01
A new method to perform the elastoplastic dynamic analysis of pipes is presented here, in which the pipe is analysed as a beam, and a bilinear eleastic-plastic behavior for the material is assumed. Pipe whip restraints are simulated as spring of bilinear elastic-plastic behavior with the provision for considering viscous damping. A numerical method was implemented in which plastic strain is treated as equivalent applied (force or moment) excitations, reducing the elastoplastic analysis of the structure to an elastic analysis of the same structure with a set of additional applied excitations. So the stiffness matrix and the eigenvectors do not vary with time. This procedure allows the response of the system to be computed by using dynamic influence coefficients, which are calculated from the elastic solution. For those structures whose dynamic elastic solutions are known in closed form, the present scheme seems to be very attractive, e.g., simple supported and cantilever beams. For those structures with unknown analytical elastic solutions, the finite element method will provide them. (orig./GL)
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.)
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
CFD simulations of a bubbly flow in a vertical pipe
International Nuclear Information System (INIS)
Krepper, E.
2000-01-01
Even at the very simple conditions of two phase flow in a vertical pipe, strong 3D effects are observed. The distribution of the gas phase over the cross section varies significantly between the different flow patterns, which are known for the vertical two-phase flow. The air water flow in a vertical tube having a diameter of 50 mm and a length of about 3 m was investigated in steady state tests for different liquid and gas superficial velocities. Several two phase flow measuring techniques were used. Applying a wire mesh sensor, developed in FZR, the void fraction could be determined over the whole cross section of the pipe. The working principle is based on the measurement of the local instantaneous conductivity of the two-phase mixture. At the investigated flow velocities, the rate of the image acquisition is sufficient to record the same bubble several times. This enables to determine bubble diameter distributions. Applying two similar wire mesh sensors with a distance of 50 mm one above the other, the influence of the wire mesh to the flow could be investigated. No essential disturbances of the two-phase flow by the mesh could be found for the investigated flow regimes. Performing an auto correlation between the signals of both sensors, also profiles of the gas velocity were determined. In the CFD code CFX-4.2 several two-phase flow models were available. Using the code, volume fraction profiles were calculated and compared to the measured results for bubble flow regimes, to investigate the capability of these models (see also Krepper and Prasser [4] (1999)). (orig.)
Leak-before-break analysis of thermally aged nuclear pipe under different bending moments
Energy Technology Data Exchange (ETDEWEB)
Lv, Xuming; Li, Shilei; Zhang, Hailong; Wang, Yanli; Wang, Xitao [University of Science and Technology Beijing, Beijing (China); Wang, Zhaoxi [CPI Nuclear Power Institute, Beijing (China); Xue, Fei [Suzhou Nuclear Power Research Institute, Suzhou (China)
2015-10-15
Cast duplex stainless steels are susceptible to thermal aging during long-term service at temperatures ranging from 280°C to 450°C. To analyze the effect of thermal aging on leak-before-break (LBB) behavior, three-dimensional finite element analysis models were built for circumferentially cracked pipes. Based on the elastic–plastic fracture mechanics theory, the detectable leakage crack length calculation and J-integral stability assessment diagram approach were carried out under different bending moments. The LBB curves and LBB assessment diagrams for unaged and thermally aged pipes were constructed. The results show that the detectable leakage crack length for thermally aged pipes increases with increasing bending moments, whereas the critical crack length decreases. The ligament instability line and critical crack length line for thermally aged pipes move downward and to the left, respectively, and unsafe LBB assessment results will be produced if thermal aging is not considered. If the applied bending moment is increased, the degree of safety decreases in the LBB assessment.
Heat pipes in modern heat exchangers
International Nuclear Information System (INIS)
Vasiliev, Leonard L.
