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Sample records for feedwater nozzle piping

  1. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning

    Energy Technology Data Exchange (ETDEWEB)

    Shah, V.N.; Ware, A.G.; Porter, A.M.

    1997-03-01

    This report presents a review of nuclear industry efforts to manage thermal fatigue, flow-accelerated corrosion, and water hammer damage to pressurized water reactor (PWR) feedwater nozzles, piping, and feedrings. The review includes an evaluation of design modifications, operating procedure changes, augmented inspection and monitoring programs, and mitigation, repair and replacement activities. Four actions were taken: (a) review of field experience to identify trends of operating events, (b) review of technical literature, (c) visits to PWR plants and a PWR vendor, and (d) solicitation of information from 8 other countries. Assessment of field experience is that licensees have apparently taken sufficient action to minimize feedwater nozzle cracking caused by thermal fatigue and wall thinning of J-tubes and feedwater piping. Specific industry actions to minimize the wall-thinning in feedrings and thermal sleeves were not found, but visual inspection and necessary repairs are being performed. Assessment of field experience indicates that licensees have taken sufficient action to minimize steam generator water hammer in both top-feed and preheat steam generators. Industry efforts to minimize multiple check valve failures that have allowed backflow of steam from a steam generator and have played a major role in several steam generator water hammer events were not evaluated. A major finding of this review is that analysis, inspection, monitoring, mitigation, and replacement techniques have been developed for managing thermal fatigue and flow-accelerated corrosion damage to feedwater nozzles, piping, and feedrings. Adequate training and appropriate applications of these techniques would ensure effective management of this damage.

  2. 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

  3. Ultrasonic pattern recognition study of feedwater nozzle inner radius indication

    International Nuclear Information System (INIS)

    Yoneyama, H.; Takama, S.; Kishigami, M.; Sasahara, T.; Ando, H.

    1983-01-01

    A study was made to distinguish defects on feed-water nozzle inner radius from noise echo caused by stainless steel cladding by using ultrasonic pattern recognition method with frequency analysis technique. Experiment has been successfully performed on flat clad plates and nozzle mock-up containing fatigue cracks and the following results which shows the high capability of frequency analysis technique are obtained

  4. Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Tsai, Y.L., E-mail: YLTsai@itri.org.t [National Chiao Tung University, Mechanical Engineering Department, 1001 TaHsueh Road, HsinChu, Taiwan 30010 (China); Industrial Technology Research Institute (ITRI), 195 Chung Hsing Rd., Sec.4 Chu Tung, HsinChu, Taiwan 310 (China); Wang, Li. H. [Industrial Technology Research Institute (ITRI), 195 Chung Hsing Rd., Sec.4 Chu Tung, HsinChu, Taiwan 310 (China); Fan, T.W. [Industrial Technology Research Institute (ITRI), 195 Chung Hsing Rd., Sec.4 Chu Tung, HsinChu, Taiwan 310 (China); Chung Hua University, Department of Civil Engineering and Engineering Informatics, 707, Sec.2, WuFu Rd., HsinChu, Taiwan 300 (China); Ranganath, Sam [Industrial Technology Research Institute (ITRI), 195 Chung Hsing Rd., Sec.4 Chu Tung, HsinChu, Taiwan 310 (China); Wang, C.K. [Taiwan Power Company (TPC), No.242, Sec. 3, Roosevelt Rd., Zhongzheng District, Taipei City 100, Taiwan (China); Chou, C.P. [National Chiao Tung University, Mechanical Engineering Department, 1001 TaHsueh Road, HsinChu, Taiwan 30010 (China)

    2010-01-15

    Inspection of the weld between the feedwater nozzle and the safe end at one Taiwan BWR showed axial indications in the Alloy 182 weld. The indication was sufficiently deep that continued operation could not be justified considering the crack growth for one cycle. A weld overlay was decided to implement for restoring the structural margin. This study reviews the cracking cases of feedwater nozzle welds in other nuclear plants, and reports the lesson learned in the engineering project of this weld overlay repair. The overlay design, the FCG calculation and the stress analysis by FEM are presented to confirm that the Code Case structural margins are met. The evaluations of the effect of weld shrinkage on the attached feedwater piping are also included. A number of challenges encountered in the engineering and analysis period are proposed for future study.

  5. Simulation of the behaviour of a servo actuated check valve upon rupture of the feedwater pipe

    International Nuclear Information System (INIS)

    Lucas, A.M. de; Perezagua, R.L.; Rosa, B. de la; Sanz, J.

    1995-01-01

    The steam generator replacement programme at Almaraz NPP, provides for the installation of a replacement damped non-return valve for the feedwater system. the function of this valve is to protect the steam generator in the event of a rupture in the feedwater pipe. Sudden closure of the check valve, against the flow and following rupture of the feedwater pipe, causes overpressure in the valve which is transmitted to the steam generator nozzle. It is therefore necessary to know this when designing the internal systems of the steam generator. Using the RELAP5/MODE3 code, it has been possible to simulate the dynamic behaviour of a check valve upon rupture of a feedwater pipe postulated outside the containment. The calculation model has been applied to different types of check valve. (Author)

  6. 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

  7. Thermal-hydraulics of PGV-4 water volume during damage of the feedwater collector nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Titov, V.F. [OKB Gidropress (Russian Federation); Notaros, U.; Lenkei, I. [NPP Paks (Hungary)

    1995-12-31

    A number of VVER-440 plants has experienced the distributing nozzles of feedwater collector being damaged due to corrosion-erosion wearing. Such phenomenon could result in feedwater redistribution within the SG inventory with undesirable consequences. The collector with damaged nozzles has to be replaced but a certain time is needed for the preparatory works. The main objective of the investigation conducted is to assess if the safe operation of SG is possible before collector replacement. It was shown that the nozzle damage as observed did not result in the dangerous disturbances of thermobydraulics as compared with the conditions existing at the initial period of operation. (orig.).

  8. Device for detecting the water leak within a feedwater nozzle in water cooled reactors

    International Nuclear Information System (INIS)

    Hattori, Tsunekazu.

    1984-01-01

    Purpose: To enable exact recognition and detection for the state of water leak. Constitution: The detection device comprises a thermocouple disposed to the outer surface of a feedwater nozzle, a distortion meter for detecting the change in the outer diameter of a nozzle and an acoustic emission generator disposed to the inside of the nozzle for generating a signal upon temperature change. These sensors previously monitor the states during normal operation, and thus detect the change in each of the states upon occurrence of water leakage to issue an alarm. (Kamimura, M.)

  9. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  10. Design criteria for piping and nozzles program

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  11. Heat exchanger in which the feedwater is injected into the upper part by a feed pipe open downing

    International Nuclear Information System (INIS)

    Poussin, C.

    1994-01-01

    The feedwater is injected into the upper part of the annular space formed between the exterior casing and the tube bundle envelope of the steam generator. The feedwater is injected via a feed pipe with an inversed T-piece and which opens into an overflowing channel open down. 10 figs

  12. 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.

  13. Condensation driven water hammer studies for feedwater distribution pipe

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland); Pullinen, J. [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K. [EDO Gidropress (Russian Federation)

    1997-12-31

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

  14. Condensation driven water hammer studies for feedwater distribution pipe

    International Nuclear Information System (INIS)

    Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J.; Pullinen, J.; Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K.

    1997-01-01

    Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.)

  15. BWR feedwater nozzle and control rod drive return line nozzle cracking: resolution of generic technical activity A-10. Technical report

    International Nuclear Information System (INIS)

    Snaider, R.

    1980-11-01

    This report summarizes work performed by the NRC staff in the resolution of Generic Technical Activity A-10, 'BWR Nozzle Cracking'. Generic Technical Activity A-10 is one of the generic technical subjects designated as 'unresolved safety issues' pursuant to Section 210 of the Energy Reorganization Act of 1974. The report describes the technical issues, the technical studies and analyses performed by the General Electric Company and the NRC staff, the staff's technical positions based on these studies, and the staff's plans for continued implementation of its technical positions. It also provides information for further work to resolve the non-destructive examination issue

  16. Replacement of the feedwater pipe system in reactor building outside containment at the nuclear power plant Philippsburg; Austausch der Speisewasserleitung im Reaktorgebaeude ausserhalb SHB im Kernkraftwerk Philippsburg I

    Energy Technology Data Exchange (ETDEWEB)

    Kessler, A. [Energie-Versorgung Schwaben AG, Stuttgart (Germany); Labes, M. [Siemens AG Unternehmensbereich KWU, Offenbach am Main (Germany); Schwenk, B. [Kernkraftwerk Philippsburg GmbH (Germany)

    1998-11-01

    After full replacement of the feedwater pipe system during the inspection period in 1997, combined with a modern materials, manufacturing and analysis concept, the entire pipe system of the water/steam cycle in the reactor building of KKP 1 now consists of high-toughness materials. The safety level of the entire plant has been increased by leaving aside postulation of F2 breaks in the reactor building and providing for protection against 0.1 leaks. Based on fluid-dynamic calculations for the cases of pump failure and pipe break, as well as pipe system calculations in 5 extensive calculation cycles, about 130 documents were filed for inspection and approval (excluding preliminary test documents on restraints). Points of main interest for safety analysis in this context were the optimised closing performance of the 3rd check valves and the integrity of the nozzle region at the RPV. (oirg./CB) [Deutsch] Durch den Restaustausch der Speisewasserleitungen in der Revision 1997, verbunden mit einem modernen Werkstoff-, Fertigungs- und Nachweiskonzept, sind im Reaktorgebaeude von KKP 1 in den Hauptleitungen des Wasser-Dampf-Kreislaufes nur noch hochzaehe Werkstoffe eingesetzt. Durch den Verzicht auf das Postulat von 2F-Bruechen im Reaktorgebaeude und durch die Auslegung gegen 0,1F-Lecks wird das Sicherheitsniveau der Anlage insgesamt gesteigert. Ausgehend von fluiddynamischen Berechnungen fuer Pumpenausfall und Rohrbruch sowie Rohrsystem-Berechnungen in 5 umfangreichen Berechnungskreisen wurden fuer die Genehmigung und Begutachtung ca. 130 Unterlagen (ohne Halterungs-Vorpruefunterlagen) eingereicht und vom Gutachter geprueft. Schwerpunkte der Nachweisfuehrung waren die Optimierung des Schliessverhaltens der 3. Rueckschlagarmaturen sowie der Integritaetsnachweis des RDB-Anschlusses. (orig./MM)

  17. Feedwater heater

    International Nuclear Information System (INIS)

    Murata, Shigeto; Minato, Akihiko; Yokomizo, Osamu; Masuhara, Yasuhiro.

    1991-01-01

    The present invention concerns a feedwater heater for a BWR type reactor. A cylinder is fit into the lower portion of a drain inlet pipe, to which drain water inflows from a turbine, and a disk is disposed to the lower end of the cylinder vertically to the axis of the cylinder, to constitute a drain water dispersing mechanism. Drain water inflown from the drain inlet pipe is fallen in the cylinder and collides against the disk. The collided drain water is splashed horizontally by its kinetic energy to reach the heat transfer pipe and conducts heat exchange. In this case, the drain water is converted into fine droplets by the collision against the disk and scattered in a wide range in the heater. As a result, sensible heat in the drain water can be transferred to feedwater effectively. Then, even the heat energy of the drain water can be utilized effectively for heat exchange, to improve the heat exchange efficiency. (I.N.)

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  19. Feedwater control device for reactor pressure vessels

    International Nuclear Information System (INIS)

    Oonuma, Takeshi.

    1982-01-01

    Purpose: To prevent the generation of thermal stresses at the junction between a clean-up water pipe and a feedwater pipe. Constitution: Hot water containing impurities in a pressure vessel is caused to flow by a recycling pump through a heat exchanger, a cooler and a clean-up desalter and again by way of the heat exchanger into the feedwater pipe at the junction with the clean-up water pipe, where it is mixed with the feedwater passed by way of a feedwater heater and supplied to the pressure vessel. The feedwater temperature for the feedwater pipe and the set temperature for the clean-up water are compared with each other by using temperature sensors disposed to the feedwater pipe between the junction and the feedwater heater at the upstream of the junction. If the temperature difference is increased, for instance, upon transient state where the operation of the feedwater heater is not yet stabilized, the recycling pump is controlled to stop the supply of the clean-up water to the junction while flowing only the feedwater. This makes the temperature distribution uniform and prevents the generation of the thermal stresses at the junction, by which reactor safety can be improved. (Moriyama, K.)

  20. Nozzle

    Science.gov (United States)

    Chen, Alexander G.; Cohen, Jeffrey M.

    2009-06-16

    A fuel injector has a number of groups of nozzles. The groups are generally concentric with an injector axis. Each nozzle defines a gas flowpath having an outlet for discharging a fuel/air mixture jet. There are means for introducing the fuel to the air. One or more groups of the nozzles are oriented to direct the associated jets skew to the injector axis.

  1. A Study on the Influence of Fuel Pipe on Fuel Injection Characteristics of Each Nozzle Hole in Diesel Injector

    Directory of Open Access Journals (Sweden)

    Luo Fuqiang

    2016-01-01

    Full Text Available The inner diameter of high pressure fuel pipe has a significant effect on the fuel injection process and the performance of a diesel engine. The spray impact force of each nozzle hole of a conventional injection system of pump-line-nozzle for diesel engine (based on the spray momentum flux and the injection pressure (on a fuel injection pump test rig were measured. With varying fuel injection quantities and pump speed, the effects of the inner diameter of the high pressure fuel pipe on fuel injection process and the fuel injection characteristics of each nozzle hole were analyzed. It was noted from experimental results that the fuel injection pressure changes with variations in the inner diameter of the high pressure fuel pipe and also the injection duration gradually increases with increase in the inner diameter. At low injection pump speed, even with the same geometric fuel deliver rate, the injection duration also increases gradually. Due to throttling effect and reduction in injection pressure, the fuel injection quantities of the injection nozzle were relatively minimal when the inner diameters of the high pressure fuel pipe were respectively small and large. The optimum injection pipe inner diameter for the right quantity for fuel injection falls between the two cases (between small and large. In addition, the injection rate of each nozzle hole increases with the decrease in angle between the needle axis and each of the nozzle hole axis. The fuel injection quantity of each nozzle hole increases while their relative difference decreases with increasing pump speed.

  2. A Study on the Influence of Fuel Pipe on Fuel Injection Characteristics of Each Nozzle Hole in Diesel Injector

    OpenAIRE

    Luo Fuqiang; Wang Chuqiao; Xue Fuying; Ye Bingjian; Wu Xiwen

    2016-01-01

    The inner diameter of high pressure fuel pipe has a significant effect on the fuel injection process and the performance of a diesel engine. The spray impact force of each nozzle hole of a conventional injection system of pump-line-nozzle for diesel engine (based on the spray momentum flux) and the injection pressure (on a fuel injection pump test rig) were measured. With varying fuel injection quantities and pump speed, the effects of the inner diameter of the high pressure fuel pipe on fuel...

  3. Investigation of surface oxide morphology in SG feedwater pipes and study of its influence on flow accelerated corrosion rate

    International Nuclear Information System (INIS)

    Qiu, G.; Alos-Ramos, O.; Monchecourt, D.; Mansour, C.; Delaunay, S.; Trevin, S.

    2015-01-01

    Flow accelerated corrosion (FAC) affects carbon steel components in the secondary circuits of PWR plants. The mandatory use of the prediction tool BRT-CICERO in all its PWR plants enables EDF to perform efficient inspections programs and minimize the number of leaks in the secondary circuits. Due to the operating conditions, SG feedwater flow regulation (ARE) circuits can be affected by FAC phenomenon. Thickness loss has been reported by several plants during the last 10 years, although significant damage by FAC remains very rare. This paper describes the surface features observed on an ARE straight tube that has orange peel pattern with thickness loss on the one half of its inner surface and a thick fouling layer without much thickness loss on the other. An analysis of the oxide porosity and structure by SEM investigation has been carried out. The origin of fouling layer and its behavior in the ARE circuits environment (oxide solubility, flow stability/turbulence) have been discussed. Finally by comparing with the classic FAC models, an attempt of correlation between the presence of the fouling layer and the lower corrosion rate is proposed. (authors)

  4. Feedwater control system

    International Nuclear Information System (INIS)

    Cook, B.M.

    1982-01-01

    Excessive swing of the feedwater in nuclear reactor power supply apparatus on the occurrence of a transient is suppressed by injecting an anticipatory compensating signal (δWsub(fw)) into the control for the feedwater. Typical overshoot occurs on removal of a large part of the load, the steam flow is reduced so that the conventional control system reduces the flow of feedwater. At the same time there is a reduction of feedwater level in the steam generator because of the collapse of the bubbles under increased steam pressure. By the time the control responds to the drop in level, the apparatus has begun to stabilize so that there is overshoot. The anticipatory signal is derived from the boiling power (BP) which is a function of the nuclear power (Qsub(N)) developed, the enthalpy of saturated water (hsub(s)) and the enthalpy of the feedwater injected into the steam generator (hsub(fw)). From the boiling power (BP) and the increment in steam pressure resulting from the transient an anticipatory increment of feedwater flow is derived. This increment is added to the other parameters controlling the feedwater. (author)

  5. Simplified method to evaluate seismic nozzle loads on mechanical equipment connected to unbraced piping

    International Nuclear Information System (INIS)

    Detroux, P.; Lafaille, J.P.

    1991-01-01

    After ten years of operation, the Belgian Nuclear Power Plants had to be seismically reassessed; especially, new requirements were imposed to the oldest units. The method, presented in this paper, is based on the principle that all the piping connected to the equipment is replaced by a clamped-hinged beam with or without concentrated mass and of a characteristic length depending on the diameter, schedule, mass per length of the connected piping and on the floor response spectra applicable at the location of the equipment. A theoretical justification of the method is presented for the simplest cases. The case of added concentrated mass is investigated. Finally, several comparisons with a full modal spectral analysis are presented

  6. Jet flow issuing from an axisymmetric pipe-cavity-orifice nozzle

    Directory of Open Access Journals (Sweden)

    Broučková Zuzana

    2016-01-01

    Full Text Available An axisymmetric air jet flow is experimentally investigated under passive flow control. The jet issues from a pipe of the inner diameter and length of 10 mm and 150 mm which is equipped with an axisymmetric cavity at the pipe end. The cavity operates as a resonator creating self-sustained acoustic excitations of the jet flow. A mechanism of excitations is rather complex – in comparison with a common Helmholtz resonator. The experiments were performed using flow visualization, microphone measurements and time-mean velocity measurements by the Pitot probe. The power spectral density (PSD and the sound pressure level (SPL were evaluated from microphone measurements. The jet Reynolds number ranged Re = 1600–18 000. Distinguishable peaks in PSD indicated a function of the resonator. Because the most effective acoustic response was found at higher Re, a majority of experiments focused on higher Re regime. The results demonstrate effects of the passive control on the jet behavior. Fluid mixing and velocity decay along the axis is intensified. It causes shortening of the jet transition region. On the other hand, an inverse proportionality of the velocity decay (u ~ 1/x in the fully developed region is not changed. The momentum and kinetic energy fluxes decrease more intensively in the controlled jets in comparison with common jets.

  7. Feed water distribution pipe replacement at Loviisa NPP

    Energy Technology Data Exchange (ETDEWEB)

    Savolainen, S.; Elsing, B. [Imatran Voima Loviisa NPP (Finland)

    1995-12-31

    Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

  8. Reactor feedwater facility

    Energy Technology Data Exchange (ETDEWEB)

    Fujii, Tadashi; Kinoshita, Shoichiro; Akatsu, Jun-ichi

    1996-04-30

    In a reactor feedwater facility in which one stand-by system and at least three ordinary systems are disposed in parallel, each of the feedwater pumps is driven by an electromotor, and has substantially the same capacity. At least two systems among the ordinary systems have a pump rotation number variable means. Since the volume of each of the feedwater pump of each system is determined substantially equal, standardization is enabled to facilitate the production. While the number of electromotors is increased, since they are driven by electromotors, turbines, steam pipelines and valves for driving feed water pumps can be eliminated. Therefore, the feedwater pumps can be disposed to a region of low radiation dose being separated from a main turbine and a main condensator, to improve the degree of freedom in view of the installation. In addition, accessibility to equipments during operation is improved to improve the maintenance of feed water facilities. The number of parts for equipments can be reduced compared with that in a turbine-driving system thereby capable of reducing the operation amount for the maintenance and inspection. (N.H.)

  9. Feedwater temperature control device in nuclear power plants

    International Nuclear Information System (INIS)

    Nakamoto, Masashi.

    1985-01-01

    Purpose: To automatically and optimally control the temperature of feedwater supplied to the nuclear reactor of a nuclear power plant regardless the load on a steam turbine. Constitution: A pressure switch is disposed for detecting the turbine extract pressure within an extract pipeway leading to the feedwater heater and heating steams are supplied by selectively switching the control valve disposed to the pipeway for introducing the turbine extract or main steams to the feedwater heater by the signal from the pressure switch. Since the temperature at the exit of the feedwater heater is determined by the pressure inside of the respective equipments, the pressure inside the extract pipe is detected by the pressure switch, and the control valve is put to close and open based on the value to thereby control the entrance of steams to the feedwater heater. As a result, the temperature of the feedwater supplied to the nuclear reactor can be set and controlled automatically within a region where the steam is generated stably in the nuclear reactor. (Kamimura, M.)

  10. Reactor feedwater pump control device

    International Nuclear Information System (INIS)

    Nishiyama, Hiroyuki.

    1990-01-01

    An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

  11. Analysis of ultrasound propagation in high-temperature nuclear reactor feedwater to investigate a clamp-on ultrasonic pulse doppler flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige; Sakai, Yukihiro

    2008-01-01

    The flow rate of nuclear reactor feedwater is an important factor in the operation of a nuclear power reactor. Venturi nozzles are widely used to measure the flow rate. Other types of flowmeters have been proposed to improve measurement accuracy and permit the flow rate and reactor power to be increased. The ultrasonic pulse Doppler system is expected to be a candidate method because it can measure the flow profile across the pipe cross section, which changes with time. For accurate estimation of the flow velocity, the incidence angle of ultrasound entering the fluid should be estimated using Snell's law. However, evaluation of the ultrasound propagation is not straightforward, especially for a high-temperature pipe with a clamp-on ultrasonic Doppler flowmeter. The ultrasound beam path may differ from what is expected from Snell's law due to the temperature gradient in the wedge and variation in the acoustic impedance between interfaces. Recently, simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation, using 3D-FEM simulation code plus the Kirchhoff method, as it relates to flow profile measurement in nuclear reactor feedwater with the ultrasonic pulse Doppler system. (author)

  12. The development of new analysis procedures for reactor internals under pipe breaks

    International Nuclear Information System (INIS)

    Song, Heuy Gap; Jhung, Myung Jo; Chang, Sang Gyun; Lee, Gyu Man

    1993-04-01

    This study investigates the horizontal responses of the reactor internals due to a 14 inch safety injection nozzle break which is expected to cause the largest loads of the branch line pipe breaks defined for the YGN 3 and 4. It examines the effects of two forcing terms, RV motions and internals hydraulic loads, and suggests new procedure which can be used for the tributary pipe break analysis. The analysis result confirms the applicability of suggested procedure to a small size tributary pipe break analysis. Also, this study calculates the horizontal responses of the reactor internals due to a 3 inch pressurizer spray line nozzle break which is the only one remaining in the primary side after leak-before-break evaluation, and secondary side pipe breaks such as main steam line and economizer feedwater line. The responses are compared with those of safe shutdown earthquake(SSE) to show that SSE loads with a conservative margin may be used for the pipe break loads in the preliminary design. (Author)

  13. Piping Stress Analysis

    International Nuclear Information System (INIS)

    Setjo, Renaningsih

    2000-01-01

    Piping stress analysis on Primary Sampling System, Reactor Cooling System, and Feedwater System for AP600 have been performed. Piping stress analysis is one of the requirements in the design of piping system. Piping stress is occurred due to static and dynamic loads during service. Analysis was carried out. Using PS+CAEPIPE software based on the individual and combination loads with assumption that failure could be happened during normal, upset, emergency and faulted condition as describe in ASME III/ANSI B31.1. With performing the piping stress analysis, the layout (proper pipe routing) of the piping system can be design with the requirements of piping stress and pipe supports in mind I.e sufficient flexibility for thermal expansion, etc to commensurate with the i tended service such as temperatures, pressure, seismic and anticipated loading

  14. Feedwater control method and device therefor

    International Nuclear Information System (INIS)

    Nakahara, Mitsugu; Ichikawa, Yoshiaki; Ishii, Yoshikazu; Suzuki, Katsuyuki; Tanikawa, Naoshi; Mizuki, Fumio.

    1997-01-01

    The present invention provides a method of and a device for easily changing the constitution of feedwater systems without causing change in the water level of a reactor even when a plurality of feedwater systems have imbalance points. Namely, a feedwater control device comprises at least two feedwater systems capable of feeding water to tanks independently respectively and a controller capable of controlling water level in the tanks by controlling these feedwater systems. There is disposed a means for outputting gradually increasing driving signals to other feedwater systems, when the water level controller automatically controls one of the feedwater systems. There is also disposed a means for switching from automatic control for one of the feedwater systems to automatic control for the other feedwater system by a water level controller when the other feedwater system is in a stable operation region. As a result, entire feedwater flow rate is not temporarily changed and the water level in the tanks can be maintained constant. (N.H.)

  15. Water hammer calculation and analysis in main feedwater system of PWR nuclear power plants

    International Nuclear Information System (INIS)

    Wang Xin; Han Weishi

    2010-01-01

    The main feedwater system of a nuclear power plant is an important part in ensuring the cooling of the steam generator. Moreover, it is the main pipe section where water hammers frequently occur. Studying the regular patterns of water hammers to the main feedwater system is significant to the stable operation of the system. The paper focuses on the study of water hammers through Flowmaster's transient calculating function to establish a mathematical model with boundary conditions such as a feedwater pump, control valves, etc.; calculation of the water hammers pressure when feedwater pumps and control valves shut down; exporting the instantaneous change in solution of pressure. Combined with engineering practical examples, the conclusions verify the viability of calculating the water hammers pressure through Flowmaster's transient function, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively, changing the intervals of closing signals to feedwater pumps and control valves to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (authors)

  16. Feedwater heat condition assessment - An NDE perspective

    International Nuclear Information System (INIS)

    Krzywosz, K.J.

    1995-01-01

    An integral part of the feedwater heater condition assessment program consists of an inspection program performed on a periodic basis to help extend the availability of feedwater heaters. If applied effectively, NDE can extend the life of feedwater heaters by eliminating insurance plugs, minimizing tube leaks, assisting where necessary with design modifications and improved water chemistry, and repairing defective tubes for safety and remedial reasons. By performing inspection on a periodic basis, more realistic, up-to-date heater conditions can be obtained for a planned and timely heater replacements. This paper presents essential components of an effective NDE program to better prepare and assist utilities with reliable feedwater heater inspections

  17. Ultrasound propagation in steel piping at electric power plant using clamp-on ultrasonic pulse doppler velocity-profile flowmeter

    International Nuclear Information System (INIS)

    Tezuka, Kenichi; Mori, Michitsugu; Wada, Sanehiro; Aritomi, Masanori; Kikura, Hiroshige

    2008-01-01

    Venturi nozzles are widely used to measure the flow rates of reactor feedwater. This flow rate of nuclear reactor feedwater is an important factor in the operation of nuclear power reactors. Some other types of flowmeters have been proposed to improve measurement accuracy. The ultrasonic pulse Doppler velocity-profile flowmeter is expected to be a candidate method because it can measure the flow profiles across the pipe cross sections. For the accurate estimation of the flow velocity, the incidence angle of ultrasonic entering the fluid should be carefully estimated by the theoretical approach. However, the evaluation of the ultrasound propagation is not straightforward for the several reasons such as temperature gradient in the wedge or mode conversion at the interface between the wedge and pipe. In recent years, the simulation code for ultrasound propagation has come into use in the nuclear field for nondestructive testing. This article analyzes and discusses ultrasound propagation in steel piping and water, using the 3D-FEM simulation code and the Kirchhoff method, as it relates to the flow profile measurements in power plants with the ultrasonic pulse Doppler velocity-profile flowmeter. (author)

  18. Acoustic resonances in the steam and feedwater lines at the Dukovany NPP, unit 1

    International Nuclear Information System (INIS)

    Krupa, V.; Pecinka, L.

    1995-04-01

    The steam and feedwater line integrity and reliability programme for the loops of the Dukovany 1 reactor unit beyond the hermetic zone required calculation of the probability of their damage during normal operation and during a shock event such as a pressure wave propagation due to collapse of a steam or air bubble at the highest lying site of the feedwater piping (i. e. before the inlet to the main steam header) or to the closure of the quick-acting valve of the turbine, etc. For this purpose, the steam and feedwater eigenfrequencies were calculated for the segments from the steam generator to the turbine quick-acting valve or to the second high-pressure heater. The simple waveguide variants as well as the inclusion of the steam generator or main steam header as resonators were considered. The electromechanical analogy was employed for the calculation. (P.A.). 7 tabs., 8 figs., 6 refs

  19. Thermal-hydraulic analysis for changing feedwater check valve leakage rate testing methodology

    Energy Technology Data Exchange (ETDEWEB)

    Fuller, R.; Harrell, J.

    1996-12-01

    The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station are established from original licensing requirements that necessitate extremely restrictive air testing with tight allowable leakage limits. As a direct result of these requirements, the original high endurance hard seats in the FWCVs were modified with elastomeric seals to provide a sealing surface capable of meeting the stringent air leakage limits. However, due to the relatively short functional life of the elastomeric seals compared to the hard seats, the overall reliability of the sealing function actually decreased. This degraded performance was exhibited by frequent seal failures and subsequent valve repairs. The original requirements were based on limited analysis and the belief that all of the high energy feedwater vaporized during the LOCA blowdown. These phenomena would have resulted in completely voided feedwater lines and thus a steam environment within the feedwater leak pathway. To challenge these criteria, a comprehensive design basis accident analysis was developed using the RELAP5/MOD3.1 thermal-hydraulic code. Realistic assumptions were used to more accurately model the post-accident fluid conditions within the feedwater system. The results of this analysis demonstrated that no leak path exists through the feedwater lines during the reactor blowdown phase and that sufficient subcooled water remains in various portions of the feedwater piping to form liquid water loop seals that effectively isolate this leak path. These results provided the bases for changing the leak testing requirements of the FWCVs from air to water. The analysis results also established more accurate allowable leakage limits, determined the real effective margins associated with the FWCV safety functions, and led to design changes that improved the overall functional performance of the valves.

  20. 49 CFR 230.57 - Injectors and feedwater pumps.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 4 2010-10-01 2010-10-01 false Injectors and feedwater pumps. 230.57 Section 230... Appurtenances Injectors, Feedwater Pumps, and Flue Plugs § 230.57 Injectors and feedwater pumps. (a) Water.... Injectors and feedwater pumps must be kept in good condition, free from scale, and must be tested at the...

  1. Condition assessment of closed feedwater heaters

    International Nuclear Information System (INIS)

    Bell, R.J.

    1995-01-01

    Feedwater heaters are often forgotten in condition assessments plans. While they have no moving parts, these components have a significant impact on plant performance equivalent availability. Condition assessment of feedwater heaters includes not only an analysis of the tubing, which because of its thin wall nature is the primary objective of analysis, but other failure causes, such as tube joint leaks, an adverse condition which can and often does occur. For these reasons a comprehensive condition assessment program should be employed. This paper will identify the three level approach suggested by EPRI and many of the testing methods used to assess feedwater heater condition

  2. Auxiliary feedwater system aging study

    International Nuclear Information System (INIS)

    Kueck, J.D.

    1993-07-01

    This report documents the results of a Phase I follow-on study of the Auxiliary Feedwater (AFW) System that has been conducted for the US Regulatory Commission's Nuclear Plant Aging research Program. The Phase I study found a number of significant AFW System functions that are not being adequately tested by conventional test methods and some that are actually being degraded by conventional testing. Thus, it was decided that this follow-on study would focus on these testing omissions nd equipment degradation. The deficiencies in current monitoring and operating practice are categorized and evaluated. Areas of component degradation caused by current practice are discussed. Recommendations are made for improved diagnostic methods and test procedures

  3. Method for measuring feedwater flow rate using ultrasonic technique in PWR power plant

    International Nuclear Information System (INIS)

    Ozaki, Yoshihiko; Oda, Minoru; Tanaka, Mitsuo

    1988-01-01

    At present, differential pressure type flowmeters are widely used in feedwater systems of PWR plants. In these flowmeters, however, scales gradually deposit at the nozzle throat during the plant operation, causing the apparent flow rate to increase and consequently becoming a serious problem for efficient plant operations. Therefore, a new type of ultrasonic flowmeter (USFM) having good stability and free of the above phenomenon has been developed. A method to compensate for the effect of dependency of sound velocity on water temperature and pressure corresponding to PWR feedwater conditions was contrived. The validity of the method was confirmed in an experiment for investigating the sound velocity dependency in practice. The performance of the USFM was also examined using a water loop in various flow conditions with satisfactory results. After the basic studies, finally, the USFM was tested in an actual PWR feedwater system for almost 3 yr. The USFM met all the required characteristics for PWR feedwater systems, those being linearity, accuracy and stability. (author)

  4. Multi-unit shutdown due to boiler feedwater chemical excursion

    International Nuclear Information System (INIS)

    Diebel, M.E.

    1991-01-01

    Ontario Hydro's Bruce Nuclear Generating Station 'B' consists of four 935 W CANDU units located on the east shore of Lake Huron in the province of Ontario, Canada. On July 25 and 26, 1989 three of the four operating units were shutdown due to boiler feedwater chemical excursions initiated by a process upset in the Water Treatment Plant that provides demineralized make-up water to all four units. The chemicals that escaped from an ion exchange vessel during a routine regeneration very quickly spread through the condensate make-up system and into the boiler feedwater systems. This resulted in boiler sulfate levels exceeding shutdown limits. A total of 260 GWH of electrical generation was unexpectedly made unavailable to the grid at a time of peak seasonal demand. This event exposed several unforeseen deficiencies and vulnerabilities in the automatic demineralized water make-up quality protection scheme, system designs, operating procedures and the ability of operating personnel to recognize and appropriately respond to such an event. The combination of these factors contributed towards turning a minor system upset into a major multi-unit shutdown. This paper provides the details of the actual event initiation in the Water Treatment Plant and describes the sequence of events that led to the eventual shutdown of three units and near shutdown of the fourth. The design inadequacies, procedural deficiencies and operating personnel responses and difficulties are described. The process of recovering from this event, the flushing out of system piping, boilers and the feedwater train is covered as well as our experiences with setting up supplemental demineralized water supplies including trucking in water and the use of rental trailer mounted demineralizing systems. System design, procedural and operational changes that have been made and that are still being worked on in response to this event are described. The latest evidence of the effect of this event on boiler tube

  5. Feedwater heaters functional analysis at Embalse NGS

    International Nuclear Information System (INIS)

    Lolis, R.R.

    1992-01-01

    This study is concerned with the analysis or feedwater heaters, to detect actual failure or a bad trend beyond acceptable operating limits. When these situations are identified, preventive or corrective maintenance must be done. 2 tabs., 14 figs

  6. Aging and low-flow degradation of auxiliary feedwater pumps

    International Nuclear Information System (INIS)

    Adams, M.L.