2005-01-01
Heat pipes are very flexible systems with regard to effective thermal control. They can easily be implemented as heat exchangers inside sorption and vapour-compression heat pumps, refrigerators and other types of heat transfer devices. Their heat transfer coefficient in the evaporator and condenser zones is 10 3 -10 5 W/m 2 K, heat pipe thermal resistance is 0.01-0.03 K/W, therefore leading to smaller area and mass of heat exchangers. Miniature and micro heat pipes are welcomed for electronic components cooling and space two-phase thermal control systems. Loop heat pipes, pulsating heat pipes and sorption heat pipes are the novelty for modern heat exchangers. Heat pipe air preheaters are used in thermal power plants to preheat the secondary-primary air required for combustion of fuel in the boiler using the energy available in exhaust gases. Heat pipe solar collectors are promising for domestic use. This paper reviews mainly heat pipe developments in the Former Soviet Union Countries. Some new results obtained in USA and Europe are also included
DEFF Research Database (Denmark)
Mohammadi, Soma; Bojesen, Carsten
2015-01-01
the temperature in DH systems. The main focus is on modeling transient heat transfer in pipe networks regarding the time delays between the heat supply unit and the consumers, the heat loss in the pipe networks and the consumers’ dynamic heat loads. A pseudo-dynamic approach is adopted and also the implicit...... district heating networks [DHN] characteristics. This paper is presenting a new developed model, which reflects the thermo-dynamic behavior of DHN. It is designed for tree network topologies. The purpose of the model is to serve as a basis for applying a variety of scenarios towards lowering...... finite element method is applied to simulate transient temperature changes in pipe networks. The model is calculating time series data related to supply temperature to the DHN from heat production units, heat loads and return temperature related to each consumer to calculate dynamic temperature changes...
Analysis of the main causes of failures in the Atucha I PWR moderator circuit branch piping
International Nuclear Information System (INIS)
Porto, J.; Sarmiento, G.S.
1983-01-01
From 1977 to 1979 four through cracks were detected in the auxiliary connection of the moderator piping with the coolant circuit in the PWR Atucha I Nuclear Plant. The failures were observed to occur systematically in the same place of the pipe, where mechanical stresses were detected experimentally and thermal stresses were calculated based on temperature values measured on the pipe. The temperature field in steady state conditions as well as during thermal shocks was modelled by finite element codes, and the corresponding thermal stresses were than numerically calculated. Considering those thermal and mechanical solicitations, a crack propagation analysis based on the elastoplastic fracture mechanics and the finite element method is now being developed. Among other causes such as fatigue corrosion and vibrations, the results of the analysis show that the most preponderant factors determining the cracking are mechanical stress, thermal stress and thermal fatigue
Engineering design aspects of the heat-pipe power system
Capell, B. M.; Houts, M. G.; Poston, D. I.; Berte, M.
1997-01-01
The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations.
Uncoupled and coupled analysis of a large HDR pipe
International Nuclear Information System (INIS)
Muller, W.C.
1987-01-01
The main differences are in the structural response. There is no clear tendency that a coupled calculation will result in lower amplitudes of the structural response, but it can be seen from the results that there is a typical difference between coupled and uncoupled analysis which increases with time. This increase is mainly due to the fact that in a coupled analysis the speed of sound of the fluid and the eigenmodes of the piping system structure are lower than in the uncoupled analysis. Coupled and uncoupled piping transient analyses show similar results for the fluiddynamic data. The differences are less than 10% and as long as the fluid is in the two phase domain they can almost be neglected
Engineering design aspects of the heat-pipe power system
International Nuclear Information System (INIS)
Capell, B.M.; Houts, M.G.; Poston, D.I.; Berte, M.
1997-10-01
The Heat-pipe Power System (HPS) is a near-term, low-cost space power system designed at Los Alamos that can provide up to 1,000 kWt for many space nuclear applications. The design of the reactor is simple, modular, and adaptable. The basic design allows for the use of a variety of power conversion systems and reactor materials (including the fuel, clad, and heat pipes). This paper describes a project that was undertaken to develop a database supporting many engineering aspects of the HPS design. The specific tasks discussed in this paper are: the development of an HPS materials database, the creation of finite element models that will allow a wide variety of investigations, and the verification of past calculations
Leak before break piping evaluation diagram
International Nuclear Information System (INIS)
Fabi, R.J.; Peck, D.A.