    1991-01-01

    This paper documents the results of research done under the auspices of the Nuclear Regulatory Commission Nuclear Plant Aging Research Program. It examines the degradation imparted to safety Auxiliary Feedwater System pumps at nuclear plants due to the low flow operation. The Auxiliary Feedwater (AFW) System is normally a stand-by system. As such it is operated most often in the test mode. Since few plants are equipped with full flow test loops, most testing is accomplished at minimum flow conditions in pump by-pass lines. It is the vibration and hydraulic forces generated at low flow conditions that have been shown to be the major causes of AFW pump aging and degradation. The wear can be manifested in a number of ways, such as impeller or diffuser breakage, thrust bearing and/or balance device failure due to excessive loading, cavitation damage on such stage impellers, increase seal leakage or failure, sear injection piping failure, shaft or coupling breakage, and rotating element seizure

  7. Feedwater temperature control methods and systems

    Science.gov (United States)

    Moen, Stephan Craig; Noonan, Jack Patrick; Saha, Pradip

    2014-04-22

    A system for controlling the power level of a natural circulation boiling water nuclear reactor (NCBWR) is disclosed. The system, in accordance with an example embodiment of the present invention, may include a controller configured to control a power output level of the NCBWR by controlling a heating subsystem to adjust a temperature of feedwater flowing into an annulus of the NCBWR. The heating subsystem may include a steam diversion line configured to receive steam generated by a core of the NCBWR and a steam bypass valve configured to receive commands from the controller to control a flow of the steam in the steam diversion line, wherein the steam received by the steam diversion line has not passed through a turbine. Additional embodiments of the invention may include a feedwater bypass valve for controlling an amount of flow of the feedwater through a heater bypass line to the annulus.

  8. Interim status report on the revision of ASME PTC 12.1 -- closed feedwater heaters

    International Nuclear Information System (INIS)

    Stellern, J.L.; Hoobler, J.V.; Milton, J.W.; Welch, T.; Kona, C.; Thompson, H.N.; Tsou, J.L.

    1993-01-01

    The ASME Performance Test Code (PTC) 12.1-1978 for the performance testing of feedwater heaters is being revised extensively and updated. The committee anticipates that the final draft of the proposed Code will be ready for industry review in 1993. This Code revision will greatly enhance the usefulness and cost effectiveness of feedwater heater performance testing. This paper has been prepared to report on the progress of the committee and to disseminate information on the nature of the revision. Included in this paper are some of the notable changes intended for the Code. The most extensive change is the calculation method, which is described in step-by-step detail. An approach is also described for using ultrasonic flow techniques to test individual or split-string feedwater heaters, when flow nozzles are not available. Additionally some educational information on the use and limitations of ultrasonic measurement instrumentation is included. Discussion is also included on the required uncertainty analysis. 3 refs., 2 figs., 2 tabs

  9. A simplified approach to feedwater train modeling

    International Nuclear Information System (INIS)

    Ollat, X.; Smoak, R.A.

    1990-01-01

    This paper presents a method to simplify feedwater train models for power plants. A simple set of algebraic equations, based on mass and energy balances, is developed to replace complex representations of the components under certain assumptions. The method was tested and used to model the low pressure heaters of the Sequoyah Nuclear Plant in a larger simulation

  10. Nozzle seal

    International Nuclear Information System (INIS)

    Groff, R.D.; Vatovec, R.J.

    1978-01-01

    In an illustrative embodiment of the invention, a nuclear reactor pressure vessel, having an internal hoop from which the heated coolant emerges from the reactor core and passes through to the reactor outlet nozzles, is provided with annular sealing members operatively disposed between the outlet nozzle and the hoop and partly within a retaining annulus formed in the hoop. The sealing members are biased against the pressure vessel and the hoop and one of the sealing members is provided with a piston type pressure ring sealing member which effectively closes the path between the inlet and outlet coolants in the region about the outlet nozzle establishing a leak-proof condition. Furthermore, the flexible responsiveness of the seal assures that the seal will not structurally couple the hoop to the pressure vessel

  11. Scramjet Nozzles

    Science.gov (United States)

    2010-09-01

    integration et gestion thermique ) 14. ABSTRACT The lecture is given in four parts, each being a step in the process of nozzle design, and within each part...nose acts as a compressor at flight Mach numbers below 2.5, feeding a transfer duct which moves air rearwards below the 40m cabin to ramjet combustors...the fuselage, but with fuel tanks rather than a cabin above the transfer duct. The single nozzle along the wing trailing edge, highlighted in blue, was

  12. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  13. A Smart Soft Sensor Predicting Feedwater Flow Rate

    International Nuclear Information System (INIS)

    Yang, Heon Young; Na, Man Gyun

    2009-01-01

    Since we evaluate thermal nuclear reactor power with secondary system calorimetric calculations based on feedwater flow rate measurements, we need to measure the feedwater flow rate accurately. The Venturi flow meters that are being used to measure the feedwater flow rate in most pressurized water reactors (PWRs) measure the flow rate by developing a differential pressure across a physical flow restriction. The differential pressure is then multiplied by a calibration factor that depends on various flow conditions in order to calculate the feedwater flow rate. The calibration factor is determined by the feedwater temperature and pressure. However, Venturi meters cause a buildup of corrosion products near the orifice of the meter. This fouling increases the measured pressure drop across the meter, thereby causing an overestimation of the feedwater flow rate

  14. Condensate and feedwater systems, pumps, and water chemistry. Volume seven

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Subject matter includes condensate and feedwater systems (general features of condensate and feedwater systems, condenser hotwell level control, condensate flow, feedwater flow), pumps (principles of fluid flow, types of pumps, centrifugal pumps, positive displacement pumps, jet pumps, pump operating characteristics) and water chemistry (water chemistry fundamentals, corrosion, scaling, radiochemistry, water chemistry control processes, water pretreatment, PWR water chemistry, BWR water chemistry, condenser circulating water chemistry

  15. Improvements of feedwater controller for the super fast reactor

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Peng, Changhong; Ikejiri, Satoshi; Oka, Yoshiaki

    2010-01-01

    The main steam temperature of SCWRs sensitively changes with the power-to-flow ratio. In this article, the feedwater controller of the Super FR (fast-spectrum SCWR) is modified from that of the Super LWR (thermal-spectrum SCWR) for suppressing the variation of the main steam temperature. A plant system analysis code SPRAT-F is used. One of three feedback terms is added to the original feedwater controller that took only the deviation of the main steam temperature into consideration. In the feedwater controller (A), the deviation of the power-to-flow ratio is considered. In the feedwater controller (B), the deviation of the power is considered. In the feedwater controller (C), the time derivative of the power is considered. All the modified feedwater controllers keep the variation of the main steam temperature within 2degC, which has been achieved in recent supercritical coal-fired power plants, against typical load change. In order to further confirm the performance of the modified feedwater controllers, five typical perturbations are analyzed. All the feedwater controllers including the original one stably control the Super FR against all the perturbations without a significant oscillation or offset. Among them, the feedwater controller (B) gives a smaller or at least not larger variation of the main steam temperature compared with the original one at all the perturbations, while the feedwater controllers (A) and (C) give a larger variation in particular cases. From these results, it is concluded that the original feedwater controller is successfully improved as the feedwater controller (B). (author)

  16. Analysis of the effects of postulated pipe breaks on the LOFT containment building and on building TAN 650

    Energy Technology Data Exchange (ETDEWEB)

    Mosby, W.R.

    1978-08-31

    This report presents the results of analyses of the consequences of pipe whips and jets occurring as a result of pipe breaks postulated to occur in the LOFT main steam and feedwater lines both inside and adjacent to the LOFT Containment Vessel and Building TAN 650. Pipe whip and jet cases resulting in breach of containment or damage to Building TAN 650 are identified.

  17. Fuel nozzle assembly

    Science.gov (United States)

    Johnson, Thomas Edward [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC; Lacey, Benjamin Paul [Greer, SC; York, William David [Greer, SC; Stevenson, Christian Xavier [Inman, SC

    2011-08-30

    A fuel nozzle assembly is provided. The assembly includes an outer nozzle body having a first end and a second end and at least one inner nozzle tube having a first end and a second end. One of the nozzle body or nozzle tube includes a fuel plenum and a fuel passage extending therefrom, while the other of the nozzle body or nozzle tube includes a fuel injection hole slidably aligned with the fuel passage to form a fuel flow path therebetween at an interface between the body and the tube. The nozzle body and the nozzle tube are fixed against relative movement at the first ends of the nozzle body and nozzle tube, enabling the fuel flow path to close at the interface due to thermal growth after a flame enters the nozzle tube.

  18. System Study: Auxiliary Feedwater 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2014-12-31

    This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the AFW results.

  19. System Study: Auxiliary Feedwater 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the auxiliary feedwater (AFW) system at 69 U.S. commercial nuclear power plants. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing or decreasing trends were identified in the AFW results.

  20. Development of 2-loop feedwater control system

    International Nuclear Information System (INIS)

    Omori, Takashi; Watanabe, Takao; Hirose, Masao.

    1981-01-01

    A 2-loop feedwater control system has been developed for automatic transfer control of the reactor feed pumps (RFP's) in BWR plants. This system consists of a master level controller and sub-loop flow controllers for each of the RFP's. Control characteristics of the 2-loop control system were investigated using a dynamic analysis code for the condensate feedwater system. Although the RFP system has a hydraulic coupling effect, the flow control loops become stable by setting adequate controller gains in the sub-loop flow controllers. The control characteristics in the major loop were modified in their initial response to level setpoint change by using a lead/lag compensator. Moreover, reactor core cooling was protected sufficiently during the transient in a trip of a turbine driven RFP. From simulation results of the transfer controls from the motor driven RFP to turbine driven RFP, it was ascertained that the 2-loop control system has such advantages as shorter completion time and superior controllability against ON-OFF action of a RFP recirculation valve during transfer control. (author)

  1. Thermal stratification in steam generator feedwater line. Life assessment from real transients

    International Nuclear Information System (INIS)

    Morilhat, P.

    1989-01-01

    Thermal stratification in the vicinity of steam generator feedwater nozzle at low flow rates has led to the surveillance of two specially instrumented PWR units during a complete fuel cycle. This paper is an overview of the tests performed. The analysis of the operating conditions of these units shows their representativness with the characteristics of the other plants and identifies the main phases responsible for stratified states. A simplified model for computing axial stresses is carried out. Computed stresses are compared with the values given by extensometric gauges. The fatigue usage factor is calculated from transients actually occurred on these units and the share of damage due to each type of event is assessed

  2. Operating experiences and degradation detection for auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Casada, D.; Farmer, W.S.

    1992-01-01

    A study of Pressurized Water Reactor Auxiliary Feedwater (AFW) Systems has been conducted by Oak Ridge National Laboratory (ORNL) under the auspices of the Nuclear Regulatory Commission's Nuclear Plant Aging Research Program. The results of the study are documented in NUREG/CR-5404, Vol. 1, Auxiliary Feedwater System Aging Study. The study reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results

  3. Piping failure analysis for the Korean nuclear piping including the effect of in-service inspection

    Energy Technology Data Exchange (ETDEWEB)

    Choi, S.Y. [Korea Atomic Energy Research Inst.(KAERI), Daejeon (Korea); Choi, Y.H. [Korea Inst. of Nuclear Safety(KINS), Daejeon (Korea)

    2004-07-01

    The purposes of this paper are to perform piping failure analysis for the failed safety class piping in Korean nuclear power plants(NPPs) and evaluate the effect of an in-service inspection(ISI) on the piping failure probability. For data collection, a database for piping failure events was constructed with 135 data fields including population data, event data, and service history data. A total of 6 kinds of events with 25 failure cases up to June 30, 2003 were identified from Korean NPPs. The failed systems were main feedwater system, CVCS, primary sampling system, essential service water system, and CANDU purification system. Piping failure analyses such as evaluation of the impact on nuclear safety and piping integrity and the root cause analysis were performed and the piping failure frequencies for the failed piping were calculated by using population data. The result showed that although the integrity was not maintained in the failed piping, the safety of the plants was maintained for all the events. And the root causes of the events were analyzed as FAC, vibration, thermal fatigue, corrosion, and/or an improper weld joint. The piping failure frequencies ranged from 6.08E-5/Cr-Yr to 1.15E-3/Cr-Yr for the events. According to the ASME Code sec. XI requirements, the small bore piping less than the nominal diameter of 4 inch is exempt from ISI. There, however, were many piping failures reported in the small bore piping. The effect of ISI considering the pipe size on the piping failure probability was investigated by using the Win-PRAISE program based on probabilistic fracture mechanics. The results showed that there is no significant difference between the small and large bore piping from the viewpoint of the ISI effect on the piping failure probability. It means that ISI for a small bore piping is recommended as well as the large bore piping. (orig.)

  4. Cold spray nozzle design

    Science.gov (United States)

    Haynes, Jeffrey D [Stuart, FL; Sanders, Stuart A [Palm Beach Gardens, FL

    2009-06-09

    A nozzle for use in a cold spray technique is described. The nozzle has a passageway for spraying a powder material, the passageway having a converging section and a diverging section, and at least the diverging section being formed from polybenzimidazole. In one embodiment of the nozzle, the converging section is also formed from polybenzimidazole.

  5. Controlling method for impurity concentration in the feedwater system of a BWR type reactor

    International Nuclear Information System (INIS)

    Yamazaki, Kenji

    1987-01-01

    Purpose: To decrease ionic radioactive deposition and reduce the exposure dose upon inspection in the recycling system of a nuclear reactor. Method: Fe concentration is analyzed based on the crud concentration measured by a turbidimeter and the concentration is calculated. Further, Ni density is analyzed by an ion crossed meter and the density is calculated. From the calculated values, the density between Fe density and Ni density, i.e., Fe/Ni is calculated. If the calculated value for Fe/Ni is lower than a setting value (Fe/Ni = 2), an open signal is outputted for an iron injection valve disposed at the midway of a pipeway connected from a tank containing suspended iron oxide or iron hydroxide to a feedwater pipe. Then, the opening degree of the iron injection valve is adjusted based on the deviation value between the setting value (Fe/Ni = 2) and the calculated value for Fe/Ni. (Yoshino, Y.)

  6. Aging assessment of auxiliary feedwater pumps

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1987-01-01

    ORNL is conducting aging assessments of auxiliary feedwater pumps to provide recommendations for monitoring and assessing the severity of time-dependent degradation as well as to recommend maintenance and replacement practices. Cornerstones of these activities are the identification of failure modes and causes and ranking of causes. Failure modes and causes of interest are those due to aging and service wear. Design details, functional requirements, and operating experience data were used to identify failure modes and causes and to rank the latter. Based on this input, potentially useful inspection, surveillance, and condition monitoring methods that are currently available for use or in the developmental stage were examined and recommendations made. The methods selected are listed and discussed in terms of use and information to be obtained. Relationships between inspection, surveillance, and monitoring and maintenance practices entered prominently into maintenance recommendations. These recommendations, therefore, embrace predictive as well as corrective and preventative maintenance practices. The recommendations are described, inspection details are discussed, and periodic inspection and maintenance interval guidelines are given. Surveillance testing at low-flow conditions is also discussed. It is shown that this type of testing can lead to accelerated aging

  7. 46 CFR 52.01-115 - Feedwater supply (modifies PG-61).

    Science.gov (United States)

    2010-10-01

    ... BOILERS General Requirements § 52.01-115 Feedwater supply (modifies PG-61). Boiler feedwater supply must meet the requirements of PG-61 of section I of the ASME Boiler and Pressure Vessel Code (incorporated...

  8. Robotic cleaning of radwaste tank nozzles

    International Nuclear Information System (INIS)

    Boughman, G.; Jones, S.L.

    1992-01-01

    The Susquehanna radwaste processing system includes two reactor water cleanup phase separator tanks and one waste sludge phase separator tank. A system of educator nozzles and associated piping is used to provide mixing in the tanks. The mixture pumped through the nozzles is a dense resin-and-water slurry, and the nozzles tend to plug up during processing. The previous method for clearing the nozzles had been for a worker to enter the tanks and manually insert a hydrolaser into each nozzle, one at a time. The significant radiation exposure and concern for worker safety in the tank led the utility to investigate alternate means for completing this task. The typical tank configuration is shown in a figure. The initial approach investigated was to insert a manipulator arm in the tank. This arm would be installed by workers and then teleoperated from a remote control station. This approach was abandoned because of several considerations including educator location and orientation, excessive installation time, and cost. The next approach was to use a mobile platform that would operate on the tank floor. This approach was selected as being the most feasible solution. After a competitive selection process, REMOTEC was selected to provide the mobile platform. Their proposal was based on the commercial ANDROS Mark 5 platform

  9. Comparison and evaluation of flexible and stiff piping systems

    International Nuclear Information System (INIS)

    Hahn, W.; Tang, H.T.; Tang, Y.K.

    1983-01-01

    An experimental and numerical study was performed on a piping system, with various support configurations, to assess the difference in piping response for flexible and stiff piping systems. Questions have arisen concerning a basic design philosophy employed in present day piping designs. One basic question is, the reliability of a flexible piping system greater than that of a stiff piping system by virtue of the fact that a flexible system has fewer snubber supports. With fewer snubbers, the pipe is less susceptible to inadvertent thermal stresses introduced by snubber malfunction during normal operation. In addition to the technical issue, the matter of cost savings in flexible piping system design is a significant one. The costs associated with construction, in-service inspection and maintenance are all significantly reduced by reducing the number of snubber supports. The evaluation study, sponsored by the Electric Power Research Institute, was performed on a boiler feedwater line at Consolidated Edison's Indian Point Unit 1. In this study, the boiler feedwater line was tested and analyzed with two fundamentally different support systems. The first system was very flexible, employing rod and spring hangers, and represented the 'old' design philosophy. The pipe system was very flexible with this support system, due to the long pipe span lengths between supports and the fact that there was only one lateral support. This support did not provide much restraint since it was near an anchor. The second system employed strut and snubber supports and represented the 'modern' design philosophy. The pipe system was relatively stiff with this support system, primarily due to the increased number of supports, including lateral supports, thereby reducing the pipe span lengths between supports. The second support system was designed with removable supports to facilitate interchange of the supports with different support types (i.e., struts, mechanical snubbers and hydraulic

  10. Firefighter Nozzle Reaction

    DEFF Research Database (Denmark)

    Chin, Selena K.; Sunderland, Peter B.; Jomaas, Grunde

    2017-01-01

    to anchor forces, the hose becomes straight. The nozzle reaction is found to equal the jet momentum flow rate, and it does not change when an elbow connects the hose to the nozzle. A forward force must be exerted by a firefighter or another anchor that matches the forward force that the jet would exert...... on a perpendicular wall. Three reaction expressions are derived, allowing it to be determined in terms of hose diameter, jet diameter, flow rate, and static pressure upstream of the nozzle. The nozzle reaction predictions used by the fire service are 56% to 90% of those obtained here for typical firefighting hand......Nozzle reaction and hose tension are analyzed using conservation of fluid momentum and assuming steady, inviscid flow and a flexible hose in frictionless contact with the ground. An expression that is independent of the bend angle is derived for the hose tension. If this tension is exceeded owing...

  11. Operation of the main feedwater system turbopump following plant trip with total failure of the auxiliary feedwater system

    International Nuclear Information System (INIS)

    Lucas Alvaro, A.M. de; Rosa Martinez, B. de la; Alcaide, F.; Toledano Camara, C.

    1993-01-01

    The Auxiliary Feedwater System (AF) is a safeguard system which has been designed to supply feedwater to the steam generators, cool the primary system and remove decay heat from the reactor when the main feedwater pumps fail due to loss of power or any other reason. Thus, when plant trip occurs, the AF system pumps start up automatically, allowing removal of decay heat from the reactor. However, even though this system (2 motor-driven pumps and 1 turbopump) is highly reliable, injection of water to the steam generators must be ensured when it fails completely. To do this, if plant trip has not been caused by loss of off site power or failure of the Main Feedwater System (FW) turbopumps, one of these turbopumps can be used to achieve removal of decay heat. Since a large amount of steam is consumed by these turbopumps, an analysis has been performed to determine whether one of these pumps can be used and what actions are necessary to inject water into the steam generators. Results show that, for the case in question, a FW turbopump can be used to remove decay heat from the reactor. (author)

  12. Control of Surge in Centrifugal Compressor by Using a Nozzle Injection System: Universality in Optimal Position of Injection Nozzle

    Directory of Open Access Journals (Sweden)

    Toshiyuki Hirano

    2012-01-01

    Full Text Available The passive control method for surge and rotating stall in centrifugal compressors by using a nozzle injection system was proposed to extend the stable operating range to the low flow rate. A part of the flow at the scroll outlet of a compressor was recirculated to an injection nozzle installed on the inner wall of the suction pipe of the compressor through the bypass pipe and injected to the impeller inlet. Two types of compressors were tested at the rotational speeds of 50,000 rpm and 60,000 rpm with the parameter of the circumferential position of the injection nozzle. The present experimental results revealed that the optimum circumferential position, which most effectively reduced the flow rate for the surge inception, existed at the opposite side of the tongue of the scroll against the rotational axis and did not depend on the compressor system and the rotational speeds.

  13. SU-E-T-239: Monte Carlo Modelling of SMC Proton Nozzles Using TOPAS

    International Nuclear Information System (INIS)

    Chung, K; Kim, J; Shin, J; Han, Y; Ju, S; Hong, C; Kim, D; Kim, H; Shin, E; Ahn, S; Chung, S; Choi, D

    2014-01-01

    Purpose: To expedite and cross-check the commissioning of the proton therapy nozzles at Samsung Medical Center using TOPAS. Methods: We have two different types of nozzles at Samsung Medical Center (SMC), a multi-purpose nozzle and a pencil beam scanning dedicated nozzle. Both nozzles have been modelled in Monte Carlo simulation by using TOPAS based on the vendor-provided geometry. The multi-purpose nozzle is mainly composed of wobbling magnets, scatterers, ridge filters and multi-leaf collimators (MLC). Including patient specific apertures and compensators, all the parts of the nozzle have been implemented in TOPAS following the geometry information from the vendor.The dedicated scanning nozzle has a simpler structure than the multi-purpose nozzle with a vacuum pipe at the down stream of the nozzle.A simple water tank volume has been implemented to measure the dosimetric characteristics of proton beams from the nozzles. Results: We have simulated the two proton beam nozzles at SMC. Two different ridge filters have been tested for the spread-out Bragg peak (SOBP) generation of wobbling mode in the multi-purpose nozzle. The spot sizes and lateral penumbra in two nozzles have been simulated and analyzed using a double Gaussian model. Using parallel geometry, both the depth dose curve and dose profile have been measured simultaneously. Conclusion: The proton therapy nozzles at SMC have been successfully modelled in Monte Carlo simulation using TOPAS. We will perform a validation with measured base data and then use the MC simulation to interpolate/extrapolate the measured data. We believe it will expedite the commissioning process of the proton therapy nozzles at SMC

  14. Transition nozzle combustion system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won-Wook; McMahan, Kevin Weston; Maldonado, Jaime Javier

    2016-11-29

    The present application provides a combustion system for use with a cooling flow. The combustion system may include a head end, an aft end, a transition nozzle extending from the head end to the aft end, and an impingement sleeve surrounding the transition nozzle. The impingement sleeve may define a first cavity in communication with the head end for a first portion of the cooling flow and a second cavity in communication with the aft end for a second portion of the cooling flow. The transition nozzle may include a number of cooling holes thereon in communication with the second portion of the cooling flow.

  15. An effect of downcomer feedwater fraction on steam generator performance with an axial flow economizer

    International Nuclear Information System (INIS)

    Jung, Byung Ryul; Park, Hu Shin; Chung, Duk Muk; Baik, Se Jin

    2000-01-01

    The effects of feedwater flow fraction introduced into the downcomer region have been evaluated in terms of steam generator performance based on the same steam generator thermal output for the Korea Standard Nuclear Power Plant (KSNP) steam generator. The KSNP steam generator design has an integral axial flow economizer which is designed such that most of the feedwater is introduced through the economizer region and only a portion of feedwater through the downcomer region. The feedwater flow introduced into the downcomer region is not normally controlled during the power operation. However, the actual feedwater fraction into the downcomer region may differ from the design flow depending on the as-built system and component characteristics. Investigated in this paper were the downcomer feedwater flow effects on the steam pressure, circulation ratio, internal void fraction and velocity distribution in the tube bundle region at the steady state operation using SAFE and ATHOS3 codes. The results show that the steam pressure increases and the resultant total feedwater flow increases with reducing the downcomer feedwater flow fraction for the same steam generator thermal output. The slight off-design condition of downcomer feedwater flow fraction renders no significant effect on the steam generator performance such as circulation ratios, steam qualities, void fractions and internal velocity distributions. The evaluation shows that the slight off-design downcomer feedwater flow fraction deviation up to ± 5% is acceptable for the steam generator performance

  16. Wear characterization of abrasive waterjet nozzles and nozzle materials

    Science.gov (United States)

    Nanduri, Madhusarathi

    Parameters that influence nozzle wear in the abrasive water jet (AWJ) environment were identified and classified into nozzle geometric, AWJ system, and nozzle material categories. Regular and accelerated wear test procedures were developed to study nozzle wear under actual and simulated conditions, respectively. Long term tests, using garnet abrasive, were conducted to validate the accelerated test procedure. In addition to exit diameter growth, two new measures of wear, nozzle weight loss and nozzle bore profiles were shown to be invaluable in characterizing and explaining the phenomena of nozzle wear. By conducting nozzle wear tests, the effects of nozzle geometric, and AWJ system parameters on nozzle wear were systematically investigated. An empirical model was developed for nozzle weight loss rate. To understand the response of nozzle materials under varying AWJ system conditions, erosion tests were conducted on samples of typical nozzle materials. The effect of factors such as jet impingement angle, abrasive type, abrasive size, abrasive flow rate, water pressure, traverse speed, and target material was evaluated. Scanning electron microscopy was performed on eroded samples as well as worn nozzles to understand the wear mechanisms. The dominant wear mechanism observed was grain pullout. Erosion models were reviewed and along the lines of classical erosion theories a semi-empirical model, suitable for erosion of nozzle materials under AWJ impact, was developed. The erosion data correlated very well with the developed model. Finally, the cutting efficiency of AWJ nozzles was investigated in conjunction with nozzle wear. The cutting efficiency of a nozzle deteriorates as it wears. There is a direct correlation between nozzle wear and cutting efficiency. The operating conditions that produce the most efficient jets also cause the most wear in the nozzle.

  17. Application of neural networks to validation of feedwater flow rate in a nuclear power plant

    International Nuclear Information System (INIS)

    Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

    1993-01-01

    Feedwater flow rate measurement in nuclear power plants requires periodic calibration. This is due to the fact that the venturi surface condition of the feedwater flow rate sensor changes because of a chemical reaction between the surface coating material and the feedwater. Fouling of the venturi surface, due to this chemical reaction and the deposits of foreign materials, has been observed shortly after a clean venturi is put in operation. A fouled venturi causes an incorrect measurement of feedwater flow rate, which in turn results in an inaccurate calculation of the generated power. This paper presents two methods for verifying incipient and continuing fouling of the venturi of the feedwater flow rate sensors. Both methods are based on the use of a set of dissimilar process variables dynamically related to the feedwater flow rate variable. The first method uses a neural network to generate estimates of the feedwater flow rate readings. Agreement, within a given tolerance, of the feedwater flow rate instrument reading, and the corresponding neural network output establishes that the feedwater flow rate instrument is operating properly. The second method is similar to the first method except that the neural network predicts the core power which is calculated from measurements on the primary loop, rather than the feedwater flow rates. This core power is referred to the primary core power in this paper. A comparison of the power calculated from the feedwater flow measurements in the secondary loop, with the calculated and neural network predicted primary core power provides information from which it can be determined whether fouling is beginning to occur. The two methods were tested using data from the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant

  18. Piping Inelastic Fracture Mechanics Analysis.

    Science.gov (United States)

    1980-06-30

    120 160 200 2410 280 320 2 (DEGREES) Fig. 16-Comparison on limit moment predictions with experimental rcsults- AISI 304 piping property data. For...1975, cracks were discovered on many 4 in. diameter 304 s.s. pipes for recirculation loop valve bypass, and on 10 in. diameter 304 s.s. reactor core...LOCATIONd THERM4AL SLEEVE REPAIR WELD TYPE 310 STAINLESS TEL C FVICt AREA SPO PCE Fig. 3.1-Duane Arnold recirculation-inlet-nozzle safe end configuration

  19. Aging assessment of PWR [Pressurized Water Reactor] Auxiliary Feedwater Systems

    International Nuclear Information System (INIS)

    Casada, D.A.

    1988-01-01

    In support of the Nuclear Regulatory Commission's Nuclear Plant Aging Research (NPAR) Program, Oak Ridge National Laboratory is conducting a review of Pressurized Water Reactor Auxiliary Feedwater Systems. Two of the objectives of the NPAR Program are to identify failure modes and causes and identify methods to detect and track degradation. In Phase I of the Auxiliary Feedwater System study, a detailed review of system design and operating and surveillance practices at a reference plant is being conducted to determine failure modes and to provide an indication of the ability of current monitoring methods to detect system degradation. The extent to which current practices are contributing to aging and service wear related degradation is also being assessed. This paper provides a description of the study approach, examples of results, and some interim observations and conclusions. 1 fig., 1 tab

  20. Boiler feedwater treatment using reverse osmosis at Suncor OSG

    International Nuclear Information System (INIS)

    Brown, T.

    1997-01-01

    The installation of a new 1000 cu m/hr reverse osmosis water treatment system for boiler feedwater at a Suncor plant was discussed. The selection process began in 1993 when Suncor identified a need to increase its boiler feedwater capacity. The company reviewed many options available to increase the treated water capacity. These included: contracting the supply of treated water, adding additional capacity, replacing the entire plant, reverse osmosis, and demineralization. The eventual decision was to build a new 1000 cu m/hr reverse osmosis water treatment plant with the following key components: a Degremont Infilco Ultra Pulsator Clarifier and a Glegg Water Conditioning multimedia filter, Amberpack softeners and reverse osmosis arrays. The reverse osmosis plant was environmentally favourable over an equivalent demineralization plant. A technical comparison was provided between demineralization and reverse osmosis. The system has proven to be successful and economical in meeting the plants needs. 5 figs

  1. Smart Soft-Sensing for the Feedwater Flowrate at PWRs Using a GMDH Algorithm

    Science.gov (United States)

    Lim, Dong Hyuk; Lee, Sung Han; Na, Man Gyun

    2010-02-01

    The thermal reactor power in pressurized water reactors (PWRs) is typically assessed using secondary system calorimetric calculations based on accurate measurements of the feedwater flowrate. Therefore, precise measurements of the feedwater flowrate are essential. In most PWRs, Venturi meters are used to measure the feedwater flowrate. However, the fouling phenomena of the Venturi meter deteriorate the accuracy of the existing hardware sensors. Consequently, it is essential to resolve the inaccurate measurements of the feedwater flowrate. In this study, in order to estimate the feedwater flowrate online with high precision, a smart soft sensing model for monitoring the feedwater flowrate was developed using a group method of data handling (GMDH) algorithm combined with a sequential probability ratio test (SPRT). The uncertainty of the GMDH model was also analyzed. The proposed sensing and monitoring algorithm was verified using the acquired real plant data from Yonggwang Nuclear Power Plant Unit 3.

  2. Feedwater flow measurements: challenges, current solutions, and 'soft' developments

    International Nuclear Information System (INIS)

    Ruan, D.; Roverso, D.; Fantoni, P.F.; Sanabrias, J.I.; Carrasco, J.A.; Fernandez, L.

    2002-07-01

    This report presents an early progress of a feasibility study of a computational intelligence approach to the enhancement of the accuracy of feedwater flow measurements in the framework of an ongoing cooperation between Tecnatom s.a. in Madrid and the OECD Halden Reactor Project (HRP) in Halden. The aim of this research project is to contribute to the development and validation of a flow sensor in a nuclear power plant (NPP). The basic idea is to combine the use of applied computational intelligence approaches (noise analysis, neural networks, fuzzy systems, wavelets etc.) with existing traditional flow measurements, and in particular with cross correlation flow meter concepts. In this report, Section 2 outlines relevant aspects of thermal power calculations on electrical power plants. Section 3 reviews from the available literature possible approaches and solutions for feedwater flow measurement, including ultrasonic flow meters, cross-correlation flow meters, and 'Virtural' flow meters with artificial neural networks. Section 4 reports typical experimental measurements at the Tecnatom's facility. Section 5 presents an integration approach and preliminary experimental tests. Section 6 discusses the role of soft computing techniques in the context of feedwater flow measurements related nuclear fields, and Section 7 highlights the future research direction. (Author)

  3. Replacement of the condenser necked feedwater heaters for nuclear power plant

    International Nuclear Information System (INIS)

    Shimada, A.; Yamaguchi, H.; Kimura, M.; Ooshima, Y.; Nakashima, Y.; Ogawa, K.; Akiba, T.; Hoshi, T.

    2008-01-01

    In Fukushima Daiichi Nuclear Power Station Unit 2, the feedwater heaters, which were placed in the condenser neck, were replaced in 2006. Many feedwater heaters had been replaced in BWR plants, but, only a few condenser necked heaters were replaced. As the necked heaters were large-scale and placed in the condenser, various technical problems should be solved for replacing. This report shows some solutions applied in the condenser necked feedwater heater replacement. (author)

  4. 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)

  5. Considerations for surviving the loss of a main feedwater pump at full power

    International Nuclear Information System (INIS)

    Gaydos, K.A.; Calvo, R.; Conroy, P.W.; Klein, C.M.; Mellers, J.E.

    1990-01-01

    Today's economics dictate that nuclear power operational costs be contained by addressing frequently-occurring trips that might be minimized or avoided via specific upgrades. Much recent attention has focused on the significant percentage of plant trips related to feedwater flow regulation; however, another frequent feedwater-related trip stems from the loss of a single main feedwater pump while operating at high power levels, causing a plant trip on low steam generator water-level. This paper summarizes the results of several plant-specific studies that evaluate a unit's capabilities to consistently survive the loss of a main feedwater pump from full power, and outlines a methodology for analyzing this capability

  6. Fluid/structure interaction in piping systems

    International Nuclear Information System (INIS)

    Kellner, A.; Schoenfelder, C.

    1982-01-01

    The global movement of piping systems caused by pressure pulses as well as the associated loads on bends, nozzles and piping support structures are usually computed by using the pressures given by a hydrodynamic calculation as driving functions in a consecutive dynamic structure analysis without taking into account the secondary pressure pulses induced by the piping movement in the fluid. It is shown how including this feed-back of the structure dynamics on the fluid can lead to a drastic reduction of the computed loads

  7. On-line validation of feedwater flow rate in nuclear power plants using neural networks

    International Nuclear Information System (INIS)

    Khadem, M.; Ipakchi, A.; Alexandro, F.J.; Colley, R.W.

    1994-01-01

    On-line calibration of feedwater flow rate measurement in nuclear power plants provides a continuous realistic value of feedwater flow rate. It also reduces the manpower required for periodic calibration needed due to the fouling and defouling of the venturi meter surface condition. This paper presents a method for on-line validation of feedwater flow rate in nuclear power plants. The method is an improvement of the previously developed method which is based on the use of a set of process variables dynamically related to the feedwater flow rate. The online measurements of this set of variables are used as inputs to a neural network to obtain an estimate of the feedwater flow rate reading. The difference between the on-line feedwater flow rate reading, and the neural network estimate establishes whether there is a need to apply a correction factor to the feedwater flow rate measurement for calculation of the actual reactor power. The method was applied to the feedwater flow meters in the two feedwater flow loops of the TMI-1 nuclear power plant. The venturi meters used for flow measurements are susceptible to frequent fouling that degrades their measurement accuracy. The fouling effects can cause an inaccuracy of up to 3% relative error in feedwater flow rate reading. A neural network, whose inputs were the readings of a set of reference instruments, was designed to predict both feedwater flow rates simultaneously. A multi-layer feedforward neural network employing the backpropagation algorithm was used. A number of neural network training tests were performed to obtain an optimum filtering technique of the input/output data of the neural networks. The result of the selection of the filtering technique was confirmed by numerous Fast Fourier Transform (FFT) tests. Training and testing were done on data from TMI-1 nuclear power plant. The results show that the neural network can predict the correct flow rates with an absolute relative error of less than 2%

  8. Expert system for nuclear power plant feedwater system diagnosis

    International Nuclear Information System (INIS)

    Meguro, R.; Kinoshita, Y.; Sato, T.; Yokota, Y.; Yokota, M.

    1987-01-01

    The Expert System for Nuclear Power Plant Feedwater System Diagnosis has been developed to assist maintenance engineers in nuclear power plants. This system adopts the latest process computer TOSBAC G8050 and the expert system developing tool TDES2, and has a large scale knowledge base which consists of the expert knowledge and experience of engineers in many fields. The man-machine system, which has been developed exclusively for diagnosis, improves the man-machine interface and realizes the graphic displays of diagnostic process and path, stores diagnostic results and searches past reference

  9. Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Sang, Seok Yoon [Engineering and Technical Center, Korea Hydro, Daejeon (Korea, Republic of)

    2014-08-15

    Since failure in, damage to, and performance degradation of power generation components in operation under harsh environment of high pressure and high temperature may cause both economic and human loss at power plants, highly reliable operation and control of these components are necessary. Therefore, a systematic method of diagnosing the condition of these components in its early stages is required. There have been many researches related to the diagnosis of these components, but our group developed an approach using a regression model and diagnosis table, specializing in diagnosis relating to thermal efficiency degradation of power plant. However, there was a difficulty in applying the method using the regression model to power plants with different operating conditions because the model was sensitive to value. In case of the method that uses diagnosis table, it was difficult to find the level at which each performance degradation factor had an effect on the components. Therefore, fuzzy logic was introduced in order to diagnose performance degradation using both qualitative and quantitative results obtained from the components' operation data. The model makes performance degradation assessment using various performance degradation variables according to the input rule constructed based on fuzzy logic. The purpose of the model is to help the operator diagnose performance degradation of components of power plants. This paper makes an analysis of power plant feedwater heater by using fuzzy logic. Feedwater heater is one of the core components that regulate life-cycle of a power plant. Performance degradation has a direct effect on power generation efficiency. It is not easy to observe performance degradation of feedwater heater. However, on the other hand, troubles such as tube leakage may bring simultaneous damage to the tube bundle and therefore it is the object of concern in economic aspect. This study explains the process of diagnosing and verifying typical

  10. Loss of Normal Feedwater analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Parzer, I.; Prosek, A.; Hrvatin, S.

    2000-01-01

    The purpose of these analyses was to perform calculations of a Loss of Normal (Main) Feedwater transient for Krsko NPP. The results of calculations were used for the verification of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. To perform the thermal-hydraulic analyses, the RELAP5/MOD2 computer code and the NPP Krsko input card deck were used. In the presented paper two scenarios have been analyzed. Both of them started with a loss of normal feedwater event. Thus, a reduction or an interruption of the heat removal by the secondary system occurred. The first scenario assumed that auxiliary feedwater was available during the transient, while in the second scenario both normal and auxiliary feedwater were unavailable. The results showed that with auxiliary feedwater pumps unavailable additional operator actions would be needed to prevent overheating of the core. (author)

  11. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae [FNC Technology Co., Yongin (Korea, Republic of)

    2016-10-15

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air.

  12. Evaluation of Flood Level under Main Feedwater Line Break Accident using GOTHIC Computer Code and Analytical Calculation by ANSI 56.11

    International Nuclear Information System (INIS)

    Kim, Keon Yeop; Park, Jae Won; Jeon, Woo Jae

    2016-01-01

    The design basis internal flooding is caused by postulated pipe ruptures or component failures. The flooding can cause failure of safety-related equipment and affect the integrity of the structure. Though large diameter pipe rupture is significant in flooding analysis, split breaks should also be considered with consideration of a spectrum of pipe break size and power level. The pipe rupture analysis should be based on the most severe single active failure. For enveloping spectrum of pipe break condition, flood relief paths are necessary and passive flood protection without operating action, basically, shall be applied. In this study, the evaluation of flood level in case of Main Feedwater Line Break (MFLB) was performed by using GOTHIC computer program and hand calculation. The flooding analyses were performed by hand calculation and GOTHIC analysis for an assumed MFLB condition. The calculated flood levels were 0.823m and 0.691m for hand calculation and GOTHIC analysis, respectively. In comparison to the GOTHIC analysis, hand calculation showed conservative results. However, in actual flood protection design, margin for uncertainty shall be considered, in order to reflect the outflow reducing effect due to vortex and intake of air

  13. Nuclear reactor pressure vessel

    International Nuclear Information System (INIS)

    McDonald, B.N.

    1976-01-01

    In nuclear power reactor systems which have a reactor core inside a pressure vessel, the feedwater inlet pipe and steam discharge nozzle usually require separate pressure vessel penetrations. This requirement involves a great deal of expensive high quality special machining, welding and weld joint testing. The invention overcomes most of these problems by nestling the feedwater inlet pipe inside the steam discharge nozzle. At the same time the individual heat exchanger modules are supported from the pressure vessel at the same location as the nested feedwater inlet pipe and steam discharge nozzle combination, thus eliminating the need to accomodate troublesome differential thermal expansion problems through special structures within the pressure vessel

  14. Operational challenges to feedwater/steam generator water level control

    International Nuclear Information System (INIS)

    Thomas, V.M.; Whaley, S.D.; Federico, P.A.

    2012-01-01

    Feedwater control and turbine control have historically been at the top of the list of contributors to unplanned outages and forced curtailments in the nuclear industry, and they remain so according to recent industry data. Much has been done and is available by way of measures to improve this area and, in spite of much progress, opportunities remain to extend implementation. Toward this end, this paper aims to focus upon feedwater control and provide background on associated characteristics and attributes as a context for identifying the issues which are key challenges that lie at the root of this concern. Primary groupings of these issues will be discussed in order to better define their nature and to establish a basis for a presentation of the range of solutions which have been implemented and remain available to address them. The need for a systems engineering approach, and the role of I&C and field-mounted equipment to application of these solutions will be discussed. (author)

  15. Operational challenges to feedwater/steam generator water level control

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, V.M.; Whaley, S.D.; Federico, P.A. [Westinghouse Electric Company, Cranberry Township, Pennsylvania (United States)

    2012-07-01

    Feedwater control and turbine control have historically been at the top of the list of contributors to unplanned outages and forced curtailments in the nuclear industry, and they remain so according to recent industry data. Much has been done and is available by way of measures to improve this area and, in spite of much progress, opportunities remain to extend implementation. Toward this end, this paper aims to focus upon feedwater control and provide background on associated characteristics and attributes as a context for identifying the issues which are key challenges that lie at the root of this concern. Primary groupings of these issues will be discussed in order to better define their nature and to establish a basis for a presentation of the range of solutions which have been implemented and remain available to address them. The need for a systems engineering approach, and the role of I&C and field-mounted equipment to application of these solutions will be discussed. (author)

  16. Inferential smart sensing for feedwater flowrate in PWRs

    International Nuclear Information System (INIS)

    Na, M. G.; Hwang, I. J.; Lee, Y. J.

    2006-01-01

    The feedwater flowrate that is measured by Venturi flow meters in most pressurized water reactors can be over-measured because of the fouling phenomena that make corrosion products accumulate in the Venturi meters. Therefore, in this work, two kinds of methods, a support vector regression method and a fuzzy modeling method, combined with a sequential probability ratio test, are used in order to accurately estimate online the feedwater flowrate, and also to monitor the status of the existing hardware sensors. Also, the data for training the support vector machines and the fuzzy model are selected by using a subtractive clustering scheme to use informative data from among all acquired data. The proposed inferential sensing and monitoring algorithm is verified by using the acquired real plant data of Yonggwang Nuclear Power Plant Unit 3. In the simulations, it was known that the root mean squared error and the relative maximum error are so small and the proposed method early detects the degradation of an existing hardware sensor. (authors)

  17. Inferential smart sensing for feedwater flowrate in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Na, M. G.; Hwang, I. J. [Nuclear Eng. Dept., Chosun Univ., 375 Seosuk-dong, Dong-gu, Gwangju 501-759 (Korea, Republic of); Lee, Y. J. [Nuclear and Energy Eng. Dept., Cheju National Univ., 1 Ara-il-dong, Jeju-do 690-756 (Korea, Republic of)

    2006-07-01

    The feedwater flowrate that is measured by Venturi flow meters in most pressurized water reactors can be over-measured because of the fouling phenomena that make corrosion products accumulate in the Venturi meters. Therefore, in this work, two kinds of methods, a support vector regression method and a fuzzy modeling method, combined with a sequential probability ratio test, are used in order to accurately estimate online the feedwater flowrate, and also to monitor the status of the existing hardware sensors. Also, the data for training the support vector machines and the fuzzy model are selected by using a subtractive clustering scheme to use informative data from among all acquired data. The proposed inferential sensing and monitoring algorithm is verified by using the acquired real plant data of Yonggwang Nuclear Power Plant Unit 3. In the simulations, it was known that the root mean squared error and the relative maximum error are so small and the proposed method early detects the degradation of an existing hardware sensor. (authors)

  18. Simulation of a passive auxiliary feedwater system with TRACE5

    Energy Technology Data Exchange (ETDEWEB)

    Lorduy, María; Gallardo, Sergio; Verdú, Gumersindo, E-mail: maloral@upv.es, E-mail: sergalbe@iqn.upv.es, E-mail: gverdu@iqn.upv.es [Instituto Universitario de Seguridad Industrial, Radiofísica y Medioambiental (ISIRYM), València (Spain)

    2017-07-01

    The study of the nuclear power plant accidents occurred in recent decades, as well as the probabilistic risk assessment carried out for this type of facility, present human error as one of the main contingency factors. For this reason, the design and development of generation III, III+ and IV reactors, which include inherent and passive safety systems, have been promoted. In this work, a TRACE5 model of ATLAS (Advanced Thermal- Hydraulic Test Loop for Accident Simulation) is used to reproduce an accidental scenario consisting in a prolonged Station BlackOut (SBO). In particular, the A1.2 test of the OECD-ATLAS project is analyzed, whose purpose is to study the primary system cooling by means of the water supply to one of the steam generators from a Passive Auxiliary Feedwater System (PAFS). This safety feature prevents the loss of secondary system inventory by means of the steam condensation and its recirculation. Thus, the conservation of a heat sink allows the natural circulation flow rate until restoring stable conditions. For the reproduction of the test, an ATLAS model has been adapted to the experiment conditions, and a PAFS has been incorporated. >From the simulation test results, the main thermal-hydraulic variables (pressure, flow rates, collapsed water level and temperature) are analyzed in the different circuits, contrasting them with experimental data series. As a conclusion, the work shows the TRACE5 code capability to correctly simulate the behavior of a passive feedwater system. (author)

  19. Remotely replaceable fuel and feed nozzles for the NWCF calciner vessel

    International Nuclear Information System (INIS)

    Fletcher, R.D.; Carter, J.A.; May, K.W.

    1978-01-01

    The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility (NWCF) being built at the Idaho National Engineering Laboratory are described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

  20. Welding residual stress distributions for dissimilar metal nozzle butt welds in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Soo; Kim, Ju Hee; Bae, Hong Yeol; OH, Chang Young; Kim, Yun Jae [Korea Univ., Seoul (Korea, Republic of); Lee, Kyungsoo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of); Song, Tae Kwang [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2012-02-15

    In pressurized water nuclear reactors, dissimilar metal welds are susceptible to primary water stress corrosion cracking. To access this problem, accurate estimation of welding residual stresses is important. This paper provides general welding residual stress profiles in dissimilar metal nozzle butt welds using finite element analysis. By introducing a simplified shape for dissimilar metal nozzle butt welds, changes in the welding residual stress distribution can be seen using a geometry variable. Based on the results, a welding residual stress profile for dissimilar metal nozzle butt welds is proposed that modifies the existing welding residual stress profile for austenitic pipe butt welds.

  1. Development of Weld Overlay Technology for Dissimilar Welds in Pressurizer Nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Park, K. S.; Byeon, J. G.; Lee, J. B. [Doosan Heavy Industries and Construction Co., Daejeon (Korea, Republic of)

    2009-10-15

    As a result of Primary Water Stress Corrosion Cracking (PWSCC) in alloy 600, leaks in dissimilar metal welds of pressurizer nozzles were discovered recently in several US plants. The involved companies developed advanced repair techniques to prevent or repair PWSCC applying weld overlay procedures to dissimilar metal welds such as those between pipes and nozzles. Within 2 or 3 years, more than half of the nuclear power plants in Korea will have been in operation for more than 20 years. From this background, a weld overlay procedure has been developed in Korea for the dissimilar metal welds of pressurizer nozzles.

  2. Remotely replaceable fuel and feed nozzles for the new waste calcining facility calciner vessel

    International Nuclear Information System (INIS)

    Fletcher, R.D.; Carter, J.A.; May, K.W.

    1978-01-01

    The development and testing of remotely replaceable fuel and feed nozzles for calcination of liquid radioactive wastes in the calciner vessel of the New Waste Calcining Facility being built at the Idaho National Engineering Laboratory is described. A complete fuel nozzle assembly was fabricated and tested at the Remote Maintenance Development Facility to evolve design refinements, identify required support equipment, and develop handling techniques. The design also provided for remote replacement of the nozzle support carriage and adjacent feed and fuel pipe loops using two pairs of master-slave manipulators

  3. Premixed direct injection nozzle

    Science.gov (United States)

    Zuo, Baifang [Simpsonville, SC; Johnson, Thomas Edward [Greer, SC; Lacy, Benjamin Paul [Greer, SC; Ziminsky, Willy Steve [Simpsonville, SC

    2011-02-15

    An injection nozzle having a main body portion with an outer peripheral wall is disclosed. The nozzle includes a plurality of fuel/air mixing tubes disposed within the main body portion and a fuel flow passage fluidly connected to the plurality of fuel/air mixing tubes. Fuel and air are partially premixed inside the plurality of the tubes. A second body portion, having an outer peripheral wall extending between a first end and an opposite second end, is connected to the main body portion. The partially premixed fuel and air mixture from the first body portion gets further mixed inside the second body portion. The second body portion converges from the first end toward said second end. The second body portion also includes cooling passages that extend along all the walls around the second body to provide thermal damage resistance for occasional flame flash back into the second body.

  4. Study of check valve slamming in a BWR feedwater system following a postulated pipe break

    International Nuclear Information System (INIS)

    Safwat, H.H.; Arastu, A.H.; Norman, A.

    1985-01-01

    This study deals with a swing check valve slamming due to a break at relatively short distance from the valve. Under this situation, substantial flashing occurs near the valve and the result of the study are subject to what is believed to be a conservative simplifying assumption, i.e., the hydrodynamic moment acting on the valve during the transient is represented by resultant moment due to the pressure differential across the valve. It is believed that vapor voids forming at the valve would actually reduce the disk impact velocities in comparison to those predicted under this simplifying assumption. A technique used to represent a double-ended break through hypothetical valves may have some influence on the results particularly for long break opening times. The study has yielded good insight to help understand the complex problem. The study has focused on some parameters and the reader may raise questions on the effects of other parameters. Nevertheless, the present study underlines the complexity facing analysts dealing with this transient using analytical methods. Though some experimental data are available, the authors believe that an experimental study (recognizing the complexity of the experimental setup and instrumentation), would be quite useful. It can provide answers to questions facing analysts dealing with this problem and thus avoid unnecessary conservatisms due to uncertainties in input data

  5. Limit loads in nozzles

    International Nuclear Information System (INIS)

    Zouain, N.

    1983-01-01

    The static method for the evaluation of the limit loads of a perfectly elasto-plastic structure is presented. Using the static theorem of Limit Analysis and the Finite Element Method, a lower bound for the colapso load can be obtained through a linear programming problem. This formulation if then applied to symmetrically loaded shells of revolution and some numerical results of limit loads in nozzles are also presented. (Author) [pt

  6. Heat exchanger inventory cost optimization for power cycles with one feedwater heater

    International Nuclear Information System (INIS)

    Qureshi, Bilal Ahmed; Antar, Mohamed A.; Zubair, Syed M.

    2014-01-01

    Highlights: • Cost optimization of heat exchanger inventory in power cycles is investigated. • Analysis for an endoreversible power cycle with an open feedwater heater is shown. • Different constraints on the power cycle are investigated. • The constant heat addition scenario resulted in the lowest value of the cost function. - Abstract: Cost optimization of heat exchanger inventory in power cycles with one open feedwater heater is undertaken. In this regard, thermoeconomic analysis for an endoreversible power cycle with an open feedwater heater is shown. The scenarios of constant heat rejection and addition rates, power as well as rate of heat transfer in the open feedwater heater are studied. All cost functions displayed minima with respect to the high-side absolute temperature ratio (θ 1 ). In this case, the effect of the Carnot temperature ratio (Φ 1 ), absolute temperature ratio (ξ) and the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ) are qualitatively the same. Furthermore, the constant heat addition scenario resulted in the lowest value of the cost function. For variation of all cost functions, the smaller the value of the phase-change absolute temperature ratio for the feedwater heater (Φ 2 ), lower the cost at the minima. As feedwater heater to hot end unit cost ratio decreases, the minimum total conductance required increases

  7. Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

    International Nuclear Information System (INIS)

    Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

    1987-01-01

    From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 MPa were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. (orig./GL)

  8. Control of feedwater composition of BWR power plant

    International Nuclear Information System (INIS)

    Sturla, P.; D'Anna, A.; Borgese, D.

    1983-01-01

    Corrosion behaviour of fuel element cladding, cycle structural materials and dose rate increase are relevant to physico-chemical characteristics of process coolants and to adopted operational conditions. A careful control of cycle chemistry, during loading and shutdown periods, is necessary to verify material choices, the polishing system and chemistry specifications. For this purpose ENEL carried out some preliminary experimental tests employing continuous control system and samples for specific analytical determinations. The cycle points checked during about two months were: main condensate; condensate after polishing system; outlet of low pressure heathers; final feedwater; inlet and outlet of clean-up system; drains to condenser. The physico-chemical analysis were related to corrosion product levels (Cu, Fe, Ni, Co) and water chemistry (pH, conductivity, dissolved oxygen etc.). The preliminary results allow to express some considerations about sampling procedures, detection limits and reliability of analytical employed methods. The acquisition data time and some morphological oxide pictures are also showed. (author)

  9. 'Better feedwater quality through heat exchange equipment renovation'

    International Nuclear Information System (INIS)

    Pouzenc, C.

    2002-01-01

    In a fossil-fired or nuclear steam power plant, the water secondary circuit is a critical part of its thermodynamic cycle, as it achieves conditioning, pressurizing and heating of the condensate to match the conditions required at the steam generator inlet. Furthermore, the power plant electrical output and efficiency depend on availability and performances of each component of this secondary circuit from the condenser to the steam generator. Erosion and corrosion phenomena are at the origin of most significant failures in these components and related interconnecting systems. Feedwater chemistry is, together with the selection of materials and optimization of fluid velocities, one of the key levers to protect, as efficiently as possible, the components of the water secondary. (authors)

  10. Leak Detection of High Pressure Feedwater Heater Using Empirical Models

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Song Kyu; Kim, Eun Kee [Korea Power Engineering Company, Daejeon (Korea, Republic of); Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); An, Sang Ha [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-10-15

    Even small leak from tube side or pass partition within the high pressure feedwater heater (HPFWH) causes a significant deficiency in its performance. Plant operation under the HPFWH leak condition for long time will result in cost increase. Tube side leak within HPFWH can produce the high velocity jet of water and it can cause neighboring tube failures. However, most of plants are being operated without any information for internal leaks of HPFWH, even though it is prone to be damaged under high temperature and high pressure operating conditions. Leaks from tubes and/or pass partition of HPFWH occurred in many nuclear power plants, for example, Mihama PS-2, Takahama PS-2 and Point Beach Nuclear Plant Unit 1. If the internal leaks of HPFWH are monitored, the cost can be reduced by inexpensive repairs relative to loss in performance and moreover plant shutdown as well as further tube damages can be prevented.

  11. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1994-01-01

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

  12. Pipe grabber

    Energy Technology Data Exchange (ETDEWEB)

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

    1981-05-15

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

  13. Fuel nozzle tube retention

    Energy Technology Data Exchange (ETDEWEB)

    Cihlar, David William; Melton, Patrick Benedict

    2017-02-28

    A system for retaining a fuel nozzle premix tube includes a retention plate and a premix tube which extends downstream from an outlet of a premix passage defined along an aft side of a fuel plenum body. The premix tube includes an inlet end and a spring support feature which is disposed proximate to the inlet end. The premix tube extends through the retention plate. The spring retention feature is disposed between an aft side of the fuel plenum and the retention plate. The system further includes a spring which extends between the spring retention feature and the retention plate.

  14. 46 CFR 52.25-3 - Feedwater heaters (modifies PFH-1).

    Science.gov (United States)

    2010-10-01

    ... BOILERS Other Boiler Types § 52.25-3 Feedwater heaters (modifies PFH-1). In addition to the requirements in PFH-1 of section I of the ASME Boiler and Pressure Vessel Code (incorporated by reference; see 46...

  15. Study of thermohydraulic characteristics of upgraded feedwater collector in PGV-440 steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Tarankov, G.A.; Trunov, N.B.; Titov, V.F. [OKB Gidropress (Russian Federation); Urbansky, V.V. [Rovno NPP (Ukraine); Lenkei, I.; Notarosh, M. [Paks NPP (Hungary)

    1995-12-31

    Reconstruction of feedwater distribution collector was performed at unit 1 of Rowno NPP. Main results of measurements of temperatures in water volume, reparation characteristics and impurities distribution are presented. Analysis of tests results and design criteria is given. (orig.).

  16. A decision theoretic approach to an accident sequence: when feedwater and auxiliary feedwater fail in a nuclear power plant

    International Nuclear Information System (INIS)

    Svenson, Ola

    1998-01-01

    This study applies a decision theoretic perspective on a severe accident management sequence in a processing industry. The sequence contains loss of feedwater and auxiliary feedwater in a boiling water nuclear reactor (BWR), which necessitates manual depressurization of the reactor pressure vessel to enable low pressure cooling of the core. The sequence is fast and is a major contributor to core damage in probabilistic risk analyses (PRAs) of this kind of plant. The management of the sequence also includes important, difficult and fast human decision making. The decision theoretic perspective, which is applied to a Swedish ABB-type reactor, stresses the roles played by uncertainties about plant state, consequences of different actions and goals during the management of a severe accident sequence. Based on a theoretical analysis and empirical simulator data the human error probabilities in the PRA for the plant are considered to be too small. Recommendations for how to improve safety are given and they include full automation of the sequence, improved operator training, and/or actions to assist the operators' decision making through reduction of uncertainties, for example, concerning water/steam level for sufficient cooling, time remaining before insufficient cooling level in the tank is reached and organizational cost-benefit evaluations of the events following a false alarm depressurization as well as the events following a successful depressurization at different points in time. Finally, it is pointed out that the approach exemplified in this study is applicable to any accident scenario which includes difficult human decision making with conflicting goals, uncertain information and with very serious consequences

  17. Effect of Turbine Axial Nozzle-Wheel Clearance on Performance of Mark 25 Torpedo Power Plant

    Science.gov (United States)

    Hoyt, Jack W.; Kottas, Harry

    1948-01-01

    Investigations were made of the turbine from a Mark 25 torpedo to determine the performance of the unit with three different turbine nozzles at various axial nozzle-wheel clearances. Turbine efficiency with a reamed nondivergent nozzle that uses the axial clearance space for gas expansion was little affected by increasing the axial running clearance from 0.030 to 0.150 inch. Turbine efficiency with cast nozzles that expanded the gas inside the nozzle passage was found to be sensitive to increased axial nozzle-wheel clearance. A cast nozzle giving a turbine brake efficiency of 0.525 at an axial running clearance of 0.035 inch gave a brake efficiency of 0.475 when the clearance was increased to 0.095 inch for the same inlet-gas conditions and blade-jet speed ratio. If the basis for computing the isentropic power available to the turbine is the temperature inside the nozzle rather then the temperature in the inlet-gas pipe, an increase in turbine efficiency of about 0.01 is indicated.

  18. Nuclear plant power up-rate study: feedwater heater evaluations

    International Nuclear Information System (INIS)

    Svensson, Eric; Catapano, Michael; Coakley, Michael; Thomas, Dan

    2014-01-01

    Given today's nuclear industry business climate, it has become common for Utility companies to consider increasing unit capacities through turbine replacement and power up-rates. An integral part of the studies conducted by many towards this end, involve the generation of a set of turbine cycle heat balances with predicted performance parameters for the up-rated condition. Once these tentative operating values are established, it becomes necessary to evaluate the suitability of the existing components within each system to ensure they are capable of continued safe and reliable operation. The ultimate cost for the up-rate, including the cost for any major required modifications or significant replacements is weighed against increased revenue generated from the up-rate over time. Exelon's Peach Bottom Atomic Power Station (PBAPS) is currently planning for an Extended Power up-rate (EPU) for both units. To ensure the existing Feedwater Heaters (FWH) could maintain the operating and transient response margins at the EPU condition, an engineering study was conducted. Powerfect Inc. in conjunction with SPX Heat Transfer LLC were contracted to provide engineering services to analyze the design, thermal performance, reliability and operating conditions at projected EPU conditions. Specifically, to address the following with regard to the station's Feedwater Heaters (FWHs): 1. Evaluate Drain Cooler (DC) Velocities - including zone inlet velocity, cross and window velocities and outlet velocities. 2. Evaluate Drain Cooler Zone Pressure Drop for effect on drain cooler margins to flashing. 3. Evaluate differential pressure allowable across the pass partition plate. 4. Evaluate Drain Cooler Tube Vibration Potential. 5. Perform detailed steam dome velocity calculations. The goal of the study was to identify the most susceptible areas within the heaters for problems and potential failures when operating at the higher duty of the EPU condition for the remaining life

  19. Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhao Xiaoyu

    1996-01-01

    The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

  20. EDF (Electricite de France) feedback shot-peening on feedwater plants working to 360 0 C: prediction correlation and follow-up of thermal stresses relaxation

    International Nuclear Information System (INIS)

    Gauchet, J.P.

    1995-01-01

    This study predicts life duration of shot-peening effect and finally to allow the plant operator to prepare routine stopping, considering the following four steps have been: the shot-peening parameters must been carefully chosen and implementation must be reliable and perfectly reproducible; the residual stresses and cold working state checked by X-ray diffraction; the EDF feedback on different steam-water system components working at around-300 0 C and repaired by shot-peening, like feed heater water boxes, water tanks and vessels, steam pipes; a program, carried out on a feedwater tank repaired by welding and hot-peening and working at 360 0 C, on the correlation between expected and effective results. (author). 7 refs., 3 figs., 1 tab

  1. Thermal fatigue damage evaluation of a PWR NPP steam generator injection nozzle model subjected to thermal stratification phenomenon

    International Nuclear Information System (INIS)

    Leite da Silva, Luiz; Rodrigues Mansur, Tanius; Cimini Junior, Carlos Alberto

    2011-01-01

    Thermal stratification phenomenon with the same thermodynamic steam generator (SG) injection nozzle parameters was simulated. After 41 experiments, the experimental section was dismantled; cut and specimens were made of its material. Other specimens were made of the preserved pipe material. By comparing their fatigue tests results, the pipe material damage was evaluated. The water temperature layers and also the outside pipe wall temperatures were measured at the same level. Strains outside the pipe in 7 positions were measured. The experimental section develops thermal stratified flows, stresses and strains caused enlargement of material grain size and reduction in fatigue life.

  2. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

    Lu, S.C.; Tsai, N.C.

    1983-08-01

    The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

  3. Economizer water-wall damages initiated by feedwater impurities

    Directory of Open Access Journals (Sweden)

    Vidojković Sonja M.

    2014-01-01

    Full Text Available The main causes of efficiency loss in thermal power plants are boiler tube failures that diminish unit reliability and availability, and raise the cost of the electric energy. For that reason, regular examination of boiler tubes is indispensable measure for prevention future malfunctions of power units. Microscopic examination of economizer inner wall microstructure, analysis of chemical composition of deposit using x-ray diffraction (XRD and scanning electron microscopy/energy dispersive spectroscopy (SEM/EDS has been performed in a subcritical power plant. Stress corrosion cracking, pitting corrosion, destroyed protective magnetite layer, presence of magnetite and hematite in deposit and corrosive impurities within the cracks were indicated the effect of inadequate quality of feedwater that can not entirely ensure reliable operation of the boiler. It may be stated that maintenance of present boiler does not provide its reliable operation. Extensive chemical control of water/steam cycle was recommended. [Projekat Ministartsva nauke Republike Srbije, br. III 43009 i br. III 45012

  4. 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

  5. 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

  6. Low NOx nozzle tip for a pulverized solid fuel furnace

    Science.gov (United States)

    Donais, Richard E; Hellewell, Todd D; Lewis, Robert D; Richards, Galen H; Towle, David P

    2014-04-22

    A nozzle tip [100] for a pulverized solid fuel pipe nozzle [200] of a pulverized solid fuel-fired furnace includes: a primary air shroud [120] having an inlet [102] and an outlet [104], wherein the inlet [102] receives a fuel flow [230]; and a flow splitter [180] disposed within the primary air shroud [120], wherein the flow splitter disperses particles in the fuel flow [230] to the outlet [104] to provide a fuel flow jet which reduces NOx in the pulverized solid fuel-fired furnace. In alternative embodiments, the flow splitter [180] may be wedge shaped and extend partially or entirely across the outlet [104]. In another alternative embodiment, flow splitter [180] may be moved forward toward the inlet [102] to create a recessed design.

  7. Injection nozzle for a turbomachine

    Science.gov (United States)

    Uhm, Jong Ho; Johnson, Thomas Edward; Kim, Kwanwoo

    2012-09-11

    A turbomachine includes a compressor, a combustor operatively connected to the compressor, an end cover mounted to the combustor, and an injection nozzle assembly operatively connected to the combustor. The injection nozzle assembly includes a first end portion that extends to a second end portion, and a plurality of tube elements provided at the second end portion. Each of the plurality of tube elements defining a fluid passage includes a body having a first end section that extends to a second end section. The second end section projects beyond the second end portion of the injection nozzle assembly.

  8. Marangoni flow on an inkjet nozzle plate

    NARCIS (Netherlands)

    de Jong, J.; Reinten, Hans; Wijshoff, H.; Wijshoff, Herman; van den Berg, Marc; Delescen, Koos; van Dongen, Rini; Mugele, Friedrich Gunther; Versluis, Michel; Lohse, Detlef

    2007-01-01

    In piezo inkjet printing, nozzle failures are often caused by an ink layer on the nozzle plate. It is experimentally shown that the ink layer at the nozzle is formed through streamers of ink, emanating from a central ink band on the nozzle plate. The streamers propagate over a wetting nanofilm of

  9. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    International Nuclear Information System (INIS)

    Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-01-01

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  10. Axisymmetric thrust-vectoring nozzle performance prediction

    International Nuclear Information System (INIS)

    Wilson, E. A.; Adler, D.; Bar-Yoseph, P.Z

    1998-01-01

    Throat-hinged geometrically variable converging-diverging thrust-vectoring nozzles directly affect the jet flow geometry and rotation angle at the nozzle exit as a function of the nozzle geometry, the nozzle pressure ratio and flight velocity. The consideration of nozzle divergence in the effective-geometric nozzle relation is theoretically considered here for the first time. In this study, an explicit calculation procedure is presented as a function of nozzle geometry at constant nozzle pressure ratio, zero velocity and altitude, and compared with experimental results in a civil thrust-vectoring scenario. This procedure may be used in dynamic thrust-vectoring nozzle design performance predictions or analysis for civil and military nozzles as well as in the definition of initial jet flow conditions in future numerical VSTOL/TV jet performance studies

  11. A study on the relief of shell wall thinning of high pressure feedwater heater

    International Nuclear Information System (INIS)

    Kim, Hyung Joon; Park, Sang Hoon; Seo, Hyuk Ki; Kim, Kyung Hoon; Hwang, Kyung Mo

    2008-01-01

    Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which will increase as operating time progresses. Several nuclear power plants in Korea have experienced wall thinning damage in the area around the impingement baffle-installed downstream of the high pressure turbine extraction stream line- inside number 5A and 5B feedwater heaters. At that point, the extracted steam from the high pressure turbine is two phase fluid at high temperature, high pressure, and high speed. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of the number 5 high pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes operation of experience and numerical analysis composed similar condition with real high pressure feedwater heater. This study applied squared, curved and new type impingement baffle plates to feedwater heater same as previous study. In addition, it shows difference of pressure distribution and value between single phase and two phase based on experience and numerical analysis

  12. Coordination Control of SMR-Based NSSS Modules Integrated by Feedwater Distribution

    Science.gov (United States)

    Dong, Zhe; Song, Maoxuan; Huang, Xiaojin; Zhang, Zuoyi; Wu, Zongxin

    2016-10-01

    Due to its strong safety feature, the small modular reactor whose electric output is no more than 300MWe has been seen as a promising trend in nuclear engineering. By adopting multi-modular scheme, i.e. the superheated steam flows produced by multiple SMR-based nuclear heating system (NSSS) modules are combined to drive a common thermal load, the strong safety feature of a SMR can be applied to large-scale nuclear plants. To improve the economic competitiveness, it is meaningful to integrate multiple NSSS modules by the scheme of feedwater distribution, i.e. sharing a common pump and distributing feedwater by adjusting the opening of regulating valve of each module. The module coordination control of multiple SMR-based NSSS modules coupled by feedwater distribution is essentially the flowrate-pressure control of the common secondary-loop fluid flow network (FFN). In this paper, the nonlinear differential-algebraic model for the FFNs with a single feedwater pump is first given. A novel distributed adaptive flowrate-pressure control is proposed, which is then applied to realize the module coordination. Numerical simulation results in the case of coordination control of two MHTGR-based NSSS modules integrated by feedwater distribution scheme show the feasibility as well as the satisfactory transient performance of this newly-built coordination control law.

  13. Reactor pressure vessel nozzle

    Science.gov (United States)

    Challberg, Roy C.; Upton, Hubert A.

    1994-01-01

    A nozzle for joining a pool of water to a nuclear reactor pressure vessel includes a tubular body having a proximal end joinable to the pressure vessel and a distal end joinable in flow communication with the pool. The body includes a flow passage therethrough having in serial flow communication a first port at the distal end, a throat spaced axially from the first port, a conical channel extending axially from the throat, and a second port at the proximal end which is joinable in flow communication with the pressure vessel. The inner diameter of the flow passage decreases from the first port to the throat and then increases along the conical channel to the second port. In this way, the conical channel acts as a diverging channel or diffuser in the forward flow direction from the first port to the second port for recovering pressure due to the flow restriction provided by the throat. In the backflow direction from the second port to the first port, the conical channel is a converging channel and with the abrupt increase in flow area from the throat to the first port collectively increase resistance to flow therethrough.