1994-01-01
Traditionally Leak Before Break (LBB) has been applied to the evaluation of piping in existing nuclear plants. This paper presents a simple method for evaluating piping systems for LBB during the design process. This method produces a piping evaluation diagram (PED) which defines the LBB requirements to the piping designer for use during the design process. Several sets of LBB analyses are performed for each different pipe size and material considered in the LBB application. The results of this method are independent of the actual pipe routing. Two complete LBB evaluations are performed to determine the maximum allowable stability load, one evaluation for a low normal operating load, and the other evaluation for a high normal operating load. These normal operating loads span the typical loads for the particular system being evaluated. In developing the allowable loads, the appropriate LBB margins are included in the PED preparation. The resulting LBB solutions are plotted as a set of allowable curves for the maximum design basis load, such is the seismic load versus the normal operating load. Since the required margins are already accounted for in the LBB PED, the piping designer can use the diagram directly with the results of the piping analysis and determine immediately if the current piping arrangement passes LBB. Since the LBB PED is independent of pipe routing, changes to the piping system can be evaluated using the existing PED. For a particular application, all that remains is to confirm that the actual materials and pipe sizes assumed in creating the particular design are built into the plant
International Nuclear Information System (INIS)
Ohta, Takahiro; Kamo, Kazuhiko; Asada, Seiji; Terasaki, Toshio
2009-01-01
The new process called L-SIP (outer surface irradiated Laser Stress Improvement Process) is developed to improve the tensile residual stress of the inner surface near the butt welded joints of pipes in the compression stress. The temperature gradient occurs in the thickness of pipes in heating the outer surface rapidly by laser beam. By the thermal expansion difference between the inner surface and the outer surface, the compression stress occurs near the inner surface of pipes. In this paper, the theoretical equation for the temperature distributions of pipes heated by moving rectangular Gauss distribution heat source on the outer surface is derived. The temperature histories of pipes calculated by theoretical equation agree well with FEM analysis results. According to the theoretical equation, the controlling parameters of temperature distributions and histories are q/2a y , vh, a x /h and a y /h, where q is total heat input, a y is heat source length in the axial direction, a x is Gaussian radius of heat source in the hoop direction, ν is moving velocity, and h is thickness of the pipe. The essential variables for L-SIP, which are defined on the basis of the measured temperature histories on the outer surface of the pipe, are Tmax, F 0 =kτ 0 /h 2 , vh, W Q and L Q , where Tmax is maximum temperature on the monitor point of the outer surface, k is thermal diffusivity coefficient, τ 0 is the temperature rise time from 100degC to maximum temperature on the monitor point of the outer surface, W Q is τ 0 x ν, and L Q is the uniform temperature length in the axial direction. It is verified that the essential variables for L-SIP match the controlling parameters by the theoretical equation. (author)
46 CFR 61.15-5 - Steam piping.
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...
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
Influence of pipe material and surfaces on sulfide related odor and corrosion in sewers.
Nielsen, Asbjørn Haaning; Vollertsen, Jes; Jensen, Henriette Stokbro; Wium-Andersen, Tove; Hvitved-Jacobsen, Thorkild
2008-09-01
Hydrogen sulfide oxidation on sewer pipe surfaces was investigated in a pilot scale experimental setup. The experiments were aimed at replicating conditions in a gravity sewer located immediately downstream of a force main where sulfide related concrete corrosion and odor is often observed. During the experiments, hydrogen sulfide gas was injected intermittently into the headspace of partially filled concrete and plastic (PVC and HDPE) sewer pipes in concentrations of approximately 1,000 ppm(v). Between each injection, the hydrogen sulfide concentration was monitored while it decreased because of adsorption and subsequent oxidation on the pipe surfaces. The experiments showed that the rate of hydrogen sulfide oxidation was approximately two orders of magnitude faster on the concrete pipe surfaces than on the plastic pipe surfaces. Removal of the layer of reaction (corrosion) products from the concrete pipes was found to reduce the rate of hydrogen sulfide oxidation significantly. However, the rate of sulfide oxidation was restored to its background level within 10-20 days. A similar treatment had no observable effect on hydrogen sulfide removal in the plastic pipe reactors. The experimental results were used to model hydrogen sulfide oxidation under field conditions. This showed that the gas-phase hydrogen sulfide concentration in concrete sewers would typically amount to a few percent of the equilibrium concentration calculated from Henry's law. In the plastic pipe sewers, significantly higher concentrations were predicted because of the slower adsorption and oxidation kinetics on such surfaces.