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  15. Investigation and evaluation of cracking incidents in piping in pressurized water reactors. Technical report

    International Nuclear Information System (INIS)

    1980-09-01

    This report summarizes an investigation of known cracking incidents in pressurized water reactor plants. Several instances of cracking in feedwater piping in 1979, together with reported cases of stress corrosion cracking at Three Mile Island Unit 1, led to the establishment of the third Pipe Crack Study Group. Major differences between the scope of the third PCSG and the previous two are: (1) the emphasis given to systems safety implications of cracking, and (2) the consideration given all cracking mechanisms known to affect PWR piping, including the failure of small lines in secondary safety systems. The present PCSG reviewed existing information on cracking of PWR pipe systems, either contained in written records of collected from meetings in the United States, and made recommendations in response to the PCSG charter questions and to othe major items that may be considered to either reduce the potential for cracking or to improve licensing bases

  16. Design enhancement in BWR feedwater control system: Experience feedback from Kuosheng NPP

    International Nuclear Information System (INIS)

    Chen, C. C.

    2006-01-01

    Though feedwater control system is used only for normal operation of the plant, it belongs to the category cited by SRP chapter 7.7 that has the potential to affect the performance of critical safety functions. Kousheng has been commercial operation for more than 20 years. Because of equipment aging problem and difficulties in getting spare parts, feedwater control system needs upgrade. Unit II feedwater control system was upgraded to modern digital control system in EOC-17 and Unit I was upgraded in EOC-18. Before the implementation of the project, Taipower has accumulated various experiences from previous digital control system retrofits in Chin-Shan, Kousheng and Maanshan NPP. A task force lead by author is formed to assist Kousheng. This paper shares the experience of Kousheng retrofit project. (authors)

  17. Soft-Sensing for the Feedwater Flowrate at PWRs Using a GMDH Algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Su; Lee, Sung Han; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of); Lim, Dong Hyuk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2009-10-15

    Currently, Venturi flow meters are being used to measure the feedwater flowrate in most pressurized water reactors (PWRs). These meters can induce measurement drift owing to corrosion product buildup near the meter orifice through long-term operation. Fouling of the Venturi meter decreases the accuracy of the existing hardware sensors. These requirements result in nuclear power plants operating at lower-than-planned power levels. Therefore, considerable research has been focused on resolving the inaccurate measurement issue of the feedwater flowrate. This study employed the Group Method of Data Handling (GMDH) which is one of the data driven models, such as Artificial Neural Networks (ANNs), to increase the thermal efficiency by accurately estimating online the feedwater flowrate. The GMDH method was trained using the informative data selected through a subtractive clustering (SC) scheme. In addition, the uncertainty of the GMDH algorithm was analyzed using a statistical uncertainty method.

  18. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Pappx, L.

    1994-01-01

    After modification of Dukovany NPP steam generator feedwater system, the increased concentration of minerals was measured in the cold leg of modified steam generator. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators, has focused this attention on the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of flow distribution in the secondary side of SG was developed. (Author)

  19. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed

  20. Refined analysis of effects on the LOFT containment building resulting from main steam and feedwater breaks

    Energy Technology Data Exchange (ETDEWEB)

    Mosby, W.R.

    1979-01-16

    Dynamic one-degree-of-freedom analyses of main steam breaks 8/10B and 8/11 and feedwater break 9/40 were performed using refined jet-plus-reaction force-time functions. Feedwater breaks 10/8B and 10/11 were analyzed statically using a refined maximum jet-plus-reaction force and assuming a dynamic load factor of 2. None of the breaks were found to stress the containment shell in excess of the ASME Boiler and Pressure Vessel Code allowables.

  1. Influence of feedwater and blowdown systems on the mineral distribution in WWER steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Papp, L. [Inst. of Material Engineering, Ostrava (Switzerland)

    1995-12-31

    After modification of Dukovany NPP steam generator (SG) feedwater system, the increased concentration of minerals was measured in the cold leg of modified SG. Some modifications were performed on operating WWER 1000 steam generators with aim to optimize the water chemistry in the collectors area. Since the distribution of minerals can substantially affect on corrosion processes in steam generators, VITKOVICE, as a producer of WWER steam generators has focused the attention to the optimizing of these systems. To predict the mineral distribution on the secondary side of steam generators for considered feedwater/blowdown systems, the simple model of the flow distribution in the secondary side of SG was developed.

  2. Cyclic elastic analysis of a PWR nozzle subjected to a repeated thermal shock

    International Nuclear Information System (INIS)

    Locci, J.M.; Prost, J.P.

    1979-01-01

    In the primary piping system of a PWR nuclear power plant, some nozzles are subjected to strong thermal shocks due to sudden thermal variations in the internal water flow. The thermal gradients are sufficiently high to induce general elastic plastic behaviour. The design of these nozzles using the simplified elastic plastic analysis given in the ASME III Code NB-3200 generally leads to a very high usage factor. The aim of this work is to show by giving an example that a complete cyclic elastic plastic analysis makes it possible to considerably reduce the usage factor. (orig.)

  3. Additional Stress And Fracture Mechanics Analyses Of Pressurized Water Reactor Pressure Vessel Nozzles

    International Nuclear Information System (INIS)

    Walter, Matthew; Yin, Shengjun; Stevens, Gary; Sommerville, Daniel; Palm, Nathan; Heinecke, Carol

    2012-01-01

    In past years, the authors have undertaken various studies of nozzles in both boiling water reactors (BWRs) and pressurized water reactors (PWRs) located in the reactor pressure vessel (RPV) adjacent to the core beltline region. Those studies described stress and fracture mechanics analyses performed to assess various RPV nozzle geometries, which were selected based on their proximity to the core beltline region, i.e., those nozzle configurations that are located close enough to the core region such that they may receive sufficient fluence prior to end-of-life (EOL) to require evaluation of embrittlement as part of the RPV analyses associated with pressure-temperature (P-T) limits. In this paper, additional stress and fracture analyses are summarized that were performed for additional PWR nozzles with the following objectives: To expand the population of PWR nozzle configurations evaluated, which was limited in the previous work to just two nozzles (one inlet and one outlet nozzle). To model and understand differences in stress results obtained for an internal pressure load case using a two-dimensional (2-D) axi-symmetric finite element model (FEM) vs. a three-dimensional (3-D) FEM for these PWR nozzles. In particular, the ovalization (stress concentration) effect of two intersecting cylinders, which is typical of RPV nozzle configurations, was investigated. To investigate the applicability of previously recommended linear elastic fracture mechanics (LEFM) hand solutions for calculating the Mode I stress intensity factor for a postulated nozzle corner crack for pressure loading for these PWR nozzles. These analyses were performed to further expand earlier work completed to support potential revision and refinement of Title 10 to the U.S. Code of Federal Regulations (CFR), Part 50, Appendix G, Fracture Toughness Requirements, and are intended to supplement similar evaluation of nozzles presented at the 2008, 2009, and 2011 Pressure Vessels and Piping (PVP

  4. Nozzle geometry for organic vapor jet printing

    Science.gov (United States)

    Forrest, Stephen R.; McGraw, Gregory

    2017-10-25

    A first device is provided. The device includes a print head. The print head further includes a first nozzle hermetically sealed to a first source of gas. The first nozzle has an aperture having a smallest dimension of 0.5 to 500 microns in a direction perpendicular to a flow direction of the first nozzle. At a distance from the aperture into the first nozzle that is 5 times the smallest dimension of the aperture of the first nozzle, the smallest dimension perpendicular to the flow direction is at least twice the smallest dimension of the aperture of the first nozzle.

  5. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    International Nuclear Information System (INIS)

    Kim, Hyeonmin; Kang, Yeon Kwan; Heo, Gyunyoung; Song, Seok Yoon

    2014-01-01

    It is inevitable to avoid degradation of component, which operates continuously for long time in harsh environment. Since this degradation causes economical loss and human loss, it is important to monitor and diagnose the degradation of component. The diagnosis requires a well-systematic method for timely decision. Before this article, the methods using regression model and diagnosis table have been proposed to perform the diagnosis study for thermal efficiency in Nuclear Power Plants (NPPs). Since the regression model was numerically less-stable under changes of operating variables, it was difficult to provide good results in operating plants. Contrary to this, the diagnosis table was hard to use due to ambiguous points and to detect how it affects degradation. In order to cover the issues of previous researches, we proposed fuzzy approaches and applied it to diagnose Feedwater Heater (FWH) degradation to check the feasibility. The degradation of FWHs is not easy to be observed, while trouble such as tube leakage may bring simultaneous damage to the tube bundle. This study explains the steps of diagnosing typical failure modes of FWHs. In order to cover the technical issues of previous researches, we adopted fuzzy logic to suggest a diagnosis algorithm for the degradation of FHWs and performed feasibility study. In this paper, total 7 modes of FWH degradation modes are considered, which are High Drain Level, Low Shell Pressure, Tube Pressure Increase, Tube Fouling, Pass Partition Plate Leakage, Tube Leakage, Abnormal venting. From the literature survey and simulation, diagnosis table for FWH is made. We apply fuzzy logic based on diagnosis table. Authors verify fuzzy diagnosis for FWH degradation synthesized the random input sets from made diagnosis table. Comparing previous researches, suggested method more-stable under changes of operating variables, than regression model. On the contrary, the problem which ambiguous points and detect how it affects degradation

  6. Heat pipes

    CERN Document Server

    Dunn, Peter D

    1982-01-01

    A comprehensive, up-to-date coverage of the theory, design and manufacture of heat pipes and their applications. This latest edition has been thoroughly revised, up-dated and expanded to give an in-depth coverage of the new developments in the field. Significant new material has been added to all the chapters and the applications section has been totally rewritten to ensure that topical and important applications are appropriately emphasised. The bibliography has been considerably enlarged to incorporate much valuable new information. Thus readers of the previous edition, which has established

  7. Prototype Morphing Fan Nozzle Demonstrated

    Science.gov (United States)

    Lee, Ho-Jun; Song, Gang-Bing

    2004-01-01

    Ongoing research in NASA Glenn Research Center's Structural Mechanics and Dynamics Branch to develop smart materials technologies for aeropropulsion structural components has resulted in the design of the prototype morphing fan nozzle shown in the photograph. This prototype exploits the potential of smart materials to significantly improve the performance of existing aircraft engines by introducing new inherent capabilities for shape control, vibration damping, noise reduction, health monitoring, and flow manipulation. The novel design employs two different smart materials, a shape-memory alloy and magnetorheological fluids, to reduce the nozzle area by up to 30 percent. The prototype of the variable-area fan nozzle implements an overlapping spring leaf assembly to simplify the initial design and to provide ease of structural control. A single bundle of shape memory alloy wire actuators is used to reduce the nozzle geometry. The nozzle is subsequently held in the reduced-area configuration by using magnetorheological fluid brakes. This prototype uses the inherent advantages of shape memory alloys in providing large induced strains and of magnetorheological fluids in generating large resistive forces. In addition, the spring leaf design also functions as a return spring, once the magnetorheological fluid brakes are released, to help force the shape memory alloy wires to return to their original position. A computerized real-time control system uses the derivative-gain and proportional-gain algorithms to operate the system. This design represents a novel approach to the active control of high-bypass-ratio turbofan engines. Researchers have estimated that such engines will reduce thrust specific fuel consumption by 9 percent over that of fixed-geometry fan nozzles. This research was conducted under a cooperative agreement (NCC3-839) at the University of Akron.

  8. Silver clusters from nozzle expansions

    International Nuclear Information System (INIS)

    Hagena, O.F.

    1990-01-01

    This note reports on the first successful experiments to generate silver clusters (N≤100) in supersonic nozzle flows. A mixture of argon/silver-vapor was used expanding from a conical nozzle (0.35 mm, 10deg full cone angle, 17 mm long conical section). Source temperature and total pressure ranged up to 2200 K/300 kPa, and silver partial pressure up to 25 kPa. The data confirm the scaling laws developed to compare clustering of metals with that of rare gases. (orig.)

  9. The water treatment in the dual-purpose nuclear plants of Babcock and Wilcox with straight pipes

    International Nuclear Information System (INIS)

    Martynova, O.I.

    1978-01-01

    A report is given on water processing and water chemistry in the dual-purpose nuclear power plants (as compared to the single-purpose nuclear power plants) of Babcock and Wilcox, with flow steam generators with straight pipes. The most important materials, especially regarding their corrosion resistance, and the water composition during 'hot' start-up of the Okonie-I power plant, the quality factors of the feedwater, the water quality factors of the steam generator with fast start-up and the experience with numerous corrosion-caused defects in steam generator pipes are dealt with from the aspect of optimum water processing and successful continuous operation. (HK) [de

  10. Calculation of the limiting CESSAR Feedwater Line-Break and Steam Line-Break transients

    International Nuclear Information System (INIS)

    Peeler, G.B.; Kennedy, M.F.; Caraher, D.L.; Guttmann, J.; Chung, K.S.

    1983-01-01

    Argonne National Laboratory (ANL), under contract to the Nuclear Regulatory Commission, performed audit calculations of the limiting Feedwater Line Break (FLB) and Steam Line Break (SLB) transients presented in the CESSAR FSAR. The results of the FLB and SLB calculations are discussed

  11. Collector feedwater supply and stability of the power distribution in a pressurized-water reactor

    International Nuclear Information System (INIS)

    Budnikov, V.I.; Kosolapov, S.V.; Kramerov, A.Ya.

    1980-01-01

    It is necessary to determine how the collector feedwater supply affects the disposition of the stability limits and the instability period for the power distribution in such a reactor. The main reason for the fluctuations in feedwater flow rate were shown by additional calculations with the general power regulator switched out to be due to instability on the fundamental in the neutron distribution. The power-level fluctuations are due to oscillation of the feed valve in the level regulator, and consequently to oscillations in the feedwater flow rate. If collector feed is to be employed, it is desirable to improve the response of the pressure control system for the separator drum, because under certain emergency conditions there will be a considerable fall in pressure in the separator drum. The deviation from saturation for the water in the separator drum tube is less in the second method than it is in the first, so the cavitation margin in the principal pumps may be reduced somewhat. Calculations show that this reduction will not occur if the time constant of the turbine synchronizer is about 10 sec. Also, the dynamic characteristics of the nuclear power station in these modes of feedwater supply are appreciably influenced by the parameters of the pressure-control system and the water-level control for the separator drum

  12. 77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Science.gov (United States)

    2012-03-16

    ... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

  13. 77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Science.gov (United States)

    2012-09-11

    ... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Systems for Boiling Water Reactor Power Plants.'' This regulatory guide is being revised to: (1) Expand... for the condensate and feedwater systems in all types of light water reactor facilities; and (2) to...

  14. VGB conference 'Chemistry in the power plant 1984' - VGB feedwater conditioning conference

    International Nuclear Information System (INIS)

    1984-01-01

    The conference bears various aspects of feedwater conditioning for power plant cooling systems and steam generators as well as on the analytical assessment of water quality and its translation into operational method approaches. 5 out of the total 14 papers were entered separately in the database. (RB) [de

  15. Turbomachine combustor nozzle including a monolithic nozzle component and method of forming the same

    Science.gov (United States)

    Stoia, Lucas John; Melton, Patrick Benedict; Johnson, Thomas Edward; Stevenson, Christian Xavier; Vanselow, John Drake; Westmoreland, James Harold

    2016-02-23

    A turbomachine combustor nozzle includes a monolithic nozzle component having a plate element and a plurality of nozzle elements. Each of the plurality of nozzle elements includes a first end extending from the plate element to a second end. The plate element and plurality of nozzle elements are formed as a unitary component. A plate member is joined with the nozzle component. The plate member includes an outer edge that defines first and second surfaces and a plurality of openings extending between the first and second surfaces. The plurality of openings are configured and disposed to register with and receive the second end of corresponding ones of the plurality of nozzle elements.

  16. Reactor pressure vessel with forged nozzles

    Science.gov (United States)

    Desai, Dilip R.

    1993-01-01

    Inlet nozzles for a gravity-driven cooling system (GDCS) are forged with a cylindrical reactor pressure vessel (RPV) section to which a support skirt for the RPV is attached. The forging provides enhanced RPV integrity around the nozzle and substantial reduction of in-service inspection costs by eliminating GDCS nozzle-to-RPV welds.

  17. Integrated experimental test program on waterhammer pressure pulses and associated structural responses within a feedwater sparger

    Energy Technology Data Exchange (ETDEWEB)

    Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland)

    1997-12-31

    This paper describes the methods and systems as utilized in an integrated experimental thermohydraulic/mechanics analysis test program on waterhammer pressure pulses within a revised feedwater sparger of a Loviisa generation VVER-440-type reactor. This program was carried out in two stages: (1) measurements with a strictly limited set of operating parameters at Loviisa NPP, and (2) measurements with the full set of operating parameters on a test article simulating the revised feedwater sparger. The experiments at Loviisa NPS served as an invaluable source of information on the nature of waterhammer pressure pulses and structural responses. These tests thus helped to set the objectives and formulate the concept for series of tests on a test article to study the water hammer phenomena. The heavily instrumented full size test article of a steam generator feedwater sparger was placed within a pressure vessel simulating the steam generator. The feedwater sparger was subjected to the full range of operating parameters which were to result in waterhammer pressure pulse trains of various magnitudes and duration. Two different designs of revised feedwater sparger were investigated (i.e. `grounded` and `with goose neck`). The following objects were to be met within this program: (1) establish the thermohydraulic parameters that facilitate the occurrence of water hammer pressure pulses, (2) provide a database for further analysis of the pressure pulse phenomena, (3) establish location and severity of these water hammer pressure pulses, (4) establish the structural response due to these pressure pulses, (5) provide input data for structural integrity analysis. (orig.). 3 refs.

  18. Parametric Study of Sealant Nozzle

    Science.gov (United States)

    Yamamoto, Yoshimi

    It has become apparent in recent years the advancement of manufacturing processes in the aerospace industry. Sealant nozzles are a critical device in the use of fuel tank applications for optimal bonds and for ground service support and repair. Sealants has always been a challenging area for optimizing and understanding the flow patterns. A parametric study was conducted to better understand geometric effects of sealant flow and to determine whether the sealant rheology can be numerically modeled. The Star-CCM+ software was used to successfully develop the parametric model, material model, physics continua, and simulate the fluid flow for the sealant nozzle. The simulation results of Semco sealant nozzles showed the geometric effects of fluid flow patterns and the influences from conical area reduction, tip length, inlet diameter, and tip angle parameters. A smaller outlet diameter induced maximum outlet velocity at the exit, and contributed to a high pressure drop. The conical area reduction, tip angle and inlet diameter contributed most to viscosity variation phenomenon. Developing and simulating 2 different flow models (Segregated Flow and Viscous Flow) proved that both can be used to obtain comparable velocity and pressure drop results, however; differences are seen visually in the non-uniformity of the velocity and viscosity fields for the Viscous Flow Model (VFM). A comprehensive simulation setup for sealant nozzles was developed so other analysts can utilize the data.

  19. Implementation of a digital feedwater control system at Dresden Nuclear Power Plant, Units 2 and 3: Final report

    International Nuclear Information System (INIS)

    Zapotocky, A.; Popovic, J.R.; Fournier, R.D.

    1988-12-01

    This report describes the Digital Feedwater Control System Implementation at the Dresden 2 or 3 Units of the BWR Nuclear Power Plant owned by the Commonwealth Edison Company. The digital system has been operational in Unit 3 since August 1986, and in Unit 2 since April 1987. The Bailey Control's Network 90 based digital control system replaced the obsolete GE/MAC 5000 analog control system in the reactor feedwater control loop as a ''like-for-like'' replacement. Operational experience from the Digital Feedwater Control installations has been good and the system demonstrated better performance than the old analog systems. 14 refs., 15 figs., 17 tabs

  20. Impact of the operation of non-displaced feedwater heaters on the performance of Solar Aided Power Generation plants

    International Nuclear Information System (INIS)

    Qin, Jiyun; Hu, Eric; Nathan, Graham J.

    2017-01-01

    Highlights: • Impact of non-displaced feedwater heater on plant’s performance has been evaluated. • Two operation strategies for non-displaced feedwater heater has been proposed. • Constant temperature strategy is generally better. • Constant mass flow rate strategy is suit for rich solar thermal input. - Abstract: Solar Aided Power Generation is a technology in which low grade solar thermal energy is used to displace the high grade heat of the extraction steam in a regenerative Rankine cycle power plant for feedwater preheating purpose. The displaced extraction steam can then expand further in the steam turbine to generate power. In such a power plant, using the (concentrated) solar thermal energy to displace the extraction steam to high pressure/temperature feedwater heaters (i.e. displaced feedwater heaters) is the most popular arrangement. Namely the extraction steam to low pressure/temperature feedwater heaters (i.e. non-displaced feedwater heaters) is not displaced by the solar thermal energy. In a Solar Aided Power Generation plants, when solar radiation/input changes, the extraction steam to the displaced feedwater heaters requires to be adjusted according to the solar radiation. However, for the extraction steams to the non-displaced feedwater heaters, it can be either adjusted accordingly following so-called constant temperature strategy or unadjusted i.e. following so-called constant mass flow rate strategy, when solar radiation/input changes. The previous studies overlooked the operation of non-displaced feedwater heaters, which has also impact on the whole plant’s performance. This paper aims to understand/reveal the impact of the two different operation strategies for non-displaced feedwater heaters on the plant’s performance. In this paper, a 300 MW Rankine cycle power plant, in which the extraction steam to high pressure/temperature feedwater heaters is displaced by the solar thermal energy, is used as study case for this purpose. It

  1. Experimental characterization of spin motor nozzle flow.

    Energy Technology Data Exchange (ETDEWEB)

    Erven, Rocky J.; Peterson, Carl Williams; Henfling, John Francis

    2006-11-01

    The Mach number in the inviscid core of the flow exiting scarfed supersonic nozzles was measured using pitot probes. Nozzle characterization experiments were conducted in a modified section of an obsolete M = 7.3 test section/nozzle assembly on Sandia's Hypersonic Wind Tunnel. By capitalizing on existing hardware, the cost and time required for tunnel modifications were significantly reduced. Repeatability of pitot pressure measurements was excellent, and instrumentation errors were reduced by optimizing the pressure range of the transducers used for each test run. Bias errors in probe position prevented us from performing a successful in situ calibration of probe angle effects using pitot probes placed at an angle to the nozzle centerline. The abrupt throat geometry used in the Baseline and Configuration A and B nozzles modeled the throat geometry of the flight vehicle's spin motor nozzles. Survey data indicates that small (''unmeasurable'') differences in the nozzle throat geometries produced measurable flow asymmetries and differences in the flow fields generated by supposedly identical nozzles. Therefore, data from the Baseline and Configuration A and B nozzles cannot be used for computational fluid dynamics (CFD) code validation. Configuration C and D nozzles replaced the abrupt throat geometry of Baseline and Configuration A and B nozzles with a 0.500-inch streamwise radius of curvature in the throat region. This throat geometry eliminated the flow asymmetries, flow separation in the nozzle throat, and measurable differences between the flow fields from identical nozzles that were observed in Baseline/A/B nozzles. Data from Configuration C and D nozzles can be used for CFD code validation.

  2. Shield For Flexible Pipe

    Science.gov (United States)

    Ponton, Michael K.; Williford, Clifford B.; Lagen, Nicholas T.

    1995-01-01

    Cylindrical shield designed to fit around flexible pipe to protect nearby workers from injury and equipment from damage if pipe ruptures. Designed as pressure-relief device. Absorbs impact of debris ejected radially from broken flexible pipe. Also redirects flow of pressurized fluid escaping from broken pipe onto flow path allowing for relief of pressure while minimizing potential for harm.

  3. Appropriate nominal stresses for use with ASME Code pressure-loading stress indices for nozzles

    International Nuclear Information System (INIS)

    Rodabaugh, E.C.

    1976-06-01

    This program is part of a cooperative effort with industry to develop and verify analytical methods for assessing the safety of nuclear pressure-vessel and piping-system design. The study of nominal stresses and stress indices described is part of a continuing study of design rules for nozzles in pressure vessels being coordinated by the PVRC Subcommittee on Reinforced Openings and External Loadings. Results from these studies are used by appropriate ASME Code groups in drafting new and improved design rules

  4. Simulation of a Downsized FDM Nozzle

    DEFF Research Database (Denmark)

    Hofstätter, Thomas; Pimentel, Rodrigo; Pedersen, David B.

    2015-01-01

    This document discusses the simulat-ion of a downsized nozzle for fused deposition modelling (FDM), namely the E3D HotEnd Extruder with manufactured diameters of 200-400 μm in the nozzle tip. The nozzle has been simulated in terms of heat transfer and fluid flow giving an insight into the physical...... behavior of the polymer inside the nozzle. The extruder contains a nozzle, a heater block, a heatbreak and a heatsink additionally cooled by a fan. The diameter is located in the sub-mm re-gion allowing to reduce the size and surface roughness of the product. The simulation results were experimentally...

  5. Development of cooling techniques for induction heating stress improvement of reactor recirculation inlet nozzle

    International Nuclear Information System (INIS)

    Takahashi, Shirou; Shiina, Kouji; Nihei, Kenichi; Kanno, Satoshi; Hayashi, Shoji; Ootaka, Minoru

    2007-01-01

    Induction heating stress improvement (IHSI) has been used in nuclear power plants to reduce residual stress in welded sections of pipes by generating temperature differences between the inner and outer surfaces of the pipes. The outer metal surface is heated by induction heating, and the inner surface is cooled by flowing water. However, it is difficult to obtain a sufficient temperature gradient in the places where the flow stagnates and the heat transfer cannot be enhanced. In the present study, we developed cooling techniques for a reactor recirculation inlet nozzle with a closed end and very narrow annular channel. Computational fluid dynamics (CFD) analyses, half-scale tests, and full-scale tests were conducted to investigate the flow and cooling characteristics. One million grids of a reactor recirculation inlet nozzle model were used for the CFD analysis. Detached eddy simulation (DES) was used as the turbulence model to evaluate the unsteady phenomena of the jet flow and temperature distribution. The experimental apparatuses used for the half-scale tests were made of acryl to visualize the flow, and heat transfer coefficients were measured at the welded portions. In the full-scale tests, the temperature differences between the inner and outer surface of the recirculation inlet nozzle were measured, and reduction of the residual stress was verified. It was confirmed that the jet flow moved up and down when to jet nozzles were arranged symmetrically. The turbulence due to self-sustained jet fluctuation was effective for uniform cooling in the reactor recirculation inlet nozzle. The flow did not stagnate around the welded portion. The heat transfer coefficients at the welded portion were evaluated using an equation with Reynolds and Nusselt numbers in half-scale tests. It was also verified in full scale tests that the temperature difference between the inner and outer surfaces of the recirculation inlet nozzle was approximately 490degC, which satisfied the

  6. Fundamentals of piping design

    CERN Document Server

    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

  7. Nozzle geometry variations on the discharge coefficient

    Directory of Open Access Journals (Sweden)

    M.M.A. Alam

    2016-03-01

    Full Text Available Numerical works have been conducted to investigate the effect of nozzle geometries on the discharge coefficient. Several contoured converging nozzles with finite radius of curvatures, conically converging nozzles and conical divergent orifices have been employed in this investigation. Each nozzle and orifice has a nominal exit diameter of 12.7×10−3 m. A 3rd order MUSCL finite volume method of ANSYS Fluent 13.0 was used to solve the Reynolds-averaged Navier–Stokes equations in simulating turbulent flows through various nozzle inlet geometries. The numerical model was validated through comparison between the numerical results and experimental data. The results obtained show that the nozzle geometry has pronounced effect on the sonic lines and discharge coefficients. The coefficient of discharge was found differ from unity due to the non-uniformity of flow parameters at the nozzle exit and the presence of boundary layer as well.

  8. Insulated pipe clamp design

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. The effects of location, thermal stress, and residual stress on corner cracks in nozzles with cladding

    International Nuclear Information System (INIS)

    Besuner, P.M.; Cohen, L.M.; McLean, J.L.

    1977-01-01

    The stress intensity factors (Ksub(I)) for corner cracks in a boiling water reactor feedwater nozzle with stainless steel cladding are obtained for loading by internal pressure, and a fluid quench in the nozzle. Conditions with and without residual stress in the component are considered. The residual stress is simulated by means of a reference temperature change. The stress distribution for the uncracked structure is obtained from a three-dimensional finite element model. A three-dimensional influence function (IF) method, in conjunction with the boundary-integral equation method for structural analysis is employed to compute Ksub(I) values from the uncracked structure's stress distribution. It is concluded that the effects on Ksub(I) of location, thermal stresses, and residual stresses are significant and generally too complex to evaluate without advanced numerical procedures. The ulilized combination of finite element analysis of the uncracked structure and three-dimensional influence function analysis of the cracked structure is demonstrated and endorsed. (Auth.)

  10. Reconciliation of equipment flexibility effects on piping system dynamic response

    International Nuclear Information System (INIS)

    Geraets, L.H.

    1987-01-01

    Piping systems are connected to equipment; if the equipment cannot be considered as ''rigid'' relative to excitation frequencies, nozzle response spectra techniques, or equipment modeling techniques are used. If the equipment is considered rigid, a fixed anchor is assumed. However, occasionally after (seismic) dynamic analysis has been completed, tests or detailed equipment dynamic analyses demonstrate that the assumption of ''infinite stiff'' is questionable. This paper reviews several classes of equipment (pumps, vessels, reservoirs, heat exchangers), and the associated (piping stresses, support loads, equipment nozzle allowables). Significant divergences between design and ''as built'' results are shown (for heat exchangers in particular). The paper discusses the reconciliation process performed for a belgian PWR plant through the use of less conservative seismic damping data (Code Case N-411)

  11. Fluid Flow Nozzle Energy Harvesters

    Science.gov (United States)

    Sherrit, Stewart; Lee, Hyeong Jae; Walkenmeyer, Phillip; Winn, Tyler; Tosi, Luis Phillipe; Colonius, Tim

    2015-01-01

    Power generation schemes that could be used downhole in an oil well to produce about 1 Watt average power with long-life (decades) are actively being developed. A variety of proposed energy harvesting schemes could be used to extract energy from this environment but each of these has their own limitations that limit their practical use. Since vibrating piezoelectric structures are solid state and can be driven below their fatigue limit, harvesters based on these structures are capable of operating for very long lifetimes (decades); thereby, possibly overcoming a principle limitation of existing technology based on rotating turbo-machinery. An initial survey identified that spline nozzle configurations can be used to excite a vibrating piezoelectric structure in such a way as to convert the abundant flow energy into useful amounts of electrical power. This paper presents current flow energy harvesting designs and experimental results of specific spline nozzle/ bimorph design configurations which have generated suitable power per nozzle at or above well production analogous flow rates. Theoretical models for non-dimensional analysis and constitutive electromechanical model are also presented in this paper to optimize the flow harvesting system.

  12. Insulated pipe clamp design

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  13. Flexible ocean upwelling pipe

    Science.gov (United States)

    Person, Abraham

    1980-01-01

    In an ocean thermal energy conversion facility, a cold water riser pipe is releasably supported at its upper end by the hull of the floating facility. The pipe is substantially vertical and has its lower end far below the hull above the ocean floor. The pipe is defined essentially entirely of a material which has a modulus of elasticity substantially less than that of steel, e.g., high density polyethylene, so that the pipe is flexible and compliant to rather than resistant to applied bending moments. The position of the lower end of the pipe relative to the hull is stabilized by a weight suspended below the lower end of the pipe on a flexible line. The pipe, apart from the weight, is positively buoyant. If support of the upper end of the pipe is released, the pipe sinks to the ocean floor, but is not damaged as the length of the line between the pipe and the weight is sufficient to allow the buoyant pipe to come to a stop within the line length after the weight contacts the ocean floor, and thereafter to float submerged above the ocean floor while moored to the ocean floor by the weight. The upper end of the pipe, while supported by the hull, communicates to a sump in the hull in which the water level is maintained below the ambient water level. The sump volume is sufficient to keep the pipe full during heaving of the hull, thereby preventing collapse of the pipe.

  14. Using risk-informed asset management for feedwater system preventative maintenance optimization

    International Nuclear Information System (INIS)

    Kee, Ernest; Sun, Alice; Richards, Andrew; Grantom, Rick; Liming, James; Salter, James

    2004-01-01

    The initial development of a South Texas Project Nuclear Operating Company process for supporting preventative maintenance optimization by applying the Balance-Of-Plant model and Risk-Informed Asset Management alpha-level software applications is presented. Preventative maintenance activities are evaluated in the South Texas Project Risk-Informed Asset Management software while the plant maintains or improves upon high levels of nuclear safety. In the Balance-Of-Plant availability application, the level of detail in the feedwater system is enhanced to support plant decision-making at the component failure mode and human error mode level of indenture by elaborating on the current model at the super-component level of indenture. The enhanced model and modeling techniques are presented. Results of case studies in feedwater system preventative maintenance optimization sing plant-specific data are also presented. (author)

  15. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G.

    2014-10-01

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  16. Power-feedwater temperature operating domain for Sbwr applying Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar M, L. A.; Quezada G, S.; Espinosa M, E. G.; Vazquez R, A.; Varela H, J. R.; Cazares R, R. I.; Espinosa P, G., E-mail: sequega@gmail.com [Universidad Autonoma Metropolitana, Unidad Iztapalapa, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Mexico D. F. (Mexico)

    2014-10-15

    In this work the analyses of the feedwater temperature effects on reactor power in a simplified boiling water reactor (Sbwr) applying a methodology based on Monte Carlo simulation is presented. The Monte Carlo methodology was applied systematically to establish operating domain, due that the Sbwr are not yet in operation, the analysis of the nuclear and thermal-hydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. The results show that the reactor power is inversely proportional to the temperature of the feedwater, reactor power changes at 8% when the feed water temperature changes in 8%. (Author)

  17. Report of the US Nuclear Regulatory Commission Piping Review Committee. Volume 2. Evaluation of seismic designs: a review of seismic design requirements for Nuclear Power Plant Piping

    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.

  18. Equipment reliability and life cycle optimization of a nuclear plant feedwater heater

    International Nuclear Information System (INIS)

    Thomas, Daniel; Coakley, Michael; Catapano, Michael; Svensson, Eric

    2006-01-01

    Many papers published over the last 25 years have strongly emphasized the need for an ongoing program of inspection and testing with subsequent failure cause analysis of feedwater heaters. Plants must be run more competitively; therefore, Utilities must lower operation and maintenance costs, while optimizing overall plant efficiency and capacity factor. One recognized area that needs to be addressed in accomplishing this goal is the heat cycle. This paper specifically deals with the feedwater heating system. Utility engineers must monitor feedwater heater performance in order to recognize degradation, identify and mitigate failure mechanisms, and prevent in-service failures thereby optimizing availability. Periodic tube plugging without complete analysis of the degraded/failed areas resolves the immediate need for return to service; however, heater life will not be optimized. This paper illustrates a complete life cycle management inspection, testing, and maintenance program implemented at Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data and lack of root cause analysis, justified a program that included: - Removal of previously installed plugs; - Video-probe inspection of failed areas; - Extraction of tube samples for further analysis; - Eddy current testing of selected tubes; - Evaluation of the condition of 'insurance' plugged tubes for return to service; - Hydrostatic testing of selected individual tubes; - Final repair plan based on the results of the above program. This paper concludes that no single method of inspection or testing should solely be relied upon in establishing: - The extent of actual degraded conditions; - The mechanism(s) of failure; - The details of repair to be implemented. Evaluating all data affords the best chance in arresting problems and optimizing feedwater heater life. Problem heaters should be continuously monitored and inspected over time until the facts

  19. Developing the optimum boiler water and feedwater treatment for fossil plants

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, B. [Electric Power Research Inst., Palo Alto, California (United States)

    1996-12-01

    Over the last two years a new set of cycle chemistry guidelines has been developed for each of the treatments used in fossil plants. These revisions have been based on research conducted over the last ten years, much at the international collaborative level. By careful selection and optimization of the boiler water and feedwater treatments, it will be possible to accrue large financial, maintenance, availability and performance improvements. (au) 14 refs.

  20. Plant data comparisons for Comanche Peak 1/2 main feedwater pump trip transient

    Energy Technology Data Exchange (ETDEWEB)

    Boatwright, W.J.; Choe, W.G; Hiltbrand, D.W. [TU Electric, Dallas, TX (United States)] [and others

    1995-09-01

    A RETRAN-02 MOD5 model of Comanche Peak Steam Electric Station was developed by TU Electric for the purpose of performing core reload safety analyses. In order to qualify this model, comparisons against plant transient data from a partial loss of main feedwater flow were performed. These comparisons demonstrated that good representations of the plant response could be obtained with RETRAN-02 and the user-developed models of the primary-to-secondary heat transfer and plant control systems.

  1. Qualitative and Quantitative Analysis of Organic Impurities in Feedwater of a Heat-Recovery Steam Generator

    Science.gov (United States)

    Chichirov, A. A.; Chichirova, N. D.; Filimonova, A. A.; Gafiatullina, A. A.

    2018-03-01

    In recent years, combined-cycle units with heat-recovery steam generators have been constructed and commissioned extensively in the European part of Russia. By the example of the Kazan Cogeneration Power Station no. 3 (TETs-3), an affiliate of JSC TGK-16, the specific problems for most power stations with combined-cycle power units that stem from an elevated content of organic impurities in the feedwater of the heat-recovery steam generator (HRSG) are examined. The HRSG is fed with highly demineralized water in which the content of organic carbon is also standardized. It is assumed that the demineralized water coming from the chemical water treatment department of TETs-3 will be used. Natural water from the Volga River is treated to produce demineralized water. The results of a preliminary analysis of the feedwater demonstrate that certain quality indices, principally, the total organic carbon, are above the standard values. Hence, a comprehensive investigation of the feedwater for organic impurities was performed, which included determination of their structure using IR and UV spectroscopy techniques, potentiometric measurements, and element analysis; determination of physical and chemical properties of organic impurities; and prediction of their behavior in the HRSG. The estimation of the total organic carbon revealed that it exceeded the standard values in all sources of water comprising the feedwater for the HRSG. The extracted impurities were humic substances, namely, a mixture of humic and fulvic acids in a 20 : 80 ratio, respectively. In addition, an analysis was performed of water samples taken at all intermediate stages of water treatment to study the behavior of organic substances in different water treatment processes. An analysis of removal of the humus substances in sections of the water treatment plant yielded the concentration of organic substances on the HRSG condensate. This was from 100 to 150 μg/dm3. Organic impurities in boiler water can induce

  2. Common-cause failure analysis of McGuire Unit 2 auxiliary feedwater system

    International Nuclear Information System (INIS)

    Rasmuson, D.M.; Shepherd, J.C.; Fowler, R.D.; Summitt, R.L.; Logan, B.W.

    1982-01-01

    A powerful method for qualitative common cause failure analysis (CCFA) of nuclear power plant systems was developed by EG and G Idaho at the Idaho National Engineering Laboratory. As a cooperative project to demonstrate and evaluate the usefulness of the method, the Duke Power Company agreed to allow a CCFA of the auxiliary feedwater system (AFWS) in their McGuire Nuclear Station Unit 2. The results of the CCFA are the subject of this discussion

  3. Equipment reliability and life cycle optimization of a nuclear plant feedwater heater

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Daniel; Coakley, Michael [Exelon/Peach Bottom Atomic Power Station, 200 Exelon Way, Kennett Square, PA 19348 (United States); Catapano, Michael; Svensson, Eric [Powerfect, Inc. 9 Great Meadow Lane, East Hanover, N. J. 07936 (United States)

    2006-07-01

    Many papers published over the last 25 years have strongly emphasized the need for an ongoing program of inspection and testing with subsequent failure cause analysis of feedwater heaters. Plants must be run more competitively; therefore, Utilities must lower operation and maintenance costs, while optimizing overall plant efficiency and capacity factor. One recognized area that needs to be addressed in accomplishing this goal is the heat cycle. This paper specifically deals with the feedwater heating system. Utility engineers must monitor feedwater heater performance in order to recognize degradation, identify and mitigate failure mechanisms, and prevent in-service failures thereby optimizing availability. Periodic tube plugging without complete analysis of the degraded/failed areas resolves the immediate need for return to service; however, heater life will not be optimized. This paper illustrates a complete life cycle management inspection, testing, and maintenance program implemented at Peach Bottom Atomic Power Station (PBAPS). Concerns that tubes may have been too conservatively plugged due to insufficient data and lack of root cause analysis, justified a program that included: - Removal of previously installed plugs; - Video-probe inspection of failed areas; - Extraction of tube samples for further analysis; - Eddy current testing of selected tubes; - Evaluation of the condition of 'insurance' plugged tubes for return to service; - Hydrostatic testing of selected individual tubes; - Final repair plan based on the results of the above program. This paper concludes that no single method of inspection or testing should solely be relied upon in establishing: - The extent of actual degraded conditions; - The mechanism(s) of failure; - The details of repair to be implemented. Evaluating all data affords the best chance in arresting problems and optimizing feedwater heater life. Problem heaters should be continuously monitored and inspected over time until

  4. Pipe drafting and design

    CERN Document Server

    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

  5. Monitoring the performance of Aux. Feedwater Pump using Smart Sensing Model

    International Nuclear Information System (INIS)

    No, Young Gyu; Seong, Poong Hyun

    2015-01-01

    Many artificial intelligence (AI) techniques equipped with learning systems have recently been proposed to monitor sensors and components in NPPs. Therefore, the objective of this study is the development of an integrity evaluation method for safety critical components such as Aux. feedwater pump, high pressure safety injection (HPSI) pump, etc. using smart sensing models based on AI techniques. In this work, the smart sensing model is developed at first to predict the performance of Aux. feedwater pump by estimating flowrate using group method of data handing (GMDH) method. If the performance prediction is achieved by this feasibility study, the smart sensing model will be applied to development of the integrity evaluation method for safety critical components. Also, the proposed algorithm for the performance prediction is verified by comparison with the simulation data of the MARS code for station blackout (SBO) events. In this study, the smart sensing model for the prediction performance of Aux. feedwater pump has been developed. In order to develop the smart sensing model, the GMDH algorithm is employed. The GMDH algorithm is the way to find a function that can well express a dependent variable from independent variables. This method uses a data structure similar to that of multiple regression models. The proposed GMDH model can accurately predict the performance of Aux

  6. A soft-sensing model for feedwater flow rate using fuzzy support vector regression

    Energy Technology Data Exchange (ETDEWEB)

    Na, Man Gyun; Yang, Heon Young; Lim, Dong Hyuk [Chosun University, Gwangju (Korea, Republic of)

    2008-02-15

    Most pressurized water reactors use Venturi flow meters to measure the feedwater flow rate. However, fouling phenomena, which allow corrosion products to accumulate and increase the differential pressure across the Venturi flow meter, can result in an overestimation of the flow rate. In this study, a soft-sensing model based on fuzzy support vector regression was developed to enable accurate on-line prediction of the feedwater flow rate. The available data was divided into two groups by fuzzy c-means clustering in order to reduce the training time. The data for training the soft-sensing model was selected from each data group with the aid of a subtractive clustering scheme because informative data increases the learning effect. The proposed soft-sensing model was confirmed with the real plant data of Yonggwang Nuclear Power Plant Unit 3. The root mean square error and relative maximum error of the model were quite small. Hence, this model can be used to validate and monitor existing hardware feedwater flow meters.

  7. Loss-of-normal-feedwater sensitivity studies for AP600 behavior characterization

    International Nuclear Information System (INIS)

    Saiu, G.

    1996-01-01

    Activity concerning the development of a RELAP5/MOD3 model to simulate the Westinghouse Electric Corporation AP600 is summarized. The aim is to gain initial insight into the capability of RELAP5 to simulate the behavior of AP600 safety features. A-loss-of-normal-feedwater event is studied. Of the transients that must be investigated, this transient has been chosen to be one of the most relevant because the response of the AP600 to a loss-of-normal-feedwater event differs significantly from that of current pressurized water reactors in the extensive use of passive safety features peculiar to the AP600. Also, strong interactions among the AP600 safety systems, which should be further analyzed to permit full optimization of the system actuation logic and operation, are shown. Finally, a loss of normal feedwater without reactor scram, performed to investigate short-term plant behavior, shows that the pressure peak is affected by critical discharge flow coefficients applied to the pressurizer safety valves, while a relatively small reduction of the pressure peak is observed when both heat exchangers of the passive heat removal system are operating as opposed to the case in which only one is available. The data used for this study are derived from the Standard Safety Analysis Report configuration of the Westinghouse AP600 as of 1992

  8. Structural evaluation of IEA-R1 primary system pump nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel, E-mail: gfainer@ipen.br, E-mail: afaloppa@ipen.br, E-mail: calberto@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-11-01

    The IEA-R1 pumps of the primary coolant system may be required to withstand design and operational conditions. IEA-R1 nuclear research reactor is an open pool type reactor operated by IPEN since 1957. The reactor can be operated up to 5MW heating power since it was upgraded in a modernization program conducted by IPEN. The primary coolant system is composed by the piping system, decay tank, two heat pumps and two heat exchangers. In the latest arrangement upgrade of the primary system, conducted in 2014 as part of an aging management program, a partial replacement of the coolant piping and total replacement of piping and pump supports were done. As consequence, reviewed loads in the pump nozzles were obtained demanding a new evaluation of them. The aim of this report is to present the structural evaluation of the pump nozzles, considering the new loads coming from the new piping layout, according to: API 610 code verification, Supplier loads and structural analysis applying finite element method, by using the ANSYS computer program, regarding ASME VIII Div 1 & 2 recommendations. (author)

  9. Effects of location, thermal stress and residual stress on corner cracks in nozzles with cladding

    International Nuclear Information System (INIS)

    McLean, J.L.; Cohen, L.M.; Besuner, P.M.

    1979-01-01

    The stress intensity factors (K 1 ) for corner cracks in a boiling water reactor feedwater nozzle with stainless steel cladding are obtained for loading by internal pressure and a fluid quench in the nozzle. Conditions both with and without residual stress in the component are considered. The residual stress is simulated by means of a reference temperature change. The stress distribution for the uncracked structure is obtained from a three-dimensional finite element model. A three-dimensional influence function (IF) method, in conjunction with the boundary-integral equation method for structural analysis, is employed to compute K 1 values from the uncracked stress distribution. For each type of loading K 1 values are given for cracks at 15 nozzle locations and for 6 crack depths. Reasonable agreement is noted between calculated and previously published pressure-induced K 1 values. Comparisons are made to determine the effect on K 1 of crack location, thermal stress and residual stress, as compared with pressure stress. For the thermal transient it is shown that K 1 for small crack depths is maximised early in the transient, while K 1 for large cracks is maximised later under steady state conditions. Computation should, therefore, be made for several transient time points and the maximum K 1 for a given crack depth should be used for design analysis. It is concluded that the effects on K 1 of location, thermal stresses and residual stresses are significant and generally too complex to evaluate without advanced numerical procedures. The utilised combination of finite element analysis of the uncracked structure and three-dimensional influence function analysis of the cracked structure is demonstrated and endorsed. (author)

  10. Status of Nozzle Aerodynamic Technology at MSFC

    Science.gov (United States)

    Ruf, Joseph H.; McDaniels, David M.; Smith, Bud; Owens, Zachary

    2002-01-01

    This viewgraph presentation provides information on the status of nozzle aerodynamic technology at MSFC (Marshall Space Flight Center). The objectives of this presentation were to provide insight into MSFC in-house nozzle aerodynamic technology, design, analysis, and testing. Under CDDF (Center Director's Discretionary Fund), 'Altitude Compensating Nozzle Technology', are the following tasks: Development of in-house ACN (Altitude Compensating Nozzle) aerodynamic design capability; Building in-house experience for all aspects of ACN via End-to-End Nozzle Test Program; Obtaining Experimental Data for Annular Aerospike: Thrust eta, TVC (thrust vector control) capability and surface pressures. To support selection/optimization of future Launch Vehicle propulsion we needed a parametric design and performance tool for ACN. We chose to start with the ACN Aerospike Nozzles.

  11. Variable volume combustor with pre-nozzle fuel injection system

    Science.gov (United States)

    Keener, Christopher Paul; Johnson, Thomas Edward; McConnaughhay, Johnie Franklin; Ostebee, Heath Michael

    2016-09-06

    The present application provides a combustor for use with a gas turbine engine. The combustor may include a number of fuel nozzles, a pre-nozzle fuel injection system supporting the fuel nozzles, and a linear actuator to maneuver the fuel nozzles and the pre-nozzle fuel injection system.

  12. Optimized design of a hypersonic nozzle

    Science.gov (United States)

    Krishnamurthy, Ramesh

    1994-01-01

    Conventional procedures for designing nozzles involve the design of an inviscid contour (using the method of characteristics) that is corrected with a displacement thickness calculated from boundary-layer theory. However, nozzles designed using this classical procedure have been shown to exhibit poor flow quality at Mach numbers characteristic of hypersonic applications. The nozzle to be designed will be a part of the NASA HYPULSE facility which is being used for hypervelocity flight research. Thus, the flow quality of the nozzle is a critical question that needs to be addressed. Design of nozzles for hypersonic applications requires a proper assessment of the effects of the thick boundary layer on the inviscid flowfield. Since the flow field is largely supersonic, the parabolized form of the Navier-Stokes (PNS) equations can be used. The requirement of a uniform flow at the exit plane of the nozzle can be used to define an objective function as part of an optimization procedure. The design procedure used in this study involves the coupling of a nonlinear (least-squares) optimization algorithm with an efficient, explicit PNS solver. The thick boundary layers growing on the walls of the nozzle limit the extent of the usable core region (region with uniform flow) for testing models (especially rectangular). In order to maximize the region of uniform flow, it was decided to have the exit plane of this nozzle to be (nearly) rectangular. Thus, an additional constraint on the nozzle shape resulted, namely the nozzle will have a shape transitioning from a circular one at the inlet to that of a rectangle at the exit. In order to provide for a smooth shape transition, the cross sectional contour of the nozzle is defined by a superellipse. The nozzle is taken to be a meter in length. The axial variations of the major and minor radii of the superellipse are governed by cubic splines. The design parameters are the coefficients of the splines associated with the local nozzle

  13. Fractal analysis of agricultural nozzles spray

    Directory of Open Access Journals (Sweden)

    Francisco Agüera

    2012-02-01

    Full Text Available Fractal scaling of the exponential type is used to establish the cumulative volume (V distribution applied through agricultural spray nozzles in size x droplets, smaller than the characteristic size X. From exponent d, we deduced the fractal dimension (Df which measures the degree of irregularity of the medium. This property is known as 'self-similarity'. Assuming that the droplet set from a spray nozzle is self-similar, the objectives of this study were to develop a methodology for calculating a Df factor associated with a given nozzle and to determine regression coefficients in order to predict droplet spectra factors from a nozzle, taking into account its own Df and pressure operating. Based on the iterated function system, we developed an algorithm to relate nozzle types to a particular value of Df. Four nozzles and five operating pressure droplet size characteristics were measured using a Phase Doppler Particle Analyser (PDPA. The data input consisted of droplet size spectra factors derived from these measurements. Estimated Df values showed dependence on nozzle type and independence of operating pressure. We developed an exponential model based on the Df to enable us to predict droplet size spectra factors. Significant coefficients of determination were found for the fitted model. This model could prove useful as a means of comparing the behavior of nozzles which only differ in not measurable geometric parameters and it can predict droplet spectra factors of a nozzle operating under different pressures from data measured only in extreme work pressures.

  14. Influence of nozzle-exit boundary-layer conditions on the flow and acoustic fields of initially laminar jets

    OpenAIRE

    Bogey , Christophe; Bailly , Christophe

    2010-01-01

    International audience; Round jets originating from a pipe nozzle are computed by large-eddy simulations (LES) to investigate the effects of the nozzle-exit conditions on the flow and sound fields of initially laminar jets. The jets are at Mach number 0.9 and Reynolds number 105, and exhibit exit boundary layers characterized by Blasius velocity profiles, maximum root-mean-square (r.m.s.) axial velocity fluctuations between 0.2 and 1.9% of the jet velocity, and momentum thicknesses varying fr...

  15. Pipe-to-pipe impact program

    International Nuclear Information System (INIS)

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

    1984-06-01

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

  16. Limit the effects of secondary circuit water or steam piping breaks in the reactor building

    International Nuclear Information System (INIS)

    Nachev, N.

    2001-01-01

    The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

  17. Axisymmetric nozzles with chamfered contraction

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav

    2017-01-01

    Roč. 263, August (2017), s. 147-158 ISSN 0924-4247 Institutional support: RVO:61388998 Keywords : nozzles * chamfering * invariant Subject RIV: BK - Fluid Dynamics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 2.499, year: 2016 http://ac.els-cdn.com/S0924424716310329/1-s2.0-S0924424716310329-main.pdf?_tid=f953dc4c-873c-11e7-b8d0-00000aacb35d&acdnat=1503408341_51527a384c272a3c4e8f43e6046d789d

  18. Method and apparatus for setting precise nozzle/belt and nozzle/edge dam block gaps

    Science.gov (United States)

    Carmichael, Robert J.; Dykes, Charles D.; Woodrow, Ronald

    1989-05-16

    A pair of guide pins are mounted on sideplate extensions of the caster and mating roller pairs are mounted on the nozzle assembly. The nozzle is advanced toward the caster so that the roller pairs engage the guide pins. Both guide pins are remotely adjustable in the vertical direction by hydraulic cylinders acting through eccentrics. This moves the nozzle vertically. The guide pin on the inboard side of the caster is similarly horizontally adjustable. The nozzle roller pair which engage the inboard guide pin are flanged so that the nozzle moves horizontally with the inboard guide pin.

  19. High mass throughput particle generation using multiple nozzle spraying

    Science.gov (United States)

    Pui, David Y.H.; Chen, Da-Ren

    2004-07-20

    Spraying apparatus and methods that employ multiple nozzle structures for producing multiple sprays of particles, e.g., nanoparticles, for various applications, e.g., pharmaceuticals, are provided. For example, an electrospray dispensing device may include a plurality of nozzle structures, wherein each nozzle structure is separated from adjacent nozzle structures by an internozzle distance. Sprays of particles are established from the nozzle structures by creating a nonuniform electrical field between the nozzle structures and an electrode electrically isolated therefrom.

  20. A comparative simulation of feed and bleed operation during the total loss of feedwater event by RELAP5/MOD3 and CEFLASH-4AS/REM computer codes

    International Nuclear Information System (INIS)

    Kwon, Y.M.; Ro, T.S.; Song, J.H.

    1995-01-01

    The Ulchin 3 and 4 nuclear power plants, which are two-loop 2,825 MW(thermal) pressurized water reactors designed by the Korea Atomic Energy Research Institute, adopted a safety depressurization system (SDS) to mitigate the beyond-design-basis event of a total loss of feedwater (TLOFW). A comparative simulation by the CEFLASH-4AS/REM and RELAP5/MOD3 computer codes for the TLOFW event without operator recovery and the TLOFW event with feed and bleed (F and B) operation is performed for Ulchin 3 and 4. In the analyses, the SDS bleed paths are modeled by orifices located on the top of the pressurizer, where the analytical area of the bleed path is based on the Ulchin 3 and 4 SDS design flow capacity. An additional case, where the SDS piping and valves are modeled explicitly, is considered for the RELAP5 analysis. The predictions by the CEFLASH-4AS/REM of the transient two-phase system behavior show good qualitative and quantitative agreement with those by the RELAP5 simulation. The RELAP5 case with explicit piping results in less repressurization and lower reactor coolant system pressure than that of the case without explicit SDS modeling. However, the two cases of RELAP5 analyses result in essentially the same transient scenarios for TLOFW with F and B operation. The results of the simulation demonstrate the validity of the Ulchin 3 and 4 design approach, which employs CEFLASH-4AS/REM computer code and SDS bleed paths modeled by orifices located on the top of the pressurizer. The results also indicate that the decay heat removal and core inventory makeup function can be successfully accomplished by F and B operation by using the SDS for Ulchin 3 and 4

  1. Erosion-Resistant Water-Blast Nozzle

    Science.gov (United States)

    Roberts, Marion L.; Rice, R. M.; Cosby, S. A.

    1988-01-01

    Design of nozzle reduces erosion of orifice by turbulent high-pressure water flowing through it. Improved performance and resistance to erosion achieved by giving interior nozzle surface long, gradual convergence before exit orifice abrupt divergence after orifice and by machining surface to smooth finish.

  2. Hydrothermal carbonization (HTC) of wheat straw: influence of feedwater pH prepared by acetic acid and potassium hydroxide.

    Science.gov (United States)

    Reza, M Toufiq; Rottler, Erwin; Herklotz, Laureen; Wirth, Benjamin

    2015-04-01

    In this study, influence of feedwater pH (2-12) was studied for hydrothermal carbonization (HTC) of wheat straw at 200 and 260°C. Acetic acid and KOH were used as acidic and basic medium, respectively. Hydrochars were characterized by elemental and fiber analyses, SEM, surface area, pore volume and size, and ATR-FTIR, while HTC process liquids were analyzed by HPLC and GC. Both hydrochar and HTC process liquid qualities vary with feedwater pH. At acidic pH, cellulose and elemental carbon increase in hydrochar, while hemicellulose and pseudo-lignin decrease. Hydrochars produced at pH 2 feedwater has 2.7 times larger surface area than that produced at pH 12. It also has the largest pore volume (1.1 × 10(-1) ml g(-1)) and pore size (20.2 nm). Organic acids were increasing, while sugars were decreasing in case of basic feedwater, however, phenolic compounds were present only at 260°C and their concentrations were increasing in basic feedwater. Copyright © 2015 Elsevier Ltd. All rights reserved.

  3. Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G., E-mail: wrcc@cdtn.br, E-mail: egr@cdtn.br, E-mail: tanius@cdtn.br, E-mail: dhbs@cdtn.br, E-mail: tanius@cdtn.br, E-mail: raphaelmecanica@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Souto, Joao P.R.S.; Carvalho Junior, Ideir T., E-mail: joprocha@yahoo.com.br, E-mail: ideir_engenharia@yahoo.com.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Metalurgica

    2013-07-01

    Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

  4. Residual stress measurements in the dissimilar metal weld in pressurizer safety nozzle of nuclear power plant

    International Nuclear Information System (INIS)

    Campos, Wagner R.C.; Rabello, Emerson G.; Mansur, Tanius R.; Scaldaferri, Denis H.B.; Paula, Raphael G.; Souto, Joao P.R.S.; Carvalho Junior, Ideir T.

    2013-01-01

    Weld residual stresses have a large influence on the behavior of cracking that could possibly occur under normal operation of components. In case of an unfavorable environment, both stainless steel and nickel-based weld materials can be susceptible to stress-corrosion cracking (SCC). Stress corrosion cracks were found in dissimilar metal welds of some pressurized water reactor (PWR) nuclear plants. In the nuclear reactor primary circuit the presence of tensile residual stress and corrosive environment leads to so-called Primary Water Stress Corrosion Cracking (PWSCC). The PWSCC is a major safety concern in the nuclear power industry worldwide. PWSCC usually occurs on the inner surface of weld regions which come into contact with pressurized high temperature water coolant. However, it is very difficult to measure the residual stress on the inner surfaces of pipes or nozzles because of inaccessibility. A mock-up of weld parts of a pressurizer safety nozzle was fabricated. The mock-up was composed of three parts: an ASTM A508 C13 nozzle, an ASTM A276 F316L stainless steel safe-end, an AISI 316L stainless steel pipe and different filler metals of nickel alloy 82/182 and AISI 316L. This work presents the results of measurements of residual strain from the outer surface of the mock-up welded in base metals and filler metals by hole-drilling strain-gage method of stress relaxation. (author)

  5. Palo Verde Unit 3 BMI nozzle modification

    International Nuclear Information System (INIS)

    Waskey, D.

    2015-01-01

    The 61 BMI (Bottom Mount Instrumentation) nozzles of the unit 3 of the Palo Verde plant have been examined through ASME Code Case N722. The nozzle 3 was the only one with leakage noted. The ultrasound testing results are characteristic of PWSCC (Primary Water Stress Corrosion Cracking). The initiation likely occurred at a weld defect which was exposed to the primary water environment resulting in PWSCC. All other nozzles (60) showed no unacceptable indications. Concerning nozzle 3 one crack in J-groove weld connected large defect to primary water. An environmental model has been used to simulate and optimize the repair. The AREVA crew was on site 18 days after contract award and the job was completed in 12 days, 30 hours ahead of baseline schedule. This series of slides describes the examination of the BMI nozzles, the repair steps, and alternative design concepts

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

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  7. External Cylindrical Nozzle with Controlled Vacuum

    Directory of Open Access Journals (Sweden)

    V. N. Pil'gunov

    2015-01-01

    Full Text Available There is a developed design of the external cylindrical nozzle with a vacuum camera. The paper studies the nozzle controllability of flow rate via regulated connection of the evacuated chamber to the atmosphere through an air throttle. Working capacity of the nozzle with inlet round or triangular orifice are researched. The gap is provided in the nozzle design between the external wall of the inlet orifice and the end face of the straight case in the nozzle case. The presented mathematical model of the nozzle with the evacuated chamber allows us to estimate the expected vacuum amount in the compressed section of a stream and maximum permissible absolute pressure at the inlet orifice. The paper gives experimental characteristics of the fluid flow process through the nozzle for different values of internal diameter of a straight case and an extent of its end face remoteness from an external wall of the inlet orifice. It estimates how geometry of nozzle constructive elements influences on the volume flow rate. It is established that the nozzle capacity significantly depends on the shape of inlet orifice. Triangular orifice nozzles steadily work in the mode of completely filled flow area of the straight case at much more amounts of the limit pressure of the flow. Vacuum depth in the evacuated chamber also depends on the shape of inlet orifice: the greatest vacuum is reached in a nozzle with the triangular orifice which 1.5 times exceeds the greatest vacuum with the round orifice. Possibility to control nozzle capacity through the regulated connection of the evacuated chamber to the atmosphere was experimentally estimated, thus depth of flow rate regulation of the nozzle with a triangular orifice was 45% in comparison with 10% regulation depth of the nozzle with a round orifice. Depth of regulation calculated by a mathematical model appeared to be much more. The paper presents experimental dependences of the flow coefficients of nozzle input orifice

  8. Failure behaviour of a piping system with a circumferentially orientated flaw

    International Nuclear Information System (INIS)

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

    1987-01-01

    The experiments were conducted on the recently installed feedwater line of the HDR reactor in Kahl. The investigations were focused on analysing both the crack propagation of a circumferentially flowed pipe under the influence of corrosion and cyclic load, together with the pipeline's subsequent failure behaviour. The experimental conditions were selected in a manner representing those which can, for example, prevait during start-up or shut-down of reactor. To this aim, the pipes were internally stressed with high pressure and temperature oxygenic water in conjunction with an externally applied bending moment. The investigations are supplemented by elastic-plastic triaxial finite element (FE) calculations for various assumed crack configurations, both prior to and following the experiments, thus granting a fracture-mechanical assessment of the structural behaviour. (orig./DG) [de

  9. Implementation of an advanced digital feedwater control system at the Prairie Island nuclear generating station

    International Nuclear Information System (INIS)

    Paris, R.E.; Gaydos, K.A.; Hill, J.O.; Whitson, S.G.; Wirkkala, R.

    1990-05-01

    EPRI Project RP2126-4 was a cooperative effort between TVA, EPRI, and Westinghouse which resulted in the demonstration of a prototype of a full range, fully automatic feedwater control system, using fault tolerant digital technology, at the TVA Sequoyah simulator site. That prototype system also included advanced signal validation algorithms and an advanced man-machine interface that used CRT-based soft-control technology. The Westinghouse Advanced Digital Feedwater Control System (ADFCS) upgrade, which contains elements that were part of that prototype system, has since been installed at Northern States Power's Prairie Island Unit 2. This upgrade was very successful due to the use of an advanced control system design and the execution of a well coordinated joint effort between the utility and the supplier. The project experience is documented in this report to help utilities evaluate the technical implications of such a project. The design basis of the Prairie Island ADFCS signal validation for input signal failure fault tolerance is outlined first. Features of the industry-proven system control algorithms are then described. Pre-shipment hardware-in-loop and factory acceptance testing of the Prairie Island system are summarized. Post-shipment site testing, including preoperational and plant startup testing, is also summarized. Plant data from the initial system startup is included. The installation of the Prairie Island ADFCS is described, including both the feedwater control instrumentation and the control board interface. Modification of the plant simulator and operator and I ampersand C personnel training are also discussed. 6 refs., 14 figs., 3 tabs

  10. Regulatory analysis for the resolution of Generic Issue 125.II.7 ''Reevaluate Provision to Automatically Isolate Feedwater from Steam Generator During a Line Break''

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1988-09-01

    Generic Issue 125.II.7 addresses the concern related to the automatic isolation of auxiliary feedwater (AFW) to a steam generator with a broken steam or feedwater line. This regulatory analysis provides a quantitative assessment of the costs and benefits associated with the removal of the AFW automatic isolation and concludes that no new regulatory requirements are warranted. 21 refs., 7 tabs

  11. A reliability centered maintenance model applied to the auxiliary feedwater system of a nuclear power plant

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges

    1998-01-01

    The main objective of maintenance in a nuclear power plant is to assure that structures, systems and components will perform their design functions with reliability and availability in order to obtain a safety and economic electric power generation. Reliability Centered Maintenance (RCM) is a method of systematic review to develop or optimize Preventive Maintenance Programs. This study presents the objectives, concepts, organization and methods used in the development of RCM application to nuclear power plants. Some examples of this application are included, considering the Auxiliary Feedwater System of a generic two loops PWR nuclear power plant of Westinghouse design. (author)

  12. Feed-water heaters alternative design comparison; Comparacion de disenos alternativos de calentadores

    Energy Technology Data Exchange (ETDEWEB)

    Torres Toledano, Gerardo [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    A procedure is presented for the alternative design comparison of feed water heaters, based in the failure records of damaged tubes during operation. The procedure is used for cases in which non-continuous or random inspections are made to the feed-water heaters. [Espanol] Se presenta un procedimiento para comparar disenos alternativos de calentadores, basandose en los registros de fallas de los tubos rotos acumuladas durante su operacion. El procedimiento se emplea para casos en los que se realizan inspecciones a los calentadores no continuas, ya sea periodicas o al azar.

  13. Solids filtration of high-temperature feedwater in a PWR secondary circuit: Final report

    International Nuclear Information System (INIS)

    Siegwarth, D.P.; Friedman, K.A.; Chakravorti, R.K.; Alibutod, L.J.

    1988-11-01

    Pressurized water reactor steam generators and turbines have experienced a variety of corrosion problems as a result of ionic, corrosion product and oxidizing species transport to the steam generators. Installation of high temperature filters on final feedwater, high pressure drains and moisture separator drains to reduce corrosion product ingress to the steam generators of a 1160 MWe design basis plant are specified and evaluated. Cost estimates for installing electromagnetic filters, and added operating and maintenance costs are given. 18 refs., 12 figs., 9 tabs

  14. Aging and service wear of auxiliary feedwater pumps for PWR nuclear power plants

    International Nuclear Information System (INIS)

    Greenstreet, W.L.

    1989-01-01

    This paper describes investigations on auxiliary feedwater pumps being done under the Nuclear Plant Aging Research (NPAR) Program. Objectives of these studies are: to identify and evaluate practical, cost-effective methods for detecting, monitoring, and assessing the severity of time-dependent degradation (aging and service wear); recommend inspection and maintenance practices; establish acceptance criteria; and help facilitate use of the results. Emphasis is given to identifying and assessing methods for detecting failure in the incipient stage and to developing degradation trends to allow timely maintenance, repair or replacement actions. 3 refs

  15. Application of a Long Term Asset Management Strategy for HP Feedwater Heaters

    International Nuclear Information System (INIS)

    Won, Se Youl; Yun, Eun Sub; Park, Young Sheop

    2008-01-01

    As the commercial operating year of nuclear power plants is increased, it becomes imperative to develop integrated cost-effective asset management and to improve plans for degraded Structures, Systems, and Components (SSCs) in terms of safety and economical consideration. A long-term asset management (LTAM) strategy can improve the condition of nuclear plants, maximize their value, and optimize their operational life by maintaining their safety. This paper presents an optimized LTAM plan for HP feedwater heaters at a specific nuclear power plant

  16. Boiler feedwater quality improvement by replacing conventional pre-treatment with advanced membrane systems

    Energy Technology Data Exchange (ETDEWEB)

    Doll, Bernhard [Process Systems Pall GmbH, Dreieich (Germany). Marketing; Venkatadri, Ramraj [Pall Corporation, Port Washington, NY (United States). Global Marketing Energy

    2013-09-01

    Two case studies in different application fields highlight significant economical and operational improvements that were achieved by replacing conventional water treatment technologies by highly-sophisticated membrane systems. The first case study deals with boiler feedwater in a power plant, focusing on the challenges faced as well as the direct and indirect benefits gained by the new system within a utility station. The second case study deals with the conventional water treatment scheme for groundwater from 13 wells at a major oil sands facility. Operational performance as well as the cost improvements gained in both cases will be presented. (orig.)

  17. Power-feedwater enthalpy operating domain for SBWR applying Monte Carlo simulation

    International Nuclear Information System (INIS)

    Quezada-Garcia, S.; Espinosa-Martinez, E.-G.; Vazquez-Rodriguez, A.; Varela-Ham, J.R.; Espinosa-Paredes, G.

    2014-01-01

    In this work the analyses of the feedwater enthalpy effects on reactor power in a simplified boiling water reactor (SBWR) applying a methodology based on Monte Carlo's simulation (MCS), is presented. The MCS methodology was applied systematically to establish operating domain, due that the SBWR are not yet in operation, the analysis of the nuclear and thermalhydraulic processes must rely on numerical modeling, with the purpose of developing or confirming the design basis and qualifying the existing or new computer codes to enable reliable analyses. (author)

  18. Life cycle management, design review, and condition assessment of feedwater heaters

    International Nuclear Information System (INIS)

    Gammage, D.; Idvorian, N.

    2012-01-01

    OPEX from both the Nuclear and Fossil Power Generation Industries shows that Feedwater Heaters (FWHs) are subject to several degradation mechanisms and that this degradation commonly leads to replacement of these vessels in order to ensure reliable, efficient operation of the plants. Loss of feedwater heating will impact plant thermal performance. In response to inspection results showing on-going degradation as well as other factors, B&W Canada completed a project in conjunction with a US PWR utility to review the design, condition, and Life Cycle Management of their FWHs. This project involved a multi-disciplinary approach in order to consider all aspects of the FWHs in order to provide insight into the Life Cycle Management Plan (LCMP) so that the FWHs can be operated reliably into the future and so that adequate inspections can be conducted in order to produce a detailed condition assessment. The utility was interested in evaluating their FWH LCMP to determine if it was adequate in its requirements to enable reliable, leak-free operation of their FWH equipment. As inputs to this evaluation, it was required that B&W Canada evaluate both confirmed and plausible degradation mechanisms. They also required that the thermal hydraulic and functional design be evaluated for their particular FWHs. It was important to also incorporate industry OPEX in order to provide proper trending information for tube plugging. Out of this evaluation there were several findings and recommendations that could be used to update the utilities’ LCMP as it was apparent that the current version may not be truly reflective of the current condition of the equipment or of current industry OPEX of such FWHs. Several recommendations came from this evaluation, the most significant were: • Performing thermal/hydraulic, FIV (flow-induced vibration), and tube/shell interaction calculations to determine how the FWHs operate and how their performance can change over time as a function of tube

  19. Remote visual testing (RVT) for the diagnostic inspection of feedwater heaters

    International Nuclear Information System (INIS)

    Nugent, M.J.; Pellegrino, B.A.