Use of the modal superposition technique for piping system blowdown analyses
International Nuclear Information System (INIS)
Ware, A.G.; Macek, R.W.
1983-01-01
A standard method of solving for the seismic response of piping systems is the modal superposition technique. Only a limited number of structural modes are considered (typically those up to 33 Hz in the U.S.), since the effect on the calculated response due to higher modes is generally small, and the method can result in considerable computer cost savings over the direct integration method. The modal superposition technique has also been applied to piping response problems in which the forcing functions are due to fluid excitation. Application of the technique to this case is somewhat more difficult, because a well defined cutoff frequency for determining structural modes to be included has not been established. This paper outlines a method for higher mode corrections, and suggests methods to determine suitable cutoff frequencies for piping system blowdown analyses. A numerical example illustrates how uncorrected modal superposition results can produce erroneous stress results
International Nuclear Information System (INIS)
Hong, S.Y.; Yeater, M.L.
1985-01-01
This paper discusses stress intensity factor calculations and fatigue analysis for a PWR primary coolant piping system. The influence function method is applied to evaluate ASME Code Section XI Appendix A ''analysis of flaw indication'' for the application to a PWR primary piping. Results of the analysis are discussed in detail. (orig.)
Analysis of piping response to thermal and operational transients
International Nuclear Information System (INIS)
Wang, C.Y.
1987-01-01
The reactor piping system is an extremely complex three-dimensional structure. Maintaining its structural integrity is essential to the safe operation of the reactor and the steam-supply system. In the safety analysis, various transient loads can be imposed on the piping which may cause plastic deformation and possible damage to the system, including those generated from hydrodynamic wave propagations, thermal and operational transients, as well as the seismic events. At Argonne National Laboratory (ANL), a three-dimensional (3-D) piping code, SHAPS, aimed for short-duration transients due to wave propagation, has been developed. Since 1984, the development work has been shifted to the long-duration accidents originating from the thermal and operational transient. As a result, a new version of the code, SHAPS-2, is being established. This paper describes many features related to this later development. To analyze piping response generated from thermal and operational transients, a 3-D implicit finite element algorithm has been developed for calculating the hoop, flexural, axial, and torsional deformations induced by the thermomechanical loads. The analysis appropriately accounts for stresses arising from the temperature dependence of the elastic material properties, the thermal expansion of the materials, and the changes in the temperature-dependent yield surface. Thermal softening, failure, strain rate, creep, and stress ratching can also be considered
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)
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.
FOREST STEM VOLUME CALCULATION USING AIRBORNE LIDAR DATA
Directory of Open Access Journals (Sweden)
I. Büyüksalih
2017-05-01
Full Text Available Airborne LiDAR data have been collected for the city of Istanbul using Riegl laser scanner Q680i with 400 kHz and an average flight height of 600 m. The flight campaign was performed by a helicopter and covers an area of 5400 km2. According to a flight speed of 80 knot a point density of more than 16 points/m2 and a laser footprint size of 30 cm could be achieved. As a result of bundle adjustment, in total, approximately 17,000 LAS files with the file size of 500 m by 700 m have been generated for the whole city. The main object classes Ground, Building, Vegetation (medium, high were derived from these LAS files using the macros in Terrasolid software. The forest area under investigation is located northwest of the city of Istanbul, main tree species occurring in the test site are pine (pinus pinaster, oak (quercus and beech (fagus. In total, 120 LAS tiles covering the investigation area have been analysed using the software IMPACT of Joanneum Research Forschungsgesellschaft, Graz, Austria. First of all, the digital terrain model (DTM and the digital surface models (DSM were imported and converted into a raster file from the original laser point clouds with a spatial resolution of 50 cm. Then, a normalized digital surface model (nDSM was derived as the difference between DSM and the DTM. Tree top detection was performed by multi – resolution filter operations and tree crowns were segmented by a region growing algorithms develop specifically for this purpose. Breast Height Diameter (BHD was calculated on the base of tree height and crown areas derived from image segmentation applying allometric functions found in literature. The assessment of stem volume was then calculated as a function of tree height and BHD. A comparison of timber volume estimated from the LiDAR data and field plots measured by the Forest Department of Istanbul showed R2 of 0.46. The low correlation might arise either from the low quality of the field plots or
Estimation of leak rate through circumferential cracks in pipes in nuclear power plants
Directory of Open Access Journals (Sweden)
Jai Hak Park
2015-04-01
Full Text Available The leak before break (LBB concept is widely used in designing pipe lines in nuclear power plants. According to the concept, the amount of leaking liquid from a pipe should be more than the minimum detectable leak rate of a leak detection system before catastrophic failure occurs. Therefore, accurate estimation of the leak rate is important to evaluate the validity of the LBB concept in pipe line design. In this paper, a program was developed to estimate the leak rate through circumferential cracks in pipes in nuclear power plants using the Henry–Fauske flow model and modified Henry–Fauske flow model. By using the developed program, the leak rate was calculated for a circumferential crack in a sample pipe, and the effect of the flow model on the leak rate was examined. Treating the crack morphology parameters as random variables, the statistical behavior of the leak rate was also examined. As a result, it was found that the crack morphology parameters have a strong effect on the leak rate and the statistical behavior of the leak rate can be simulated using normally distributed crack morphology parameters.