    1991-01-01

    In this paper the benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters will be briefly reviewed. All Remote Visual Testing (RVT) devices including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras will be discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing will be discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections will be discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, will be detailed

  20. Probabilistic analysis of reactor safety - The auxiliary feedwater system of Angra I

    International Nuclear Information System (INIS)

    Oliveira, L.C.R. da L.C. de.

    1981-09-01

    The unavailability of the auxiliary feedwater system (AFWS) of Angra-1, was calculated. The fault tree analysis technique was used, considering two diferent types of contribution to system unavailability: The one due to hard-ware failure and the contribution due to test and maintenance which was separately analysed. The COMBO-and SAMPLE computer codes were used. The results have shown that the AFWS of Angra-1 contains enough redundancy to guarantee a safe operation under the conditions analysed, best values having been obtained for the unavailability of AFWS of Angra 1 with those codes than with the WASH-1400. (E.G.) [pt

  1. Elastic-plastic response of a piping system due to simulated double-ended guillotine break events

    International Nuclear Information System (INIS)

    Kussmaul, K.; Diem, H.; Hunger, H.; Katzenmeier, G.

    1987-01-01

    From the blowdown experiments performed on the HDR feedwater line with feedwater check valve the conclusion can be drawn that high transient loads of up to plastic strains of 3%, acting on an initially integer piping system, can be sustained without loss of integrity for a low number of load cycles due to the plasticizing capacity of the pipework materials nowadays used in the reactor technology. In the experiments carried out with ferritic piping of ND 400 pressure peaks up to about 31,5 mPA were achieved which resulted in excessive strains of up to 3%. By nonlinear finite element computations (ABAQUS) it was possible to describe the elastic-plastic behaviour of the piping in a good approximation. On account of the safety margins proved in the experiments, potential inaccuracies in theoretical structure analyses are recommended so as to be on the safe side. On the other hand, it appears that designing pipework with reference to elastic stress categories does not adequately take into account the actual reserves of the pipework material

  2. Uranium enrichment by the separation nozzle process

    International Nuclear Information System (INIS)

    Becker, E.W.; Bier, W.; Ehrfeld, W.; Schubert, K.; Schuette, R.; Seidel, D.

    1975-11-01

    The separation nozzle process for the enrichment of the light uranium isotope U-235 has been developed at the Karlsruhe Nuclear Research Center as an alternative to the gaseous diffusion and centrifuge processes. Since 1970 the STEAG company, Essen, has been involved in the commercial implementation of the nozzle process. A first separation nozzle process. A first separation nozzle demonstration plant with a separative capacity of 180 t SWU/a shall be erected in Brazil with the participation of the Brazilian company NUCLEBRAS and the German companies STEAG and INTERATOM. Methods for the mass production of separation elements were developed by industry and extensive performance tests were carried out on commercially fabricated separation elements. Two prototype separative stages were successfully tested in Karlsruhe. Besides further plant components, a prototype of a UF 6 recycle facility was developed which serves the purpose of stripping the UF 6 from the light auxiliary gas to be recycled in a separation nozzle cascade. The performance level achieved to date characterizes the separation nozzle process as reliable and feasible economically. Therefore, the erection of a separation nozzle demonstration plant can be recognized as the implementation of an enrichment process which combines a reliable and comparatively simple technology with a high potential for further improvements. (orig.) [de

  3. Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

    Energy Technology Data Exchange (ETDEWEB)

    None

    1985-07-01

    On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program.

  4. Loss of main and auxiliary feedwater event at the Davis-Besse Plant on June 9, 1985

    International Nuclear Information System (INIS)

    1985-07-01

    On June 9, 1985, Toledo Edison Company's Davis-Besse Nuclear Power Plant, located in Ottawa County, Ohio, experienced a partial loss of feedwater while the plant was operating at 90% power. Following a reactor trip, a loss of all feedwater occurred. The event involved a number of equipment malfunctions and extensive operator actions, including operator actions outside the control room. Several operator errors also occurred during the event. This report documents the findings of an NRC Team sent to Davis-Besse by the NRC Executive Director for Operations in conformance with the staff-proposed Incident Investigation Program

  5. Integrated TRAC/MELPROG analysis of core damage from a severe feedwater transient in the Oconee-1 PWR

    International Nuclear Information System (INIS)

    Henninger, R.J.; Boyack, B.E.

    1986-01-01

    A postulated complete loss-of-feedwater event in the Oconee-1 pressurized water reactor has been analyzed. With an initial version of the lonked TRAC and MELPROG codes, we have modeled the loss-of-feedwater event from initiation to the time of complete disruption of the core, which was calculated to occur by 6800 s. The highest structure temperatures otuside the vessel are on the flow path from the vessel to the pressurizer relief valve. Temperatures in excess of 1200 K could result in failure and depressurization of the primary system before vessel failure

  6. Low frequency sound absorption of orifice plates, perforated plates and nozzles

    Science.gov (United States)

    Salikuddin, M.; Plumblee, H. E., Jr.

    1980-01-01

    Analyses of impulse time history data from acoustic transmission tests for conical nozzles attached to a pipe show internal reflections from the solid contraction and open area tend to cancel. To gain understanding of the opposing reflections, tests were conducted by replacing the conical nozzles with orifice plates. The primary variable was the open to solid area ratio. Internal reflection coefficient data reveal that, at an area ratio of 10-12%, the low frequency internal reflection is reduced from unity to about 0.2. Based on comparisons of far-field and internal data, acoustic energy is not conserved. Results are presented for complex reflection coefficient and far-field noise for a series of orifice and perforated plate configurations.

  7. Introduction to Heat Pipes

    Science.gov (United States)

    Ku, Jentung

    2015-01-01

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

  8. 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)

  9. Simulation of loss of feedwater transient of MASLWR test facility by MARS-KS code

    Energy Technology Data Exchange (ETDEWEB)

    Park, Juyeop [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    MASLWR test facility is a mock-up of a passive integral type reactor equipped with helical coil steam generator. Since SMART reactor which is being current developed domestically also adopts helical coil steam generator, KINS has joined this ICSP to evaluate performance of domestic regulatory audit thermal-hydraulic code (MARS-KS code) in various respects including wall-to-fluid heat transfer model modification implemented in the code by independent international experiment database. In the ICSP, two types of transient experiments have been focused and they are loss of feedwater transient with subsequent ADS operation and long term cooling (SP-2) and normal operating conditions at different power levels (SP-3). In the present study, KINS simulation results by the MARS-KS code (KS-002 version) for the SP-2 experiment are presented in detail and conclusions on MARS-KS code performance drawn through this simulation is described. Performance of the MARS-KS code is evaluated through the simulation of the loss of feedwater transient of the MASLWR test facility. Steady state run shows helical coil specific heat transfer models implemented in the code is reasonable. However, through the transient run, it is also found that three-dimensional effect within the HPC and axial conduction effect through the HTP are not well reproduced by the code.

  10. Evaluation method for two-phase flow and heat transfer in a feed-water heater

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Minato, Akihiko

    1993-01-01

    A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)

  11. Condensation heat transfer of a feed-water heater and improvement of its performance

    International Nuclear Information System (INIS)

    Takamori, Kazuhide; Murase, Michio; Baba, Yoshikazu; Aihara, Tsuyoshi

    1995-01-01

    In this study, a condensation heat transfer model, coupled with a three-dimensional two-phase flow analysis, was developed. In the heat transfer model, the liquid film flow rate on the heat transfer tubes was calculated by a mass balance equation and the liquid film thickness was calculated from the liquid film flow rate using Nusselt's laminar flow model and Fujii's equation for the steam velocity effect. The model was verified by condensation heat transfer experiments. In the experiments, 112 horizontal, staggered tubes with an outer diameter of 16mm and length of 0.55m were used. The calculated over-all heat transfer coefficients agreed with the data within ±5% under the inlet quality conditions of 13-100%. Based on a three-dimensional two-phase flow analysis, an improved feed-water heater with support plates, which have flow holes between the upper and lower tube bundles, was designed. The total heat exchange capacity of the improved feed-water heater increased about 6%. (author)

  12. Nuclear thermal rocket nozzle testing and evaluation program

    Science.gov (United States)

    Davidian, Kenneth O.; Kacynski, Kenneth J.

    1993-01-01

    Performance characteristics of the Nuclear Thermal Rocket can be enhanced through the use of unconventional nozzles as part of the propulsion system. The Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis is outlined and the advantages of a plug nozzle are described. A facility description, experimental designs and schematics are given. Results of pretest performance analyses show that high nozzle performance can be attained despite substantial nozzle length reduction through the use of plug nozzles as compared to a convergent-divergent nozzle. Pretest measurement uncertainty analyses indicate that specific impulse values are expected to be within + or - 1.17 pct.

  13. Evolution of carbon steel corrosion in feedwater conditions reproduce by the Fortrand loop

    International Nuclear Information System (INIS)

    Delaunay, Sophie; Bescond, Aurelien; Mansour, Carine; Bretelle, Jean-Luc

    2012-09-01

    Fouling and tubes support plate blockage of steam generators (SG) are major problems in the secondary circuit of pressurized water reactor (PWR) plants. Corrosion products (CP) responsible of these phenomena are mainly constituted of magnetite. Limit the amount of these CP, generated in the feedwater system and transported to SG, constitutes one way to limit fouling and blockage of SGs. This work requires the understanding of CP behaviour in the feedwater system conditions. A specific experimental circulating water loop, FORTRAND, was built at EDF to follow the formation, the transport and the deposition of iron oxides in representative conditions of the secondary circuit feedwater system. The test section operating at high temperature (up to 250 deg. C) is made in carbon steel and includes three removable segments while all the other parts of the loop are made in stainless steel. First results confirm the formation of iron oxides on carbon steel and stainless steel surface in the conditions of PWR secondary circuits. The surface characterizations show that magnetite is the corrosion product formed on carbon steel and stainless steel at 220 deg. C and goethite is formed at room temperature on stainless steel. The aim of the most recent tests performed in FORTRAND loop was to follow the evolution of corrosion in the feedwater conditions. Tests were performed in one-phase flow conditions at 150 L.h -1 with a linear velocity of 0.82 m/s at 220 deg. C in morpholine/ammonia/hydrazine medium, at pH 25C equal to 9.2. To conduct this study, a removable segment constituted by ten tubes was added to the loop. Several tests were performed to follow the deposit thickness, the iron lost in solution and the oxide morphology with time from two to nine hundred sixty hours. Chemical conditions were controlled and the reproducibility of the results was confirmed by the observation of three tubes at each test. SEM pictures present kinetics with three steps: after the first hours the

  14. Aerospike Nozzle for Rotating Detonation Engine Application

    Data.gov (United States)

    National Aeronautics and Space Administration — This proposal presents a graduate MS research thesis on improving the efficiency of rotating detonation engines by using aerospike nozzle technologies. A rotating...

  15. Integrated Composite Rocket Nozzle Extension Project

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop and demonstrate an Integrated Composite Rocket Nozzle Extension (ICRNE) for use in rocket thrust chambers. The ICRNE will utilize an...

  16. Integrated Composite Rocket Nozzle Extension, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — ORBITEC proposes to develop and demonstrate an Integrated Composite Rocket Nozzle Extension (ICRNE) for use in rocket thrust chambers. The ICRNE will utilize an...

  17. Self-Adjusting Choke For Nozzle

    Science.gov (United States)

    Morrison, Andrew D.

    1991-01-01

    Self-adjusting choke for nozzle enables issuing stream of liquid to remain coherent, despite fluctuations in flow, along greater distance than possible with same nozzle without choke. Flexible membrane with slanted orifices deforms according to upstream pressure in flowing liquid. Advantageous for firefighting, making it possible to direct more concentrated flow of water at flame or hotspot. Also used in mining and for transferring liquids.

  18. Stress intensities for nozzle cracks in reactor vessels. Reporting period, January 1, 1976--October 31, 1976

    International Nuclear Information System (INIS)

    Smith, C.W.; Jolles, M.; Peters, W.H.

    1976-11-01

    A series of six frozen stress photoelastic tests was conducted to investigate the distribution of stress intensity factor (SIF) along a crack which occurred at the juncture of a pipe (nozzle) with a cylindrical pressure vessel. Typical photoelastic fringe patterns are shown for slices which were taken mutually orthogonal to the flaw border and the flaw surface. A typical plot of normalized apparent SIF versus square root of normalized distance from the crack tip is presented. The variation in SIF along the flaw border is given for all six different crack geometries and also the variation of SIF with varying a/T is presented. 40 references

  19. The full structural weld overlay procedure of PZR nozzles for KORI unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jang Wook; Cho, Hong Seok; Cho, Ki Hyun; Choi, Sang Hoon [Korea Plant Service and Engineering Co., Daejeon (Korea, Republic of)

    2009-10-15

    The discovery of Primary Water Stress Corrosion Cracking (PWSCC) in Dissimilar Metal Weld (DMW) has led to the use of corrosion-resistant high-nickel welding alloys for repair and mitigation. PWSCC happens if susceptible material, tensile strength and corrosive environment are concurrently satisfied, so that weld overlay of pipe outside can prevent PWSCC due to put ID surface in compression. This paper will contribute to the understanding of field application and effect on Full Structural Weld Overlay (SWOL) of KORI-1 pressurizer nozzles.

  20. CT Scan of NASA Booster Nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Schneberk, D; Perry, R; Thompson, R

    2004-07-27

    We scanned a Booster Nozzle for NASA with our 9 meV LINAC, AmSi panel scanner. Three scans were performed using different filtering schemes and different positions of the nozzle. The results of the scan presented here are taken from the scan which provided the best contrast and lowest noise of the three. Our inspection data shows a number of indications of voids in the outer coating of rubber/carbon. The voids are mostly on the side of the nozzle, but a few small voids are present at the ends of the nozzle. We saw no large voids in the adhesive layer between the Aluminum and the inner layer of carbon. This 3D inspection data did show some variation in the size of the adhesive layer, but none of the indications were larger than 3 pixels in extent (21 mils). We have developed a variety of contour estimation and extraction techniques for inspecting small spaces between layers. These tools might work directly on un-sectioned nozzles since the circular contours will fit with our tools a little better. Consequently, it would be useful to scan a full nozzle to ensure there are no untoward degradations in data quality, and to see if our tools would work to extract the adhesive layer.

  1. Radiation buildup and control in BWR recirculation piping

    International Nuclear Information System (INIS)

    Meyer, W.; Wood, R.M.; Rao, T.V.; Vook, R.W.

    1987-01-01

    Boiling water nuclear reactors (BWRs) employ stainless steel (Types 304 or 316 NG) pipes in which high-purity water at temperatures of ∼ 275 0 C are circulated. Various components of the system, such as valves and bearings, often contain hard facing metal alloys such as Stellite-6. These components, along with the stainless steel tubing and feedwater, serve as sources of 59 Co. This cobalt, along with other soluble and insoluble impurities, is carried along with the circulating water to the reactor core where it is converted to radioactive 60 Co. After reentering the circulating water, the 60 Co can be incorporated into a complex corrosion layer in the form of CoCr 2 O 4 and/or CoFe 2 O 4 . The presence of even small amounts of 60 Co on the walls of BWR cooling systems is the dominant contributor to inplant radiation levels. Thus BWR owners and their agents are expending significant time and resources in efforts to reduce both the rate and amount of 60 Co buildup. The object of this research is twofold: (a) to form a thin diffusion barrier against the outward migration of cobalt from a cobalt-containing surface and (b) to prevent the growth of a 60 Co-containing corrosion film. The latter goal was the more important since most of the radioactive cobalt will originate from sources other than the stainless steel piping itself

  2. Piping research program plan

    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)

  3. Transients in pipes

    International Nuclear Information System (INIS)

    Marchesin, D.; Paes-Leme, P.J.S.; Sampaio, R.

    1981-01-01

    The motion of a fluid in a pipe is commonly modeled utilizing the one space dimension conservation laws of mass and momentum. The development of shocks and spikes utilizing the uniform sampling method is studied. The effects of temperature variations and friction are compared for gas pipes. (Author) [pt

  4. Modernization of the feedwater heaters control level of the Almaraz I Nuclear Power Plant by OVATION system

    International Nuclear Information System (INIS)

    Madronal Rodriguez, E.; Cabrero Munoz, J. E.

    2010-01-01

    As a result of the process of technological renovation of the heaters system and the power increase project, Almaraz Nuclear Power Plant has made several design changes in the feedwater heaters system. Within these changes, the old heaters control loops are replaced because the new power will increase the heaters drainage caudal. This modernization is carried out using the OVATION control system.

  5. Oscillating heat pipes

    CERN Document Server

    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...

  6. Flow Energy Piezoelectric Bimorph Nozzle Harvester

    Science.gov (United States)

    Sherrit, Stewart (Inventor); Walkemeyer, Phillip E. (Inventor); Hall, Jeffrey L. (Inventor); Lee, Hyeong Jae (Inventor); Colonius, Tim (Inventor); Tosi, Phillipe (Inventor); Kim, Namhyo (Inventor); Sun, Kai (Inventor); Corbett, Thomas Gary (Inventor); Arrazola, Alvaro Jose (Inventor)

    2016-01-01

    A flow energy harvesting device having a harvester pipe includes a flow inlet that receives flow from a primary pipe, a flow outlet that returns the flow into the primary pipe, and a flow diverter within the harvester pipe having an inlet section coupled to the flow inlet, a flow constriction section coupled to the inlet section and positioned at a midpoint of the harvester pipe and having a spline shape with a substantially reduced flow opening size at a constriction point along the spline shape, and an outlet section coupled to the constriction section. The harvester pipe may further include a piezoelectric structure extending from the inlet section through the constriction section and point such that the fluid flow past the constriction point results in oscillatory pressure amplitude inducing vibrations in the piezoelectric structure sufficient to cause a direct piezoelectric effect and to generate electrical power for harvesting.

  7. Computational study of performance characteristics for truncated conical aerospike nozzles

    Science.gov (United States)

    Nair, Prasanth P.; Suryan, Abhilash; Kim, Heuy Dong

    2017-12-01

    Aerospike nozzles are advanced rocket nozzles that can maintain its aerodynamic efficiency over a wide range of altitudes. It belongs to class of altitude compensating nozzles. A vehicle with an aerospike nozzle uses less fuel at low altitudes due to its altitude adaptability, where most missions have the greatest need for thrust. Aerospike nozzles are better suited to Single Stage to Orbit (SSTO) missions compared to conventional nozzles. In the current study, the flow through 20% and 40% aerospike nozzle is analyzed in detail using computational fluid dynamics technique. Steady state analysis with implicit formulation is carried out. Reynolds averaged Navier-Stokes equations are solved with the Spalart-Allmaras turbulence model. The results are compared with experimental results from previous work. The transition from open wake to closed wake happens in lower Nozzle Pressure Ratio for 20% as compared to 40% aerospike nozzle.

  8. Mobile polishing system of feedwater at start-up feedback from the implementation and future prospects

    International Nuclear Information System (INIS)

    Faure, Celine; Eade, Kevin; Fontan, Guillaume

    2012-09-01

    The reduction of the quantity of Steam Generator (SG) metallic oxides deposits, and maintaining a good chemical composition of the secondary side of SG tubes are some of the main objectives being looked at, in order to reduce the risk of SG corrosion, regardless of the alloy used, right from the start-up phase. For all types of outage, obtaining and maintaining sufficient chemical cleanliness at the start-up requires treatment of the water. The treatments are notably: - Water movements using the purge / make-up water method until the chemical criteria have been met. This method can be long and generate large volumes of discharge. - Using suitable resins to remove pollutants from the water. The advantage of this method is that it is selective. - Filtration, allowing for the removal of any insoluble agent. In order to optimise the start-up process, Gravelines and Blayais Nuclear Power Plants (NPPs) put trials in place towards the end of the 1980s. These trials lead to a water supply treatment installation (mobile polishing system- in French Systeme Mobile d'Epuration, SME) being put in place for the start-up phase, made up of an up-stream filter, a mixed-bed resin pollutant trap and a down-stream filter to prevent losing the fines into the feedwater. At the same time, the manifestation of cracking on the secondary side of the steam generator tubes lead EDF to roll out a water treatment for the feedwater dedicated to the start-up. The choice was made not to install a condensate polishing plant, in order to limit notably the pollution risks (resin leaks or waste from the regeneration in the backwater) following difficulties during regeneration. The positive results from the first trials validated for EDF the choice to give priority to the roll-out of the SME to the NPPs judged to be most critical due to the SG material. The SME, installed on a mobile base, can be used on different units at the same station; this reduced the investment and maintenance costs, and

  9. Considerations on the question of applying ion exchange or reverse osmosis methods in boiler feedwater processing

    International Nuclear Information System (INIS)

    Marquardt, K.; Dengler, H.

    1976-01-01

    This consideration is to show that the method of reverse osmosis presents in many cases an interesting and economical alternative to part and total desolination plants using ion exchangers. The essential advantages of the reverse osmosis are a higher degree of automization, no additional salting of the removed waste water, small constructional volume of the plant as well as favourable operational costs with increasing salt content of the crude water to be processed. As there is a relatively high salt breakthrough compared to the ion exchange method, the future tendency in boiler feedwater processing will be more towards a combination of methods of reverse osmosis and post-purification through continuous ion exchange methods. (orig./LH) [de

  10. Qualitative accident analysis on loss of normal feedwater for AP1000

    International Nuclear Information System (INIS)

    Li Yankai; Lin Meng; Hou Dong; Li Meilin; Yang Yanhua

    2012-01-01

    For the analysis of loss of normal feedwater accident for AP1000, a thermalhydraulic model was built based on a two-fluid best estimate code, RELAP5. A control system model was built based on Matlab/Simulink code. As the main data for modeling was from AP1000 Design Control Document (AP1000 DCD), which was inadequate and not accurate enough, the accident analysis was qualitative. The calculation results, overall consistent with the same accident in DCD, show that the RELAP5 code has the ability to calculate the natural circulation. PRHRS and CMT are able to remove the core residual heat, decay heat timely and efficiently during the accident to guarantee the safety of the core. In the calculation model, the capability of PRHRS is stronger, which makes the coolant temperature quickly reduced to a lower level, leading to an earlier actuation of CMT, followed by different responses of the plant. (authors)

  11. Pipe drafting and design

    CERN Document Server

    Parisher, Roy A

    2011-01-01

    Pipe Drafting and Design, Third Edition provides step-by-step instructions to walk pipe designers, drafters, and students through the creation of piping arrangement and isometric drawings. It includes instructions for the proper drawing of symbols for fittings, flanges, valves, and mechanical equipment. More than 350 illustrations and photographs provide examples and visual instructions. A unique feature is the systematic arrangement of drawings that begins with the layout of the structural foundations of a facility and continues through to the development of a 3-D model. Advanced chapters

  12. Threshold concentration of easily assimilable organic carton in feedwater for biofouling of spiral-wound membranes.

    Science.gov (United States)

    Hijnen, W A M; Biraud, D; Cornelissen, E R; van der Kooij, D

    2009-07-01

    One of the major impediments in the application of spiral-wound membranes in water treatment or desalination is clogging of the feed channel by biofouling which is induced by nutrients in the feedwater. Organic carbon is, under most conditions, limiting the microbial growth. The objective of this study is to assess the relationship between the concentration of an easily assimilable organic compound such as acetate in the feedwater and the pressure drop increase in the feed channel. For this purpose the membrane fouling simulator (MFS) was used as a model for the feed channel of a spiral-wound membrane. This MFS unit was supplied with drinking water enriched with acetate at concentrations ranging from 1 to 1000 microg C x L(-1). The pressure drop (PD) in the feed channel increased at all tested concentrations but not with the blank. The PD increase could be described by a first order process based on theoretical considerations concerning biofilm formation rate and porosity decline. The relationship between the first order fouling rate constant R(f) and the acetate concentration is described with a saturation function corresponding with the growth kinetics of bacteria. Under the applied conditions the maximum R(f) (0.555 d(-1)) was reached at 25 microg acetate-C x L(-1) and the half saturation constant k(f) was estimated at 15 microg acetate-C x L(-1). This value is higher than k(s) values for suspended bacteria grown on acetate, which is attributed to substrate limited growth conditions in the biofilm. The threshold concentration for biofouling of the feed channel is about 1 microg acetate-C x L(-1).

  13. Steady state flow evaluations for passive auxiliary feedwater system of APR

    International Nuclear Information System (INIS)

    Park, Jongha; Kim, Jaeyul; Seong, Hoje; Kang, Kyoungho

    2012-01-01

    This paper briefly introduces a methodology to evaluate steady state flow of APR+ Passive Auxiliary Feedwater System (PAFS). The PAFS is being developed as a safety grade passive system to completely replace the existing active Auxiliary Feedwater System (AFWS). Natural circulation cooling can be generally classified into the single-phase, two-phase, and boiling-condensation modes. The PAF is designed to be operated in a boiling-condensation natural circulation mode. The steady-state flow rate should be equal to the steady-state boiling/condensation rate determined by the steady-state energy and momentum balances in the PAFS. The determined steady-state flow rate can be used in the design optimization for the natural circulation loop of the PAFS through the steady-state momentum balance. Since the retarding force, which is to be balanced by the driving force in the natural circulation system design depends on the reliable evaluation of the success of a natural circulation system design depends on the reliable evaluation of the pressure loss coefficients. In PAFS, the core decay heat is released by natural circulation flow between the S G secondary side and the Passive Condensation Heat Exchanger (PCHX) that is immersed in the Passive Condensation Cooling Tank (PCCT). The PCCT is located on the top of Auxiliary building The driving force is determined by the difference between the S/G (heat Source) secondary water level and condensation liquid (heat sink) level. It will overcome retarding force at flowrate in the system, which is determined by vaporization and condensation of the steam which is generated at the S/G by the latent heat in system. In this study, the theoretical method to estimate the steady state flow rate in boiling-condensation natural circulation system is developed and compared with test results

  14. 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.)

  15. 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)

  16. Flexible ultrasonic pipe inspection apparatus

    Energy Technology Data Exchange (ETDEWEB)

    Jenkins, C.F.; Howard, B.D.

    1994-01-01

    Pipe crawlers, pipe inspection {open_quotes}rabbits{close_quotes} and similar vehicles are widely used for inspecting the interior surfaces of piping systems, storage tanks and process vessels for damaged or flawed structural features. This paper describes the design of a flexible, modular ultrasonic pipe inspection apparatus.

  17. Upper Stage Engine Composite Nozzle Extensions

    Science.gov (United States)

    Valentine, Peter G.; Allen, Lee R.; Gradl, Paul R.; Greene, Sandra E.; Sullivan, Brian J.; Weller, Leslie J.; Koenig, John R.; Cuneo, Jacques C.; Thompson, James; Brown, Aaron; hide

    2015-01-01

    Carbon-carbon (C-C) composite nozzle extensions are of interest for use on a variety of launch vehicle upper stage engines and in-space propulsion systems. The C-C nozzle extension technology and test capabilities being developed are intended to support National Aeronautics and Space Administration (NASA) and United States Air Force (USAF) requirements, as well as broader industry needs. Recent and on-going efforts at the Marshall Space Flight Center (MSFC) are aimed at both (a) further developing the technology and databases for nozzle extensions fabricated from specific CC materials, and (b) developing and demonstrating low-cost capabilities for testing composite nozzle extensions. At present, materials development work is concentrating on developing a database for lyocell-based C-C that can be used for upper stage engine nozzle extension design, modeling, and analysis efforts. Lyocell-based C-C behaves in a manner similar to rayon-based CC, but does not have the environmental issues associated with the use of rayon. Future work will also further investigate technology and database gaps and needs for more-established polyacrylonitrile- (PAN-) based C-C's. As a low-cost means of being able to rapidly test and screen nozzle extension materials and structures, MSFC has recently established and demonstrated a test rig at MSFC's Test Stand (TS) 115 for testing subscale nozzle extensions with 3.5-inch inside diameters at the attachment plane. Test durations of up to 120 seconds have been demonstrated using oxygen/hydrogen propellants. Other propellant combinations, including the use of hydrocarbon fuels, can be used if desired. Another test capability being developed will allow the testing of larger nozzle extensions (13.5- inch inside diameters at the attachment plane) in environments more similar to those of actual oxygen/hydrogen upper stage engines. Two C-C nozzle extensions (one lyocell-based, one PAN-based) have been fabricated for testing with the larger

  18. Aeroelastic Modeling of a Nozzle Startup Transient

    Science.gov (United States)

    Wang, Ten-See; Zhao, Xiang; Zhang, Sijun; Chen, Yen-Sen

    2014-01-01

    Lateral nozzle forces are known to cause severe structural damage to any new rocket engine in development during test. While three-dimensional, transient, turbulent, chemically reacting computational fluid dynamics methodology has been demonstrated to capture major side load physics with rigid nozzles, hot-fire tests often show nozzle structure deformation during major side load events, leading to structural damages if structural strengthening measures were not taken. The modeling picture is incomplete without the capability to address the two-way responses between the structure and fluid. The objective of this study is to develop a tightly coupled aeroelastic modeling algorithm by implementing the necessary structural dynamics component into an anchored computational fluid dynamics methodology. The computational fluid dynamics component is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, while the computational structural dynamics component is developed under the framework of modal analysis. Transient aeroelastic nozzle startup analyses at sea level were performed, and the computed transient nozzle fluid-structure interaction physics presented,

  19. Nozzle dam having a unitary plug

    Science.gov (United States)

    Veronesi, L.; Wepfer, R.M.

    1992-12-15

    Apparatus for sealing the primary-side coolant flow nozzles of a nuclear steam generator is disclosed. The steam generator has relatively small diameter manway openings for providing access to the interior of the steam generator including the inside surface of each nozzle, the manway openings having a diameter substantially less than the inside diameter of each nozzle. The apparatus includes a bracket having an outside surface for matingly sealingly engaging the inside surface of the nozzle. The bracket also has a plurality of openings longitudinally therethrough and a plurality of slots transversely therein in communication with each opening. A plurality of unitary plugs sized to pass through the manway opening are matingly sealingly disposed in each opening of the bracket for sealingly plugging each opening. Each plug includes a plurality of arms operable to engage the slots of the bracket for connecting each plug to the bracket, so that the nozzle is sealed as the plugs seal the openings and are connected to the bracket. 16 figs.

  20. Improved Thin, Flexible Heat Pipes

    Science.gov (United States)

    Rosenfeld, John H.; Gernert, Nelson J.; Sarraf, David B.; Wollen, Peter J.; Surina, Frank C.; Fale, John E.

    2004-01-01

    Flexible heat pipes of an improved type are fabricated as layers of different materials laminated together into vacuum- tight sheets or tapes. In comparison with prior flexible heat pipes, these flexible heat pipes are less susceptible to leakage. Other advantages of these flexible heat pipes, relative to prior flexible heat pipes, include high reliability and greater ease and lower cost of fabrication. Because these heat pipes are very thin, they are highly flexible. When coated on outside surfaces with adhesives, these flexible heat pipes can be applied, like common adhesive tapes, to the surfaces of heat sinks and objects to be cooled, even if those surfaces are curved.

  1. The pipes of pan.

    Science.gov (United States)

    Chalif, David J

    2004-12-01

    The pipes of pan is the crowning achievement of Pablo Picasso's neoclassical period of the 1920s. This monumental canvas depicts a mythological Mediterranean scene in which two sculpted classical giants stare out, seemingly across the centuries, toward a distant and lost Arcadia. Picasso was influenced by Greco-Roman art during his travels in Italy, and his neoclassical works typically portray massive, immobile, and pensive figures. Pan and his pipes are taken directly from Greek mythological lore by Picasso and placed directly into 20th century art. He frequently turned to various mythological figures throughout his metamorphosing periods. The Pipes of Pan was also influenced by the painter's infatuation with the beautiful American expatriate Sara Murphy, and the finished masterpiece represents a revision of a previously conceived neoclassical work. The Pipes of Pan now hangs in the Musee Picasso in Paris.

  2. Heat Pipe Systems

    Science.gov (United States)

    1993-01-01

    The heat pipe was developed to alternately cool and heat without using energy or any moving parts. It enables non-rotating spacecraft to maintain a constant temperature when the surface exposed to the Sun is excessively hot and the non Sun-facing side is very cold. Several organizations, such as Tropic-Kool Engineering Corporation, joined NASA in a subsequent program to refine and commercialize the technology. Heat pipes have been installed in fast food restaurants in areas where humid conditions cause materials to deteriorate quickly. Moisture removal was increased by 30 percent in a Clearwater, FL Burger King after heat pipes were installed. Relative humidity and power consumption were also reduced significantly. Similar results were recorded by Taco Bell, which now specifies heat pipe systems in new restaurants in the Southeast.

  3. Development of intelligent pipe locator

    Science.gov (United States)

    Miyamoto, Y.; Wasa, Y.

    1986-08-01

    An inductive pipe locator was developed, so that the position and depth of an underground gas pipe can be accurately located by passing an ac current to the pipe and measuring the generated magnetic field. An ac current (several to 100 kHz) of several tens of mA is transmitted to the underground pipe, and a magnetic sensor above the ground catches the induced magnetic field to estimate the position and depth of the pipe.

  4. Stuck pipe prediction

    KAUST Repository

    Alzahrani, Majed

    2016-03-10

    Disclosed are various embodiments for a prediction application to predict a stuck pipe. A linear regression model is generated from hook load readings at corresponding bit depths. A current hook load reading at a current bit depth is compared with a normal hook load reading from the linear regression model. A current hook load greater than a normal hook load for a given bit depth indicates the likelihood of a stuck pipe.

  5. Design of high pressure waterjet nozzles

    Science.gov (United States)

    Mazzoleni, Andre P.

    1994-10-01

    The Hydroblast Research Cell at Marshall Space Flight Center is used to investigate the use of high pressure waterjets to strip paint, grease, adhesive and thermal spray coatings from various substrates. Current methods of cleaning often use ozone depleting chemicals (ODC) such as chlorinated solvents. High pressure waterjet cleaning has proven to be a viable alternative to the use of solvents. A popular method of waterjet cleaning involves the use of a rotating, multijet, high pressure water nozzle which is robotically controlled. This method enables rapid cleaning of a large area, but problems such as incomplete coverage and damage to the substrate from the waterjet have been observed. This report summarizes research consisting of identifying and investigating the basic properties of rotating, multijet, high pressure water nozzles, and how particular designs and modes of operation affect such things as stripping rate, standoff distance and completeness of coverage. The study involved computer simulations, an extensive literature review, and experimental studies of different nozzle designs.

  6. Li/Li2 supersonic nozzle beam

    International Nuclear Information System (INIS)

    Wu, C.Y.R.; Crooks, J.B.; Yang, S.C.; Way, K.R.; Stwalley, W.C.

    1977-01-01

    The characterization of a lithium supersonic nozzle beam was made using spectroscopic techniques. It is found that at a stagnation pressure of 5.3 kPa (40 torr) and a nozzle throat diameter of 0.4 mm the ground state vibrational population of Li 2 can be described by a Boltzmann distribution with T/sub v/ = 195 +- 30 0 K. The rotational temperature is found to be T/sub r/ = 70 +- 20 0 K by band shape analysis. Measurements by quadrupole mass spectrometer indicates that approximately 10 mole per cent Li 2 dimers are formed at an oven body temperature of 1370 0 K n the supersonic nozzle expansion. This measured mole fraction is in good agreement with the existing dimerization theory

  7. Advanced Solid Rocket Motor nozzle development status

    Science.gov (United States)

    Kearney, W. J.; Moss, J. D.

    1993-01-01

    This paper presents a status update of the design and development of an improved nozzle for the Advanced Solid Rocket Motor (ASRM). The ASRM nozzle incorporates advanced state-of-the-art design features and materials which contribute to enhanced safety, reliability, performance, and producibility for the space shuttle boosters. During 1992 the nozzle design progressed through a successful Preliminary Design Review (PDR). An improved ablative material development program also culminated in the selection of new standard and low density carbon cloth phenolic prepreg offering reduced variability and improved process attributes. A subscale motor test series to evaluate new materials and design features was also completed. An overview update of the matured design characteristics, supporting analysis, key development-program results and program status and plans is reported.

  8. Biannular Airbreathing Nozzle Rig (BANR) facility checkout and plug nozzle performance test data

    Science.gov (United States)

    Cummings, Chase B.

    2010-09-01

    The motivation for development of a supersonic business jet (SSBJ) platform lies in its ability to create a paradigm shift in the speed and reach of commercial, private, and government travel. A full understanding of the performance capabilities of exhaust nozzle configurations intended for use in potential SSBJ propulsion systems is critical to the design of an aircraft of this type. Purdue University's newly operational Biannular Airbreathing Nozzle Rig (BANR) is a highly capable facility devoted to the testing of subscale nozzles of this type. The high accuracy, six-axis force measurement system and complementary mass flowrate measurement capabilities of the BANR facility make it rather ideally suited for exhaust nozzle performance appraisal. Detailed accounts pertaining to methods utilized in the proper checkout of these diagnostic capabilities are contained herein. Efforts to quantify uncertainties associated with critical BANR test measurements are recounted, as well. Results of a second hot-fire test campaign of a subscale Gulfstream Aerospace Corporation (GAC) axisymmetric, shrouded plug nozzle are presented. Determined test article performance parameters (nozzle thrust efficiencies and discharge coefficients) are compared to those of a previous test campaign and numerical simulations of the experimental set-up. Recently acquired data is compared to published findings pertaining to plug nozzle experiments of similar scale and operating range. Suggestions relating to the future advancement and improvement of the BANR facility are provided. Lessons learned with regards to test operations and calibration procedures are divulged in an attempt to aid future facility users, as well.

  9. Numerical evaluation of weld overlay applied to a pressurized water reactor nozzle mock-up

    Energy Technology Data Exchange (ETDEWEB)

    Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V., E-mail: egr@cdtn.b, E-mail: silvall@cdtn.b, E-mail: gomespt@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Integridade Estrutural

    2011-07-01

    The primary water stress corrosion cracking (PWSCC) is a major mechanism of failure in the primary circuit of PWR type nuclear power plants. The PWSCC is associated with the presence of corrosive environment, the susceptibility to corrosion cracking of the materials involved and the tensile stresses presence. Residual stresses generated during dissimilar materials welding can contribute to PWSCC. An alternative to the PWSCC mitigation is the application of external weld layers in the regions of greatest susceptibility to corrosion cracking. This process, called Weld Overlay (WOL), has been widely used in regions of dissimilar weld (low alloy steel and stainless steel with nickel alloy addition) of nozzles and pipes on the primary circuit in order to promote internal compressive stresses on the wall of these components. This paper presents the steps required to the numerical stress evaluation (by finite element method) during the dissimilar materials welding as well as application of Weld Overlay process in a nozzle mock-up. Thus, one can evaluate the effectiveness of the application of weld overlay process to internal compressive stress generation on the wall nozzle. (author)

  10. Structural integrity analyses for preemptive weld overlay on the dissimilar metal weld of a pressurizer nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Chin-Cheng, E-mail: cchuang@iner.gov.tw [Institute of Nuclear Energy Research, Taiwan (China); Liu, Ru-Feng [Institute of Nuclear Energy Research, Taiwan (China)

    2012-02-15

    This paper presents structural integrity analyses for preemptive weld overlay on the dissimilar metal weld (DMW) of a pressurizer nozzle in a pressurized water reactor (PWR). Based on MRP-169 and ASME Code Case N-504-2, weld overlay sizing calculation, residual stress improvement, shrinkage evaluation, fatigue crack growth and fatigue usage analysis are performed. The weld overlay procedure has to be confirmed to improve the residual stresses around the inside surface of DMW. The residual compressive stress distribution is thus addressed to be resistant to subsequent primary water stress corrosion cracking (PWSCC) initiation and further crack growth. To ensure the structural integrity of the original attached piping system, the measured displacement is transformed to temperature gradient to simulate the shrinkage after overlay and is used to determine the post weld distortion and stress situation. Further, the conservative postulated surface cracks are assumed in the DMW for fatigue crack growth analysis with system design cycles. The stress limits and cumulative fatigue usages of the pressurizer nozzle with overlay are also evaluated to meet ASME Code, Section III. Based on the present results, the structural integrity of the pressurizer nozzle with preemptive weld overlay is shown.

  11. Combustor nozzles in gas turbine engines

    Science.gov (United States)

    Johnson, Thomas Edward; Keener, Christopher Paul; Stewart, Jason Thurman; Ostebee, Heath Michael

    2017-09-12

    A micro-mixer nozzle for use in a combustor of a combustion turbine engine, the micro-mixer nozzle including: a fuel plenum defined by a shroud wall connecting a periphery of a forward tube sheet to a periphery of an aft tubesheet; a plurality of mixing tubes extending across the fuel plenum for mixing a supply of compressed air and fuel, each of the mixing tubes forming a passageway between an inlet formed through the forward tubesheet and an outlet formed through the aft tubesheet; and a wall mixing tube formed in the shroud wall.

  12. Lightweight Nozzle Extension for Liquid Rocket Engines Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The ARES J-2X requires a large nozzle extension. Currently, a metallic nozzle extension is being considered with carbon-carbon composite as a backup. In Phase 1,...

  13. Turbocharger with variable nozzle having vane sealing surfaces

    Science.gov (United States)

    Arnold, Philippe [Hennecourt, FR; Petitjean, Dominique [Julienrupt, FR; Ruquart, Anthony [Thaon les Vosges, FR; Dupont, Guillaume [Thaon les Vosges, FR; Jeckel, Denis [Thaon les Vosges, FR

    2011-11-15

    A variable nozzle for a turbocharger includes a plurality of vanes rotatably mounted on a nozzle ring and disposed in a nozzle flow path defined between the nozzle ring and an opposite nozzle wall. Either or both of the faces of the nozzle ring and nozzle wall include(s) at least one step that defines sealing surfaces positioned to be substantially abutted by airfoil surfaces of the vanes in the closed position of the vanes and to be spaced from the airfoil surfaces in positions other than the closed position. This substantial abutment between the airfoil surfaces and the sealing surfaces serves to substantially prevent exhaust gas from leaking past the ends of the airfoil portions. At the same time, clearances between the nozzle ring face and the end faces of the airfoil portions can be sufficiently large to prevent binding of the vanes under all operating conditions.

  14. Study on the Effect of Branching Outlet Pipe using CFD Method

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hark; Chae, Hee Taek; Kim, Heon Il; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    For a downward flow type research reactor immersed in a pool, coolant passes through the upper guide structure, cools down fuel assemblies in the core, and return to primary cooling loop through outlet nozzle of lower plenum. If the pressure drop of reactor core is too small, flow rates of fuel assemblies in the core may be affected by the flow pattern in the outlet plenum, which depends on the location and number of outlet nozzles. A uniform distribution of coolant flow in the reactor core is required to ensure a thermal-hydraulic safety. If outlet nozzle number could affect the flow distribution in the reactor core, it should be evaluated and fed into the design of reactor structure. This is a numerical study to figure out the effect of branching outlet pipe. CFD analyses for two types of outlet plenum are conducted and the predicted results of flow distribution and fluctuation are compared with each other

  15. Noise of Embedded High Aspect Ratio Nozzles

    Science.gov (United States)

    Bridges, James E.

    2011-01-01

    A family of high aspect ratio nozzles were designed to provide a parametric database of canonical embedded propulsion concepts. Nozzle throat geometries with aspect ratios of 2:1, 4:1, and 8:1 were chosen, all with convergent nozzle areas. The transition from the typical round duct to the rectangular nozzle was designed very carefully to produce a flow at the nozzle exit that was uniform and free from swirl. Once the basic rectangular nozzles were designed, external features common to embedded propulsion systems were added: extended lower lip (a.k.a. bevel, aft deck), differing sidewalls, and chevrons. For the latter detailed Reynolds-averaged Navier-Stokes (RANS) computational fluid dynamics (CFD) simulations were made to predict the thrust performance and to optimize parameters such as bevel length, and chevron penetration and azimuthal curvature. Seventeen of these nozzles were fabricated at a scale providing a 2.13 inch diameter equivalent area throat." ! The seventeen nozzles were tested for far-field noise and a few data were presented here on the effect of aspect ratio, bevel length, and chevron count and penetration. The sound field of the 2:1 aspect ratio rectangular jet was very nearly axisymmetric, but the 4:1 and 8:1 were not, the noise on their minor axes being louder than the major axes. Adding bevel length increased the noise of these nozzles, especially on their minor axes, both toward the long and short sides of the beveled nozzle. Chevrons were only added to the 2:1 rectangular jet. Adding 4 chevrons per wide side produced some decrease at aft angles, but increased the high frequency noise at right angles to the jet flow. This trend increased with increasing chevron penetration. Doubling the number of chevrons while maintaining their penetration decreased these effects. Empirical models of the parametric effect of these nozzles were constructed and quantify the trends stated above." Because it is the objective of the Supersonics Project that

  16. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code

    International Nuclear Information System (INIS)

    Hernandez M, J.L.; Ortiz V, J.

    2005-01-01

    The obtained results when carrying out the simulation of the fault transitory of the feedwater controller (FCAA) with the Ramona-3B code, happened in the Unit 2 of the Laguna Verde power plant (CNLV), in September of the year 2000 are presented. The transitory originates as consequence of the controller's fault of speed of a turbo pump of feedwater. The work includes a short description of the event, the suppositions considered for the simulation and the obtained results. Also, a discussion of the impact of the transitory event is presented on aspects of reactor safety. Although the carried out simulation is limited by the capacities of the code and for the lack of available information, it was found that even in a conservative situation, the power was incremented only in 12% above the nominal value, while that the thermal limit determined by the minimum reason of the critical power, MCPR, always stayed above the limit values of operation and safety. (Author)

  17. Removal of Iron Oxide Scale from Feed-water in Thermal Power Plant by Using Magnetic Separation

    Science.gov (United States)

    Nakanishi, Motohiro; Shibatani, Saori; Mishima, Fumihito; Akiyama, Yoko; Nishijima, Shigehiro

    2017-09-01

    One of the factors of deterioration in thermal power generation efficiency is adhesion of the scale to inner wall in feed-water system. Though thermal power plants have employed All Volatile Treatment (AVT) or Oxygen Treatment (OT) to prevent scale formation, these treatments cannot prevent it completely. In order to remove iron oxide scale, we proposed magnetic separation system using solenoidal superconducting magnet. Magnetic separation efficiency is influenced by component and morphology of scale which changes their property depending on the type of water treatment and temperature. In this study, we estimated component and morphology of iron oxide scale at each equipment in the feed-water system by analyzing simulated scale generated in the pressure vessel at 320 K to 550 K. Based on the results, we considered installation sites of the magnetic separation system.

  18. Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

    1993-12-01

    The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and, a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation. This transient actually occurred in the plant on June 19, 1989

  19. CFD Analysis On The Performance Of Wind Turbine With Nozzles

    Directory of Open Access Journals (Sweden)

    Chunkyraj Kh

    2015-08-01

    Full Text Available In this paper an effort has been made in dealing with fluid characteristic that enters a converging nozzle and analysis of the nozzle is carried out using Computational Fluid Dynamics package ANSYS WORKBENCH 14.5. The paper is the continuation of earlier work Analytical and Experimental performance evaluation of Wind turbine with Nozzles. First the CFD analysis will be carried out on nozzle in-front of wind turbine where streamline velocity at the exit volume flow rate in the nozzle and pressure distribution across the nozzle will be studied. Experiments were conducted on the Wind turbine with nozzles and the corresponding power output at different air speed and different size of nozzles were calculated. Different shapes and dimensions with special contours and profiles of nozzles were studied. It was observed that the special contour nozzles have superior outlet velocity and low pressure at nozzle exit the design has maximum Kinetic energy. These indicators conclude that the contraction designed with the new profile is a good enhancing of the nozzle performance.

  20. Design and Analysis of Elliptical Nozzle in AJM Process using ...

    African Journals Online (AJOL)

    The common nozzle shape presently used in AJM machining process is rectangle and circular shape nozzle which gives a low flow rate and further demands on decreasing the material removal rate (MRR), so this research mainly focuses on designing nozzle geometry to improve flow rate and MRR in AJM machining ...

  1. A fundamental study of a variable critical nozzle flow

    International Nuclear Information System (INIS)

    Kim, Jea Hyung; Kim, Heuy Dong; Park, Kyung Am

    2003-01-01

    The mass flow rate of gas flow through critical nozzle depends on the nozzle supply conditions and the cross-sectional area at the nozzle throat. In order that the critical nozzle can be operated at a wide range of supply conditions, the nozzle throat diameter should be controlled to change the flow passage area. This can be achieved by means of a variable critical nozzle. In the present study, both experimental and computational works are performed to develop variable critical nozzle. A cone-cylinder with a diameter of d is inserted into conventional critical nozzle. It can move both upstream and downstream, thereby changing the cross-sectional area of the nozzle throat. Computational work using the axisymmetric, compressible Navier-Stokes equations is carried out to simulate the variable critical nozzle flow. An experiment is performed to measure the mass flow rate through variable critical nozzle. The present computational results are in close agreement with measured ones. The boundary layer displacement and momentum thickness are given as a function of Reynolds number. An empirical equation is obtained to predict the discharge coefficient of variable critical nozzle

  2. Integrated Ceramic Matrix Composite and Carbon/Carbon Structures for Large Rocket Engine Nozzles and Nozzle Extensions Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Low-cost access to space demands durable, cost-effective, efficient, and low-weight propulsion systems. Key components include rocket engine nozzles and nozzle...

  3. Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

    1993-01-01

    The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

  4. Fabrication of Microglass Nozzle for Microdroplet Jetting

    Directory of Open Access Journals (Sweden)

    Dan Xie

    2015-02-01

    Full Text Available An ejection aperture nozzle is the essential part for all microdrop generation techniques. The diameter size, the flow channel geometry, and fluid impedance are the key factors affecting the ejection capacity. A novel low-cost fabrication method of microglass nozzle involving four steps is developed in this work. In the first heating step, the glass pipette is melted and pulled. Then, the second heating step is to determine the tip cone angle and modify the flow channel geometry. The desired included angle is usually of 30~45 degrees. Fine grind can determine the exact diameter of the hole. Postheating step is the final process and it can reduce the sharpness of the edges of the hole. Micronozzles with hole diameters varying from 30 to 100 µm are fabricated by the homemade inexpensive and easy-to-operate setup. Hydrophobic treating method of microglass nozzle to ensure stable and accurate injection is also introduced in this work. According to the jetting results of aqueous solution, UV curing adhesive, and solder, the fabricated microglass nozzle can satisfy the need of microdroplet jetting of multimaterials.

  5. Remotely installed steam generator nozzle dam system

    International Nuclear Information System (INIS)

    Mc Donald, F.X.; Weisel, E.M.; Schukei, G.E.

    1990-01-01

    This patent describes a method for remotely installing a dam unit in a nozzle or a nuclear steam generator head, the head including a manway. It comprises: mounting an articulated manipulator to an internal surface of the head, the manipulator having a free end which carries a jaw member; positioning the manipulator so that the jaw member is adjacent the manway and substantially on the manway axis; passing a first dam segment through the manway and attaching the jaw member to the first segment; positioning the manipulator so that the jaw member holds the first dam segment on one side of the manway axis; passing a second dam segment through the manway into engagement with the first dam segment to form a dam subassembly; translating the manipulator through the head until the dam subassembly is adjacent the nozzle; advancing the jaw member toward the nozzle until the cam subassembly is positioned substantially at the desired location of the dam unit with respect to the nozzle; and deploying the manipulator to install dam support structure between the dam subassembly and the steam generator, thereby forming an installed dam unit

  6. Clamp and Gas Nozzle for TIG Welding

    Science.gov (United States)

    Gue, G. B.; Goller, H. L.

    1982-01-01

    Tool that combines clamp with gas nozzle is aid to tungsten/inert-gas (TIG) welding in hard-to-reach spots. Tool holds work to be welded while directing a stream of argon gas at weld joint, providing an oxygen-free environment for tungsten-arc welding.

  7. Hydrogen/Air Fuel Nozzle Emissions Experiments

    Science.gov (United States)

    Smith, Timothy D.

    2001-01-01

    The use of hydrogen combustion for aircraft gas turbine engines provides significant opportunities to reduce harmful exhaust emissions. Hydrogen has many advantages (no CO2 production, high reaction rates, high heating value, and future availability), along with some disadvantages (high current cost of production and storage, high volume per BTU, and an unknown safety profile when in wide use). One of the primary reasons for switching to hydrogen is the elimination of CO2 emissions. Also, with hydrogen, design challenges such as fuel coking in the fuel nozzle and particulate emissions are no longer an issue. However, because it takes place at high temperatures, hydrogen-air combustion can still produce significant levels of NOx emissions. Much of the current research into conventional hydrocarbon-fueled aircraft gas turbine combustors is focused on NOx reduction methods. The Zero CO2 Emission Technology (ZCET) hydrogen combustion project will focus on meeting the Office of Aerospace Technology goal 2 within pillar one for Global Civil Aviation reducing the emissions of future aircraft by a factor of 3 within 10 years and by a factor of 5 within 25 years. Recent advances in hydrocarbon-based gas turbine combustion components have expanded the horizons for fuel nozzle development. Both new fluid designs and manufacturing technologies have led to the development of fuel nozzles that significantly reduce aircraft emissions. The goal of the ZCET program is to mesh the current technology of Lean Direct Injection and rocket injectors to provide quick mixing, low emissions, and high-performance fuel nozzle designs. An experimental program is planned to investigate the fuel nozzle concepts in a flametube test rig. Currently, a hydrogen system is being installed in cell 23 at NASA Glenn Research Center's Research Combustion Laboratory. Testing will be conducted on a variety of fuel nozzle concepts up to combustion pressures of 350 psia and inlet air temperatures of 1200 F

  8. Changes in feedwater organic matter concentrations based on intake type and pretreatment processes at SWRO facilities, Red Sea, Saudi Arabia

    KAUST Repository

    Dehwah, Abdullah

    2015-03-01

    Transparent exopolymer particles (TEP), natural organic matter, and bacterial concentrations in feedwater are important factors that can lead to membrane biofouling in seawater reverse osmosis (SWRO) systems. Two methods for controlling these concentrations in the feedwater prior to pretreatment have been suggested; use of subsurface intake systems or placement of the intake at a greater depth in the sea. These proposed solutions were tested at two SWRO facilities located along the Red Sea of Saudi Arabia. A shallow well intake system was very effective in reducing the algae and bacterial concentrations and somewhat effective in reducing TEP concentrations. An intake placed at a depth of 9. m below the surface was found to have limited impact on improving water quality compared to a surface intake. The algae and bacteria concentration in the feedwater (deep) was lower compared to the surface seawater, but the overall TEP concentration was higher. Bacteria and TEP measurements made in the pretreatment process train in the plant and after the cartridge filters suggest that regrowth of bacteria is occurring within the cartridge filters.

  9. Numerical simulation of a 374 tons/h water-tube steam boiler following a feedwater line break

    International Nuclear Information System (INIS)

    Deghal Cheridi, Amina Lyria; Chaker, Abla; Loubar, Ahcène

    2016-01-01

    Highlights: • We simulate the behavior of a steam boiler during feed-water line break accident. • To perform accident analysis of the steam boiler, Relap5/Mod3.2 system code is used. • A Relap5 model of the boiler is developed and qualified at the steady state level. • A good agreement between Relap5 results and available experimental data. • The Relap5 model predicts well the main transient features of the boiler. - Abstract: To ensure the operational safety of an industrial water-tube steam boiler it is very important to assess various accident scenarios in real plant working conditions. One of the most challenging scenarios is the loss of feedwater to the steam boiler. In this paper, a simulation of the behavior of an industrial water-tube radiant steam boiler during feedwater line break accident is discussed. The simulation is carried out using the RELAP5 system code. The steam boiler is installed in an Algerian natural gas liquefaction complex. The simulation shows the capabilities of RELAP5 system code in predicting the behavior of the steam boiler at both steady state and transient working conditions. From another side, the behavior of the steam boiler following the accident shows how the control system can successfully mitigate the effects and consequences of such accident and how the evaporator tubes can undergo a severe damage due to an uncontrolled increase of the wall temperature in case of failure of this system.

  10. Pipe whip and impact

    International Nuclear Information System (INIS)

    Attwood, G.J.

    1987-01-01

    Over the past few years changes in economic and safety considerations in nuclear power plants have resulted in a need to examine the problem of pipe whip in greater detail. Consequently, experimental programmes were set up in France, North America and Britain. Results from these tests combined with analytical work indicate that pipe whip followed by impact with surrounding pipework and structures may not be as serious as had been believed. Impact loads have been found to be much less (at least five times) than those predicted to the appropriate design regulations. Hence, the use of pipe whip restraints may have been overconservative. The use of fewer, better designed restraints, would result in greater accessibility of pipework, a reduced need for inspection of restraints, and a considerable financial saving. (author)

  11. Heat pipes and use of heat pipes in furnace exhaust

    Science.gov (United States)

    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.

  12. PE 100 pipe systems

    CERN Document Server

    Brömstrup, Heiner

    2012-01-01

    English translation of the 3rd edition ""Rohrsysteme aus PE 100"". Because of the considerably increased performance, pipe and pipe systems made from 100 enlarge the range of applications in the sectors of gas and water supply, sewage disposal, industrial pipeline construction and in the reconstruction and redevelopment of defective pipelines (relining). This book applies in particular to engineers, technicians and foremen working in the fields of supply, disposal and industry. Subject matters of the book are all practice-relevant questions regarding the construction, operation and maintenance

  13. Performance of buried pipe installation.

    Science.gov (United States)

    2010-05-01

    The purpose of this study is to determine the effects of geometric and mechanical parameters : characterizing the soil structure interaction developed in a buried pipe installation located under : roads/highways. The drainage pipes or culverts instal...

  14. Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1993-12-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant

  15. ASSESSMENT OF CONDENSATION HEAT TRANSFER MODEL TO EVALUATE PERFORMANCE OF THE PASSIVE AUXILIARY FEEDWATER SYSTEM

    Directory of Open Access Journals (Sweden)

    YUN-JE CHO

    2013-11-01

    Full Text Available As passive safety features for nuclear power plants receive increasing attention, various studies have been conducted to develop safety systems for 3rd-generation (GEN-III nuclear power plants that are driven by passive systems. The Passive Auxiliary Feedwater System (PAFS is one of several passive safety systems being designed for the Advanced Power Reactor Plus (APR+, and extensive studies are being conducted to complete its design and to verify its feasibility. Because the PAFS removes decay heat from the reactor core under transient and accident conditions, it is necessary to evaluate the heat removal capability of the PAFS under hypothetical accident conditions. The heat removal capability of the PAFS is strongly dependent on the heat transfer at the condensate tube in Passive Condensation Heat Exchanger (PCHX. To evaluate the model of heat transfer coefficient for condensation, the Multi-dimensional Analysis of Reactor Safety (MARS code is used to simulate the experimental results from PAFS Condensing Heat Removal Assessment Loop (PASCAL. The Shah model, a default model for condensation heat transfer coefficient in the MARS code, under-predicts the experimental data from the PASCAL. To improve the calculation result, The Thome model and the new version of the Shah model are implemented and compared with the experimental data.

  16. Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1994-05-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant

  17. Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Vehec, T.A.

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Point Beach was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRS. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Point Beach plant

  18. Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

    International Nuclear Information System (INIS)

    Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

    1993-08-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant

  19. ECOSIM - Applied to a study on the thermo-hydraulic behaviour of feedwater heaters

    International Nuclear Information System (INIS)

    Huelamo Martinez, E.; Casado Flores, E.; Bosch Aparicio, F.

    1998-01-01

    In order to carry out a behaviour study on the secondary circuit of a nuclear power plant operating at a load level higher than originally planned, it is essential to know if the cycle heaters are valid from the thermo-dynamic point of view. This paper describes the models which were used for the study of certain heaters; these models were validated by checking that they faithfully reproduced the behaviour of the equipment (TTD and DCA) in areas where data from the manufacturer was available. The behaviour of said equipment was later obtained in the foreseen operating range. The calculations necessary for these studies were carried out by building ECOSIM models, taking into account that the behaviour of the feedwater heaters depends both on the entry conditions of the extraction steam and also on the remaining mass and energy inputs. For this reason the actual plant layout was taken into consideration, as it was different from the original design. This paper describes the starting hypothesis, the correlations used, the results obtained, an analysis of said results, and a comparison with the manufacturer's data where available. (Author)

  20. The choice of terminal temperature difference for closed feedwater heaters in a steam power cycle

    International Nuclear Information System (INIS)

    D'Errico, P.A.; Andreone, C.F.

    1989-01-01

    Terminal temperature difference (TTD) and the drains cooler approach (DCA), the thermal performance parameters for closed feedwater heaters, (CFWH) in steam power cycles, frequently are selected on the basis of economics. In the design study process, optimal heater performance is determined by analyzing the trends in overall cycle performance economics as they vary with heater TTD and DCA. TTD and DCA are selected to provide the lowest evaluated cost, taking into account unit performance cost and the investment costs associated with the heaters as these performance parameters are varied. Practical constraints such as an overall heater length, tube side velocities, desuperheating (DSH) zone wet wall, etc, must be considered as well. An exception occurs when this procedure is applied to the TTD of the highest pressure heater in the cycle. When calculating overall cycle performance, the heat rate is found to improve with improved (lower) TTD results in greater investment costs, the evaluation for the highest extraction pressure CFWH may be contrary to those obtained for similar calculations on other heaters. This paper discusses guidelines for calculating heater performance parameters and specifically addresses the practical aspects of choosing the TTD's for CFWH in a steam power cycle

  1. Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

    International Nuclear Information System (INIS)

    Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1991-07-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab

  2. Feedwater transient and small break loss of coolant accident analyses for the Bellefonte Nuclear Plant

    Energy Technology Data Exchange (ETDEWEB)

    Bayless, P D; Dobbe, C A; Chambers, R

    1987-03-01

    Specific sequences that may lead to core damage were analyzed for the Bellefonte nuclear plant as part of the US Nuclear Regulatory Commission's Severe Accident Sequence Analysis Program. The RELAP5, SCDAP, and SCDAP/RELAP5 computer codes were used in the analyses. The two main initiating events investigated were a loss of all feedwater to the steam generators and a small cold leg break loss of coolant accident. The transients of primary interest within these categories were the TMLB' and S/sub 2/D sequences. Variations on systems availability were also investigated. Possible operator actions that could prevent or delay core damage were identified, and two were investigated for a small break transient. All of the transients were analyzed until either core damage began or long-term decay heat removal was established. The analyses showed that for the sequences considered the injection flow from one high-pressure injection pump was necessary and sufficient to prevent core damage in the absence of operator actions. Operator actions were able to prevent core damage in the S/sub 2/D sequence; no operator actions were available to prevent core damage in the TMLB' sequence.

  3. Digital controller for feedwater and recirculation flow control of BWR plant

    International Nuclear Information System (INIS)

    Sato, Takao; Ito, Tetsuo; Omori, Takashi; Iida, Hiroshi; Yanai, Katsuya.

    1980-01-01

    In nuclear power plants, it is required to operate the plants on load-following basis as the proportion of nuclear power plants in whole power supply network has been increasing. For this purpose, the requirements of more reliable, more automated plants and of the flexibility of operation are becoming serious. To respond to such demands, digital controllers are inevitable because analog controllers are limited in their controllability. It is also required to devise more intelligent systems such as those enabling strengthened diagnostic functions or sophisticated predictive control. On such background, duplicated redundant digital control system has been developed, using two control microcomputers and uniting the conventional feed-water control system and recirculation flow control system. The report discribes on the design concept for this digital controller, the hardware and software of the control system and the confirmation of the performance by simulation. The verifying test for the control performance, the simulation test for recirculation pump abnormality, the test for predictive control and the test on the response characteristics of recirculation system were carried out. The digital controller attained the MTBF 10 times as much, and the down time ratio 1/5 as small as those of the analog control systems. (Wakatsuki, Y.)

  4. GEOTHERMAL / SOLAR HYBRID DESIGNS: USE OF GEOTHERMAL ENERGY FOR CSP FEEDWATER HEATING

    Energy Technology Data Exchange (ETDEWEB)

    Craig Turchi; Guangdong Zhu; Michael Wagner; Tom Williams; Dan Wendt

    2014-10-01

    This paper examines a hybrid geothermal / solar thermal plant design that uses geothermal energy to provide feedwater heating in a conventional steam-Rankine power cycle deployed by a concentrating solar power (CSP) plant. The geothermal energy represents slightly over 10% of the total thermal input to the hybrid plant. The geothermal energy allows power output from the hybrid plant to increase by about 8% relative to a stand-alone CSP plant with the same solar-thermal input. Geothermal energy is converted to electricity at an efficiency of 1.7 to 2.5 times greater than would occur in a stand-alone, binary-cycle geothermal plant using the same geothermal resource. While the design exhibits a clear advantage during hybrid plant operation, the annual advantage of the hybrid versus two stand-alone power plants depends on the total annual operating hours of the hybrid plant. The annual results in this draft paper are preliminary, and further results are expected prior to submission of a final paper.

  5. Audit calculation of the limiting CESSAR feedwater-line-break transient with RELAP5/MOD1

    International Nuclear Information System (INIS)

    Chung, K.S.; Kennedy, M.F.; Guttmann, J.

    1983-01-01

    Argonne National Laboratory (ANL) performed a series of audit calculations of the limiting FLB transient presented in Appendix 15B to the CESSAR FSAR, supported by a limited number of additional calculations to investigate the sensitivity of the results (in terms of peak primary reactor system pressure) to break area and reactor trip time. The latter calculations were performed to quantify potential benefits in crediting reactor tip on low steam generator downcomer water level, which occurs earlier than the trip shown in the limiting FSAR transient, which tripped on high pressurizer pressure. These calculations were performed to verify the break spectrum results presented by C-E and to insure that C-E did indeed analyze the limiting transient. All of the ANL calculations were performed with RELAP5/MOD1 (cycle 18) using an input deck developed at ANL from CESSAR plant data provided by C-E. In this paper we compare the results and provide insight into the generic behavior of a Feedwater Line Break transient

  6. Influence of reactor vessel nodalization in the coupled code analysis of Asymmetric Main Feedwater Isolation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Grgic, D.

    2001-01-01

    Asymmetric Main Feedwater Isolation (AMFWI) transient in one Steam Generator (SG) for NPP Krsko using RELAP5 standalone code and coupled code RELAP5- QUABOX/CUBBOX (R5QC) was analyzed. In the RELAP5 standalone calculation, a point kinetics model was used, while in the coupled code a three-dimensional (3D) neutronics model of QUABOX with different RELAP5 nodalization schemes of reactor vessel was used. Both code versions use best-estimate thermal-hydraulic system code for all components in the plant and include realistic description of plant protection and control systems. Two different types of calculations were performed: with and without automatic control rod system available. The AMFWI transient causes the great asymmetry of the transferred heat in the SGs and subsequently the asymmetry of the power produced across the core due to different reactivity feedback resulting from the thermal-hydraulic channels assigned to different loops. The work presented in the paper is a part of validation of the 3D coupled code R5QC in the analysis of asymmetric transients.(author)

  7. Impact test on a pipe

    International Nuclear Information System (INIS)

    Noronha, R.F.; Shimura, S.H.

    1995-01-01

    A carbon steel pipe was submitted to a series of progressive impact loadings on a drop table. Strain and acceleration were measured in relevant places of the pipe and recorded in time and frequency domain. The pipe withstood impact loads up to 560 G without any visual deformation being noticed. Strains 50% over the yield point strain were measured. A high level of damping coming from the supports may have attenuated the response of the pipe. (author). 4 refs., 1 fig., 1 tab

  8. Computational Fluid Dynamics Simulation of Dual Bell Nozzle Film Cooling

    Science.gov (United States)

    Braman, Kalen; Garcia, Christian; Ruf, Joseph; Bui, Trong

    2015-01-01

    Marshall Space Flight Center (MSFC) and Armstrong Flight Research Center (AFRC) are working together to advance the technology readiness level (TRL) of the dual bell nozzle concept. Dual bell nozzles are a form of altitude compensating nozzle that consists of two connecting bell contours. At low altitude the nozzle flows fully in the first, relatively lower area ratio, nozzle. The nozzle flow separates from the wall at the inflection point which joins the two bell contours. This relatively low expansion results in higher nozzle efficiency during the low altitude portion of the launch. As ambient pressure decreases with increasing altitude, the nozzle flow will expand to fill the relatively large area ratio second nozzle. The larger area ratio of the second bell enables higher Isp during the high altitude and vacuum portions of the launch. Despite a long history of theoretical consideration and promise towards improving rocket performance, dual bell nozzles have yet to be developed for practical use and have seen only limited testing. One barrier to use of dual bell nozzles is the lack of control over the nozzle flow transition from the first bell to the second bell during operation. A method that this team is pursuing to enhance the controllability of the nozzle flow transition is manipulation of the film coolant that is injected near the inflection between the two bell contours. Computational fluid dynamics (CFD) analysis is being run to assess the degree of control over nozzle flow transition generated via manipulation of the film injection. A cold flow dual bell nozzle, without film coolant, was tested over a range of simulated altitudes in 2004 in MSFC's nozzle test facility. Both NASA centers have performed a series of simulations of that dual bell to validate their computational models. Those CFD results are compared to the experimental results within this paper. MSFC then proceeded to add film injection to the CFD grid of the dual bell nozzle. A series of

  9. Jet-Surface Interaction: High Aspect Ratio Nozzle Test, Nozzle Design and Preliminary Data

    Science.gov (United States)

    Brown, Clifford; Dippold, Vance

    2015-01-01

    The Jet-Surface Interaction High Aspect Ratio (JSI-HAR) nozzle test is part of an ongoing effort to measure and predict the noise created when an aircraft engine exhausts close to an airframe surface. The JSI-HAR test is focused on parameters derived from the Turbo-electric Distributed Propulsion (TeDP) concept aircraft which include a high-aspect ratio mailslot exhaust nozzle, internal septa, and an aft deck. The size and mass flow rate limits of the test rig also limited the test nozzle to a 16:1 aspect ratio, half the approximately 32:1 on the TeDP concept. Also, unlike the aircraft, the test nozzle must transition from a single round duct on the High Flow Jet Exit Rig, located in the AeroAcoustic Propulsion Laboratory at the NASA Glenn Research Center, to the rectangular shape at the nozzle exit. A parametric nozzle design method was developed to design three low noise round-to-rectangular transitions, with 8:1, 12:1, and 16: aspect ratios, that minimizes flow separations and shocks while providing a flat flow profile at the nozzle exit. These designs validated using the WIND-US CFD code. A preliminary analysis of the test data shows that the actual flow profile is close to that predicted and that the noise results appear consistent with data from previous, smaller scale, tests. The JSI-HAR test is ongoing through October 2015. The results shown in the presentation are intended to provide an overview of the test and a first look at the preliminary results.