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)
International Nuclear Information System (INIS)
1984-08-01
IGSCC in BWR piping is occurring owing to a combination of material, environment, and stress factors, each of which can affect both the initiation of a stress-corrosion crack and the rate of its subsequent propagation. In evaluating long-term solutions to the problem, one needs to consider the effects of each of the proposed remedial actions. Mitigating actions to control IGSCC in BWR piping must be designed to alleviate one or more of the three synergistic factors: sensitized material, the convention BWR environment, and high tensile stresses. Because mitigating actions addressing each of these factors may not be fully effective under all anticipated operating conditions, mitigating actions should address two and preferably all three of the causative factors; e.g., material plus some control of water chemistry, or stress reversal plus controlled water chemistry
Joustra, Sjoerd D; van der Plas, Evelyn M; Goede, Joery; Oostdijk, Wilma; Delemarre-van de Waal, Henriette A; Hack, Wilfried W M; van Buuren, Stef; Wit, Jan M
2015-06-01
Accurate calculations of testicular volume standard deviation (SD) scores are not currently available. We constructed LMS-smoothed age-reference charts for testicular volume in healthy boys. The LMS method was used to calculate reference data, based on testicular volumes from ultrasonography and Prader orchidometer of 769 healthy Dutch boys aged 6 months to 19 years. We also explored the association between testicular growth and pubic hair development, and data were compared to orchidometric testicular volumes from the 1997 Dutch nationwide growth study. The LMS-smoothed reference charts showed that no revision of the definition of normal onset of male puberty - from nine to 14 years of age - was warranted. In healthy boys, the pubic hair stage SD scores corresponded with testicular volume SD scores (r = 0.394). However, testes were relatively small for pubic hair stage in Klinefelter's syndrome and relatively large in immunoglobulin superfamily member 1 deficiency syndrome. The age-corrected SD scores for testicular volume will aid in the diagnosis and follow-up of abnormalities in the timing and progression of male puberty and in research evaluations. The SD scores can be compared with pubic hair SD scores to identify discrepancies between cell functions that result in relative microorchidism or macroorchidism. ©2015 Foundation Acta Paediatrica. Published by John Wiley & Sons Ltd.
International Nuclear Information System (INIS)
Rahman, S.; Brust, F.; Ghadiali, N.; Krishnaswamy, P.; Wilkowski, G.; Choi, Y.H.; Moberg, F.; Brickstad, B.
1995-04-01
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 impingement 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. These leak rates depend on the crack-opening area of a 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 11. This study was requested by the NRC to review, evaluate, and refine current analytical models for crack-opening-area analyses of pipes with circumferential through-wall cracks. Twenty-five pipe experiments were analyzed to determine the accuracy of the predictive models. Several practical aspects of crack-opening such as; crack-face pressure, off-center cracks, restraint of pressure-induced bending, cracks in thickness transition regions, weld residual stresses, crack-morphology models, and thermal-hydraulic analysis, were also investigated. 140 refs., 105 figs., 41 tabs