  10. Study on high throughput nanomanufacturing of photopatternable nanofibers using tube nozzle electrospinning with multi-tubes and multi-nozzles

    Science.gov (United States)

    Fang, Sheng-Po; Jao, PitFee; Senior, David E.; Kim, Kyoung-Tae; Yoon, Yong-Kyu

    2017-12-01

    High throughput nanomanufacturing of photopatternable nanofibers and subsequent photopatterning is reported. For the production of high density nanofibers, the tube nozzle electrospinning (TNE) process has been used, where an array of micronozzles on the sidewall of a plastic tube are used as spinnerets. By increasing the density of nozzles, the electric fields of adjacent nozzles confine the cone of electrospinning and give a higher density of nanofibers. With TNE, higher density nozzles are easily achievable compared to metallic nozzles, e.g. an inter-nozzle distance as small as 0.5 cm and an average semi-vertical repulsion angle of 12.28° for 8-nozzles were achieved. Nanofiber diameter distribution, mass throughput rate, and growth rate of nanofiber stacks in different operating conditions and with different numbers of nozzles, such as 2, 4 and 8 nozzles, and scalability with single and double tube configurations are discussed. Nanofibers made of SU-8, photopatternable epoxy, have been collected to a thickness of over 80 μm in 240 s of electrospinning and the production rate of 0.75 g/h is achieved using the 2 tube 8 nozzle systems, followed by photolithographic micropatterning. TNE is scalable to a large number of nozzles, and offers high throughput production, plug and play capability with standard electrospinning equipment, and little waste of polymer.

  11. Optimization of Pipe Networks

    DEFF Research Database (Denmark)

    Hansen, C. T.; Madsen, Kaj; Nielsen, Hans Bruun

    1991-01-01

    algorithm using successive linear programming is presented. The performance of the algorithm is illustrated by optimizing a network with 201 pipes and 172 nodes. It is concluded that the new algorithm seems to be very efficient and stable, and that it always finds a solution with a cost near the best...

  12. Flexible Heat Pipe

    Science.gov (United States)

    Bienert, W. B.; Wolf, D. A.

    1985-01-01

    Narrow Tube carries 10 watts or more to moving parts. Heat pipe 12 inches long and diameter of 0.312 inch (7.92mm). Bent to minimum radius of 2.5 blocks. Flexible section made of 321 stainless steel tubing (Cajon Flexible Tubing or equivalent). Evaporator and condenser made of oxygen free copper. Working fluid methanol.

  13. HPFRCC - Extruded Pipes

    DEFF Research Database (Denmark)

    Stang, Henrik; Pedersen, Carsten

    1996-01-01

    The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique developme...... of the newly extruded material....

  14. PDE Nozzle Optimization Using a Genetic Algorithm

    Science.gov (United States)

    Billings, Dana; Turner, James E. (Technical Monitor)

    2000-01-01

    Genetic algorithms, which simulate evolution in natural systems, have been used to find solutions to optimization problems that seem intractable to standard approaches. In this study, the feasibility of using a GA to find an optimum, fixed profile nozzle for a pulse detonation engine (PDE) is demonstrated. The objective was to maximize impulse during the detonation wave passage and blow-down phases of operation. Impulse of each profile variant was obtained by using the CFD code Mozart/2.0 to simulate the transient flow. After 7 generations, the method has identified a nozzle profile that certainly is a candidate for optimum solution. The constraints on the generality of this possible solution remain to be clarified.

  15. Experiments on black liquor splashplate nozzle performance

    Energy Technology Data Exchange (ETDEWEB)

    Nieminen, K.

    1996-12-31

    The performance of a throttled black liquor splashplate nozzle was studied in this work. A series of industrial-scale experiments were performed using mass flow rate as a variable at a fixed temperature. The experiments were carried out in a spraying chamber next to the recovery boiler with real mill liquor. The disintegration process of the liquor sheet was videotaped for analyzing. The mass flow rate distribution was measured with a collector. The liquor drops produced by the nozzle were videotaped and measured with a video image analysis technique. The industrial-scale experiments were afterwards repeated on a small scale in the laboratory environment which made it possible to study the liquid sheet disintegration process thoroughly. The small-scale experiments were carried out with a solution of water and glycerol and a splashplate nozzle of approximately one tenth the size of full-scale nozzle. The whole liquid sheet and close-up exposures of the plate area were videotaped. However, the videotaping equipment (camera and objective) were not capable of observing the very thin and transparent liquid sheet. The mass flow rate distribution was measured with steps of 2.5 deg from the plate centerline with a collector device. The drop sizes were measured from various sheet angles with Malvern Particle Sizer and a phase Doppler particle anemometer (Aerometrics). The modeling was based on dimensional analysis. The objective was to compare these two experimental settings and to find out whether small-scale experiments can be used in predicting the spraying characteristics in the full-scale. It was also of interest to test the measured black liquor drop sizes against drop size correlations obtained from the literature. (31 refs.)

  16. Thiokol 260-SL Nozzle Development Program

    Science.gov (United States)

    1967-01-01

    excited by the above environments were investigated. These were: (a) lateral vibration of the nozzle exit cone as a cantilever beam , (b) radial vibration...under the debulking roller to prevent springback of the material. The cooling of the wrapped tape tended to set the material and prevent subsequent...between sheets of nylon were then placed at intervals on the surface of the shell as shown in Figure 10. The convergent ablative stack was then

  17. Wormhole Formation in RSRM Nozzle Joint Backfill

    Science.gov (United States)

    Stevens, J.

    2000-01-01

    The RSRM nozzle uses a barrier of RTV rubber upstream of the nozzle O-ring seals. Post flight inspection of the RSRM nozzle continues to reveal occurrence of "wormholes" into the RTV backfill. The term "wormholes", sometimes called "gas paths", indicates a gas flow path not caused by pre-existing voids, but by a little-understood internal failure mode of the material during motor operation. Fundamental understanding of the mechanics of the RSRM nozzle joints during motor operation, nonlinear viscoelastic characterization of the RTV backfill material, identification of the conditions that predispose the RTV to form wormholes, and screening of candidate replacement materials is being pursued by a joint effort between Thiokol Propulsion, NASA, and the Army Propulsion & Structures Directorate at Redstone Arsenal. The performance of the RTV backfill in the joint is controlled by the joint environment. Joint movement, which applies a tension and shear load on the material, coupled with the introduction of high pressure gas in combination create an environment that exceeds the capability of the material to withstand the wormhole effect. Little data exists to evaluate why the material fails under the modeled joint conditions, so an effort to characterize and evaluate the material under these conditions was undertaken. Viscoelastic property data from characterization testing will anchor structural analysis models. Data over a range of temperatures, environmental pressures, and strain rates was used to develop a nonlinear viscoelastic model to predict material performance, develop criteria for replacement materials, and quantify material properties influencing wormhole growth. Three joint simulation analogs were developed to analyze and validate joint thermal barrier (backfill) material performance. Two exploratory tests focus on detection of wormhole failure under specific motor operating conditions. A "validation" test system provides data to "validate" computer models and

  18. Analysis of film cooling in rocket nozzles

    Science.gov (United States)

    Woodbury, Keith A.

    1992-01-01

    Computational Fluid Dynamics (CFD) programs are customarily used to compute details of a flow field, such as velocity fields or species concentrations. Generally they are not used to determine the resulting conditions at a solid boundary such as wall shear stress or heat flux. However, determination of this information should be within the capability of a CFD code, as the code supposedly contains appropriate models for these wall conditions. Before such predictions from CFD analyses can be accepted, the credibility of the CFD codes upon which they are based must be established. This report details the progress made in constructing a CFD model to predict the heat transfer to the wall in a film cooled rocket nozzle. Specifically, the objective of this work is to use the NASA code FDNS to predict the heat transfer which will occur during the upcoming hot-firing of the Pratt & Whitney 40K subscale nozzle (1Q93). Toward this end, an M = 3 wall jet is considered, and the resulting heat transfer to the wall is computed. The values are compared against experimental data available in Reference 1. Also, FDNS's ability to compute heat flux in a reacting flow will be determined by comparing the code's predictions against calorimeter data from the hot firing of a 40K combustor. The process of modeling the flow of combusting gases through the Pratt & Whitney 40K subscale combustor and nozzle is outlined. What follows in this report is a brief description of the FDNS code, with special emphasis on how it handles solid wall boundary conditions. The test cases and some FDNS solution are presented next, along with comparison to experimental data. The process of modeling the flow through a chamber and a nozzle using the FDNS code will also be outlined.

  19. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

    30, also a gap of 3 mm was applied in X and Z directions of the support at the node 155. The axial force (FY that occurred in the pump outlet nozzle (dia. 4 in. of PriPump line have also exceeded the allowable limit that lead to the pump nozzle failure during an earthquake of Lembang fault. The modifications is necessary to be applied on the cooling system for PriPump line so the nozzle would not receive the force that exceed the allowable limits. The modification can be done by removing the support at node 105 and node 135 so the primary cooling system piping of Bandung TRIGA 2000 reactor would be safe to operate during an earthquake originated from Lembang fault.

  20. Nonequilibrium in a low power arcjet nozzle

    Science.gov (United States)

    Zube, Dieter M.; Myers, Roger M.

    1991-01-01

    Emission spectroscopy measurements were made of the plasma flow inside the nozzle of a 1 kW class arcjet thruster. The thruster propellant was a hydrogen-nitrogen mixture used to simulate fully decomposed hydrazine. The 0.25 mm diameter holes were drilled into the diverging section of the tungsten thruster nozzle to provide optical access to the internal flow. Atomic electron excitation, vibrational, and rotational temperatures were determined for the expanding plasma using relative line intensity techniques. The atomic excitation temperatures decreased from 18,000K at a location 3 mm downstream of the constrictor to 9,000K at a location 9 mm from the constrictor, while the molecular vibrational and rotational temperatures decreased from 6,500K to 2,500K and from 8,000K to 3,000K, respectively, between the same locations. The electron density measured using hydrogen H line Stark broadening decreased from about 10(exp 15) cm(-3) to about 2 times 10(exp 14) cm(-3) during the expansion. The results show that the plasma is highly nonequilibrium throughout the nozzle, with most relaxation times equal or exceeding the particle residence time.

  1. Coherent structures in a supersonic complex nozzle

    Science.gov (United States)

    Magstadt, Andrew; Berry, Matthew; Glauser, Mark

    2016-11-01

    The jet flow from a complex supersonic nozzle is studied through experimental measurements. The nozzle's geometry is motivated by future engine designs for high-performance civilian and military aircraft. This rectangular jet has a single plane of symmetry, an additional shear layer (referred to as a wall jet), and an aft deck representative of airframe integration. The core flow operates at a Mach number of Mj , c = 1 . 6 , and the wall jet is choked (Mj , w = 1 . 0). This high Reynolds number jet flow is comprised of intense turbulence levels, an intricate shock structure, shear and boundary layers, and powerful corner vortices. In the present study, stereo PIV measurements are simultaneously sampled with high-speed pressure measurements, which are embedded in the aft deck, and far-field acoustics in the anechoic chamber at Syracuse University. Time-resolved schlieren measurements have indicated the existence of strong flow events at high frequencies, at a Strouhal number of St = 3 . 4 . These appear to result from von Kàrmàn vortex shedding within the nozzle and pervade the entire flow and acoustic domain. Proper orthogonal decomposition is applied on the current data to identify coherent structures in the jet and study the influence of this vortex street. AFOSR Turbulence and Transition Program (Grant No. FA9550-15-1-0435) with program managers Dr. I. Leyva and Dr. R. Ponnappan.

  2. Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator

    International Nuclear Information System (INIS)

    Sanchez B, A.

    2003-01-01

    The simulation process of the component systems of the feedwater of a nucleo electric plant is presented, using several models of reduced order that represent the diverse elements that compose the systems like: the heaters of feedwater, the condenser, the feedwater pump, etc. The integration of the same ones in one simulative structure, and the development of a platform that to give the appearance of to be executed in continuous time, it is the objective of the feedwater simulator, as well as of the SUN-RAH simulator, of which is part. The simulator uses models of reduced order that respond to the observed behavior of a nuclear plant of BWR type. Likewise, it is presented a model of a flow controller of feedwater that will be the one in charge of regulating the demand of the system according to the characteristics and criticize restrictions of safety and controllability, assigned according to those wanted parameters of performance of this system inside the nucleo electric plant. The integration of these models, the adaptation of the variables and parameters, are presented in a way that the integration with the other ones models of the remaining systems of the plant (reactor, steam lines, turbine, etc.), be direct and coherent with the principles of thermodynamic cycles relative to this type of generation plants. The design of those graphic interfaces and the environment where the simulator works its are part of those developments of this work. The reaches and objectives of the simulator complement the description of the simulator. (Author)

  3. 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

  4. 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

  5. The mitigation of flow-accelerated corrosion in the feedwater systems of nuclear reactors - the influence of dissolved oxygen under different operating conditions

    International Nuclear Information System (INIS)

    Lister, D.; Feicht, A.; Fujiwara, K.; Khatibi, M.; Liu, L.; Ohira, T.; Uchida, S.

    2010-01-01

    In order to improve our understanding of the flow-accelerated corrosion (FAC) of carbon steel piping in feedwater systems, a collaborative research program between Japan and Canada has investigated the combined effects of system operating parameters. A major objective was to optimize techniques for minimizing degradation; accordingly, we report here the influence of dissolved oxygen on FAC rate under a range of conditions as examined in the laboratory with a high-temperature water loop. Most of the experiments were done at 140°C in neutral water, in ammoniated water at pH 25°C 9.2 and in ammoniated water at pH 25°C 9.2 with 100 ppb hydrazine. Several flow rates were imposed and two grades of carbon steel were employed for test probes: one containing 0.019% chromium and the other containing 0.001% chromium. Probes were designed to monitor continuously both FAC (by an electrical resistance technique) and electrochemical corrosion potential, ECP (relative to a Ag/AgCl reference electrode). During a typical experiment, probes of different diameter were installed in series. Downstream of the 'resistance probes' to measure FAC, removable probes for detailed surface examination by techniques such as laser-Raman microscopy, scanning-electron microscopy and energy-dispersive x-ray analysis were also installed. In neutral water, FAC rates reached ~6 mm/a and were apparently controlled by mass transfer, which led to a numerical correlation with fluid shear stress at the tube wall. The steel with the lower chromium content had a greater FAC rate by a factor of about 2.4. An oxygen concentration of almost 40 ppb was required to stifle FAC. In ammoniated water, FAC rates were relatively low and correlations depending on mass transfer were dubious. Hydrazine reduced the FAC rate, possibly because it affected the pH locally at the metal surface. It also increased the stifling oxygen concentration at pH 25°C 9.2 to ~3 ppb. At pH 25°C 10, FAC could not be detected

  6. Drill pipe protector development

    Energy Technology Data Exchange (ETDEWEB)

    Thomerson, C.; Kenne, R. [Regal International Corp., Corsicanna, TX (United States); Wemple, R.P. [Sandia National Lab., Albuquerque, NM (United States)] [ed.] [and others

    1996-03-01

    The Geothermal Drilling Organization (GDO), formed in the early 1980s by the geothermal industry and the U.S. Department of Energy (DOE) Geothermal Division, sponsors specific development projects to advance the technologies used in geothermal exploration, drilling, and production phases. Individual GDO member companies can choose to participate in specific projects that are most beneficial to their industry segment. Sandia National Laboratories is the technical interface and contracting office for the DOE in these projects. Typical projects sponsored in the past have included a high temperature borehole televiewer, drill bits, muds/polymers, rotary head seals, and this project for drill pipe protectors. This report documents the development work of Regal International for high temperature geothermal pipe protectors.

  7. Diffusion in flexible pipes

    Energy Technology Data Exchange (ETDEWEB)

    Brogaard Kristensen, S.

    2000-06-01

    This report describes the work done on modelling and simulation of the complex diffusion of gas through the wall of a flexible pipe. The diffusion and thus the pressure in annulus depends strongly on the diffusion and solubility parameters of the gas-polymer system and on the degree of blocking of the outer surface of the inner liner due to pressure reinforcements. The report evaluates the basis modelling required to describe the complex geometries and flow patterns. Qualitatively results of temperature and concentration profiles are shown in the report. For the program to serve any modelling purpose in 'real life' the results need to be validated and possibly the model needs corrections. Hopefully, a full-scale test of a flexible pipe will provide the required temperatures and pressures in annulus to validate the models. (EHS)

  8. LHCb: Beam Pipe portrait

    CERN Multimedia

    LHCb, Collaboration

    2005-01-01

    The proton beams circulate in the accelerator in Ultra High Vacuum to make them interact only with each other when colliding at the interaction point. A special beam pipe "holds" the vacuum where they pass through the LHCb detector: it has to be mechanically very strong to stand the difference in pressure between the vacuum inside it and the air in the cavern but also be as transparent as possible for the particles originating in the proton−proton collisions.

  9. LHCb: Beam Pipe

    CERN Multimedia

    LHCb, Collaboration

    2005-01-01

    The proton beams circulate in the accelerator in Ultra High Vacuum to make them interact only with each other when colliding at the interaction point. A special beam pipe "holds" the vacuum where they pass through the LHCb detector:it has to be mechanically very strong to stand the difference in pressure between the vacuum inside it and the air in the cavern but also be as transparent as possible for the particles originating in the proton−proton collisions.

  10. 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

  11. Pressurizer with a mechanically attached surge nozzle thermal sleeve

    Energy Technology Data Exchange (ETDEWEB)

    Wepfer, Robert M

    2014-03-25

    A thermal sleeve is mechanically attached to the bore of a surge nozzle of a pressurizer for the primary circuit of a pressurized water reactor steam generating system. The thermal sleeve is attached with a series of keys and slots which maintain the thermal sleeve centered in the nozzle while permitting thermal growth and restricting flow between the sleeve and the interior wall of the nozzle.

  12. Fluidized-bed calciner with combustion nozzle and shroud

    International Nuclear Information System (INIS)

    Wielang, J.A.; Palmer, W.B.; Kerr, W.B.

    1977-01-01

    A nozzle employed as a burner within a fluidized bed is coaxially enclosed within a tubular shroud that extends beyond the nozzle length into the fluidized bed. The open-ended shroud portion beyond the nozzle end provides an antechamber for mixture and combustion of atomized fuel with an oxygen-containing gas. The arrangement provides improved combustion efficiency and excludes bed particles from the high-velocity, high-temperature portions of the flame to reduce particle attrition. 4 claims, 2 figures

  13. TRAC-PF1 analysis of LOFT steam-generator feedwater transient test L9-1

    International Nuclear Information System (INIS)

    Meier, J.K.

    1983-01-01

    The Transient Reactor Analysis Code (TRAC-PF1) calculations were compared to test data from Loss-of-Fluid Test (LOFT) L9-1, which was a loss-of-feedwater transient. This paper includes descriptions of the test and the TRAC input and compares the TRAC-calculated results with the test data. We conclude that the code predicted the experiment well, given the uncertainties in the boundary conditions. The analysis indicates the need to model all the flow paths and heat structures, and to improve the TRAC wall condensation heat-transfer model

  14. Active Nozzle Control and Integrated Design Optimization of a Beam Subject to Fluid-Dynamic Forces

    Science.gov (United States)

    Borglund, D.

    1999-02-01

    Active nozzle control is used to improve the stability of a beam subject to forces induced by fluid flow through attached pipes. The control system has a significant effect on the structural stability, making both flutter and divergence type of instabilities possible. The stability analysis is carried out using a state-variable approach based on a finite element formulation of the structural dynamics. The simultaneous design of the control system and the beam shape minimizing structural mass is performed using numerical optimization. The inclusion of the control system in the optimization gives a considerable reduction of the structural mass but results in an optimal design which is very sensitive to imperfections. Using a simple model of the control system uncertainties, a more robust design is obtained by solving a modified optimization problem. Throughout the study, the theoretical findings are verified by experiments.

  15. Mitigation of stress corrosion cracking in pressurized water reactor (PWR) piping systems using the mechanical stress improvement process (MSIPR) or underwater laser beam welding

    International Nuclear Information System (INIS)

    Rick, Grendys; Marc, Piccolino; Cunthia, Pezze; Badlani, Manu

    2009-01-01

    A current issue facing pressurized water reactors (PWRs) is primary water stress corrosion cracking (PWSCC) of bi metallic welds. PWSCC in a PWR requires the presence of a susceptible material, an aggressive environment and a tensile stress of significant magnitude. Reducing the potential for SCC can be accomplished by eliminating any of these three elements. In the U.S., mitigation of susceptible material in the pressurizer nozzle locations has largely been completed via the structural weld overlay (SWOL) process or NuVision Engineering's Mechanical Stress Improvement Process (MSIP R) , depending on inspectability. The next most susceptible locations in Westinghouse designed power plants are the Reactor Vessel (RV) hot leg nozzle welds. However, a full SWOL Process for RV nozzles is time consuming and has a high likelihood of in process weld repairs. Therefore, Westinghouse provides two distinctive methods to mitigate susceptible material for the RV nozzle locations depending on nozzle access and utility preference. These methods are the MSIP and the Underwater Laser Beam Welding (ULBW) process. MSIP applies a load to the outside diameter of the pipe adjacent to the weld, imposing plastic strains during compression that are not reversed after unloading, thus eliminating the tensile stress component of SCC. Recently, Westinghouse and NuVision successfully applied MSIP on all eight RV nozzles at the Salem Unit 1 power plant. Another option to mitigate SCC in RV nozzles is to place a barrier between the susceptible material and the aggressive environment. The ULBW process applies a weld inlay onto the inside pipe diameter. The deposited weld metal (Alloy 52M) is resistant to PWSCC and acts as a barrier to prevent primary water from contacting the susceptible material. This paper provides information on the approval and acceptance bases for MSIP, its recent application on RV nozzles and an update on ULBW development

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-05-01

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

  17. 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

  18. Erosion resistant nozzles for laser plasma extreme ultraviolet (EUV) sources

    Science.gov (United States)

    Kubiak, Glenn D.; Bernardez, II, Luis J.

    2000-01-04

    A gas nozzle having an increased resistance to erosion from energetic plasma particles generated by laser plasma sources. By reducing the area of the plasma-facing portion of the nozzle below a critical dimension and fabricating the nozzle from a material that has a high EUV transmission as well as a low sputtering coefficient such as Be, C, or Si, it has been shown that a significant reduction in reflectance loss of nearby optical components can be achieved even after exposing the nozzle to at least 10.sup.7 Xe plasma pulses.

  19. Low Cost Carbon-Carbon Rocket Nozzle Development, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This development will provide an inexpensive vacuum nozzle manufacturing option for NOFBXTM monopropellant systems that are currently being developed under NASA SBIR...

  20. Optimization of Profile and Material of Abrasive Water Jet Nozzle

    Science.gov (United States)

    Anand Bala Selwin, K. P.; Ramachandran, S.

    2017-05-01

    The objective of this work is to study the behaviour of the abrasive water jet nozzle with different profiles and materials. Taguchi-Grey relational analysis optimization technique is used to optimize the value with different material and different profiles. Initially the 3D models of the nozzle are modelled with different profiles by changing the tapered inlet angle of the nozzle. The different profile models are analysed with different materials and the results are optimized. The optimized results would give the better result taking wear and machining behaviour of the nozzle.

  1. Heat and fluid flow properties of circular impinging jet with a low nozzle to plate spacing. Improvement by nothched nozzle; Nozzle heibankan kyori ga chiisai baai no enkei shototsu funryu no ryudo dennetsu tokusei. Kirikaki nozzle ni yoru kaizen kojo

    Energy Technology Data Exchange (ETDEWEB)

    Shakouchih, T. [Mie University, Mie (Japan). Faculty of Engineering; Matsumoto, A.; Watanabe, A.

    2000-10-25

    It is well known that as decreasing the nozzle to plate spacing considerably the heat transfer coefficient of circular impinging jet, which impinges to the plate normally, increases remarkably. At that time, the flow resistance of nozzle-plate system also increases rapidly. In this study, in order to reduce the flow resistance and to enhance the heat transfer coefficient of the circular impinging jet with a considerably low nozzle to plate spacing, a special nozzle with notches is proposed, and considerable improvement of the flow and heat transfer properties are shown. The mechanism of enhancement of the heat transfer properties is also discussed. (author)

  2. Variable volume combustor with aerodynamic fuel flanges for nozzle mounting

    Science.gov (United States)

    McConnaughhay, Johnie Franklin; Keener, Christopher Paul; Johnson, Thomas Edward; Ostebee, Heath Michael

    2016-09-20

    The present application provides a combustor for use with a gas turbine engine. The combustor may include a number of micro-mixer fuel nozzles and a fuel injection system for providing a flow of fuel to the micro-mixer fuel nozzles. The fuel injection system may include a number of support struts supporting the fuel nozzles and for providing the flow of fuel therethrough. The fuel injection system also may include a number of aerodynamic fuel flanges connecting the micro-mixer fuel nozzles and the support struts.

  3. Simulation and Optimization of Diffuser/Nozzle Micropump

    Directory of Open Access Journals (Sweden)

    Chandika S.

    2011-12-01

    Full Text Available Design and analysis of diffuser/nozzle micropump using ANSYS-FLUENT is attempted for fuel delivery in automobile. To enhance the performance of the micropump a historic dimensional design such as the diffuser length, the diffuser angle, and the throat/neck width of diffuser/nozzle elements are obtained from the simulation results. The fluid velocity of the diffuer/nozzle and the pressure loss rates are calculated. The simulation result shows that there is an optimal dimension of the diffuser/nozzle to obtain a large flow rate and to minimize the velocity and the pressure losses.

  4. Jet flow issuing from an axisymmetric pipe-cavity-orifice nozzle

    Czech Academy of Sciences Publication Activity Database

    Broučková, Zuzana; Pušková, P.; Trávníček, Zdeněk; Šafařík, P.

    2016-01-01

    Roč. 114, March (2016), č. článku 02006. ISSN 2101-6275. [International Conference on Experimental Fluid Mechanics /10./. Praha, 17.11.2015-20.11.2015] R&D Projects: GA ČR GA14-08888S Institutional support: RVO:61388998 Keywords : passive control * acoustic excitations * resonator Subject RIV: JU - Aeronautics, Aerodynamics, Aircrafts

  5. Structure of the gas-liquid annular two-phase flow in a nozzle section

    International Nuclear Information System (INIS)

    Yoshida, Kenji; Kataoka, Isao; Ohmori, Syuichi; Mori, Michitsugu

    2006-01-01

    Experimental studies on the flow behavior of gas-liquid annular two-phase flow passing through a nozzle section were carried out. This study is concerned with the central steam jet injector for a next generation nuclear reactor. In the central steam jet injector, steam/water annular two-phase flow is formed at the mixing nozzle. To make an appropriate design and to establish the high-performance steam injector system, it is very important to accumulate the fundamental data of the thermo-hydro dynamic characteristics of annular flow passing through a nozzle section. On the other hand, the transient behavior of multiphase flow, in which the interactions between two-phases occur, is one of the most interesting scientific issues and has attracted research attention. In this study, the transient gas-phase turbulence modification in annular flow due to the gas-liquid phase interaction is experimentally investigated. The annular flow passing through a throat section is under the transient state due to the changing cross sectional area of the channel and resultantly the superficial velocities of both phases are changed compared with a fully developed flow in a straight pipe. The measurements for the gas-phase turbulence were precisely performed by using a constant temperature hot-wire anemometer, and made clear the turbulence structure such as velocity profiles, fluctuation velocity profiles. The behavior of the interfacial waves in the liquid film flow such as the ripple or disturbance waves was also observed. The measurements for the liquid film thickness by the electrode needle method were also performed to measure the base film thickness, mean film thickness, maximum film thickness and wave height of the ripple or the disturbance waves. (author)

  6. 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)

  7. Bottom nozzle of a LWR fuel assembly

    International Nuclear Information System (INIS)

    Leroux, J.C.

    1991-01-01

    The bottom nozzle consists of a transverse element in form of box having a bending resistant grid structure which has an outer peripheral frame of cross-section corresponding to that of the fuel assembly and which has walls defining large cells. The transverse element has a retainer plate with a regular array of openings. The retainer plate is fixed above and parallel to the grid structure with a spacing in order to form, between the grid structure and the retainer plate a free space for tranquil flow of cooling water and for debris collection [fr

  8. One- and Two-Phase Nozzle Flows.

    Science.gov (United States)

    1980-01-31

    PROJECT. TASK The Aerospace Corporation El Segundo, Calif. 90245 11. CONTROLLING OFFICE NAME AND ADDRESS Space Division31jnv 087 Air Force Systems Command...and identify by block .eintber) Gas-particle Two- phase Nozzle Transonic Flow Corn utational Method 20. AS Tf ACT (Continue an reverse side it...Dec. 1978. -51- 74.22 in. Fig.~~~~~~~ U 28.L USmalMOTOR Itro ofgrto n AEXI Fig. 2. BFC Gridl foor Smaio CUonfM igrtho n Somutaterged Noeglock x -344in

  9. Study of nozzle deposit formation mechanism for direct injection gasoline engines; Chokufun gasoline engine yo nozzle no deposit seisei kaiseki

    Energy Technology Data Exchange (ETDEWEB)

    Kinoshita, M.; Saito, A. [Toyota Central Research and Development Labs., Inc., Aichi (Japan); Matsushita, S. [Toyota Motor Corp., Aichi (Japan); Shibata, H. [Nippon Soken, Inc., Tokyo (Japan); Niwa, Y. [Denso Corp., Aichi (Japan)

    1997-10-01

    Nozzles in fuel injectors for direct injection gasoline engines are exposed to high temperature combustion gases and soot. In such a rigorous environment, it is a fear that fuel flow rate changes in injectors by deposit formation on nozzles. Fundamental factors of nozzle deposit formation were investigated through injector bench tests and engine dynamometer tests. Deposit formation processes were observed by SEM through engine dynamometer tests. The investigation results reveal nozzle deposit formation mechanism and how to suppress the deposit. 4 refs., 8 figs., 3 tabs.

  10. Water hammer in elastic pipes

    International Nuclear Information System (INIS)

    Gale, J.; Tiselj, I.

    2002-01-01

    One dimensional two-fluid six-equation model of two-phase flow, that can be found in computer codes like RELAP5, TRAC, and CATHARE, was upgraded with additional terms, which enable modelling of the pressure waves in elastic pipes. It is known that pipe elasticity reduces the propagation velocity of the shock and other pressure waves in the piping systems. Equations that include the pipe elasticty terms are used in WAHA code, which is being developed within the WAHALoads project of 5't'h EU research program.(author)

  11. Turbofan Noise Reduction Associated With Increased Bypass Nozzle Flow

    Science.gov (United States)

    Woodward, Richard P.; Hughes, Christopher E.

    2005-01-01

    An advanced 22-in. scale model turbofan, typical of a current-generation aircraft engine design by GE Aircraft Engines, was tested in NASA Glenn Research Center s 9- by 15- Foot Low-Speed Wind Tunnel to explore the far-field acoustic effects of an increased bypass nozzle area at simulated aircraft speeds of takeoff, approach, and landing. The wind-tunnel-scale model consisted of the bypass stage fan, stators, and nacelle (including the fan exit nozzle) of a typical turbofan. This fan-stage test was part of the NASA Glenn Fan Broadband Source Diagnostic Test, second entry, which acquired aeroacoustic results over a range of test conditions. A baseline nozzle was selected, and the nozzle area was chosen for maximum performance at sea-level conditions. Two additional nozzles were also tested--one with a 5.4-percent increase in nozzle area over the baseline nozzle (sized for design point conditions), corresponding to a 5-percent increase in fan weight flow, and another nozzle with a 10.9-percent increase in nozzle area over the baseline nozzle (sized for maximum weight flow at sea-level conditions), corresponding to a 7.5 percent increase in fan weight flow. Measured acoustic benefits with increased nozzle area were very encouraging, showing overall sound power level reductions of 2 dB or more (left graph) while the stage adiabatic efficiency (right graph) and thrust (final graph) actually increased by several percentage points. These noise-reduction benefits were seen to include both rotor-interaction tones and broadband noise, and were evident throughout the range of measured sideline angles.

  12. Analysis of film cooling in rocket nozzles

    Science.gov (United States)

    Woodbury, Keith A.

    1993-01-01

    This report summarizes the findings on the NASA contract NAG8-212, Task No. 3. The overall project consists of three tasks, all of which have been successfully completed. In addition, some supporting supplemental work, not required by the contract, has been performed and is documented herein. Task 1 involved the modification of the wall functions in the code FDNS (Finite Difference Navier-Stokes) to use a Reynolds Analogy-based method. This task was completed in August, 1992. Task 2 involved the verification of the code against experimentally available data. The data chosen for comparison was from an experiment involving the injection of helium from a wall jet. Results obtained in completing this task also show the sensitivity of the FDNS code to unknown conditions at the injection slot. This task was completed in September, 1992. Task 3 required the computation of the flow of hot exhaust gases through the P&W 40K subscale nozzle. Computations were performed both with and without film coolant injection. This task was completed in July, 1993. The FDNS program tends to overpredict heat fluxes, but, with suitable modeling of backside cooling, may give reasonable wall temperature predictions. For film cooling in the P&W 40K calorimeter subscale nozzle, the average wall temperature is reduced from 1750R to about 1050R by the film cooling. The average wall heat flux is reduced by a factor of 3.

  13. 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

  14. Finite element analysis of inclined nozzle-plate junctions

    International Nuclear Information System (INIS)

    Dixit, K.B.; Seth, V.K.; Krishnan, A.; Ramamurthy, T.S.; Dattaguru, B.; Rao, A.K.

    1979-01-01

    Estimation of stress concentration at nozzle to plate or shell junctions is a significant problem in the stress analysis of nuclear reactors. The topic is a subject matter of extensive investigations and earlier considerable success has been reported on analysis for the cases when the nozzle is perpendicular to the plate or is radial to the shell. Analytical methods for the estimation of stress concentrations for the practical situations when the intersecting nozzle is inclined to the plate or is non-radial to the shell is rather scanty. Specific complications arise in dealing with the junction region when the nozzle with circular cross-section meets the non-circular cut-out on the plate or shell. In this paper a finite element analysis is developed for inclined nozzles and results are presented for nozzle-plate junctions. A method of analysis is developed with a view to achieving simultaneously accuracy of results and simplicity in the choice of elements and their connectivity. The circular nozzle is treated by axisymmetric conical shell elements. The nozzle portion in the region around the junction and the flat plate is dealt with by triangular flat shell elements. Special transition elements are developed for joining the flat shell elements with the axisymmetric elements under non-axisymmetric loading. A substructure method of analysis is adopted which achieves considerable economy in handling the structure and also conveniently combines the different types of elements in the structure. (orig.)

  15. Combustor nozzle for a fuel-flexible combustion system

    Science.gov (United States)

    Haynes, Joel Meier [Niskayuna, NY; Mosbacher, David Matthew [Cohoes, NY; Janssen, Jonathan Sebastian [Troy, NY; Iyer, Venkatraman Ananthakrishnan [Mason, OH

    2011-03-22

    A combustor nozzle is provided. The combustor nozzle includes a first fuel system configured to introduce a syngas fuel into a combustion chamber to enable lean premixed combustion within the combustion chamber and a second fuel system configured to introduce the syngas fuel, or a hydrocarbon fuel, or diluents, or combinations thereof into the combustion chamber to enable diffusion combustion within the combustion chamber.

  16. Analytical study of nozzle performance for nuclear thermal rockets

    International Nuclear Information System (INIS)

    Davidian, K.O.; Kacynski, K.J.

    1991-01-01

    Nuclear propulsion has been identified as one of the key technologies needed for human exploration of the Moon and Mars. The Nuclear Thermal Rocket (NTR) uses a nuclear reactor to heat hydrogen to a high temperature followed by expansion through a conventional convergent-divergent nozzle. A parametric study of NTR nozzles was performed using the Rocket Engine Design Expert System (REDES) at the NASA Lewis Research Center. The REDES used the JANNAF standard rigorous methodology to determine nozzle performance over a range of chamber temperatures, chamber pressures, thrust levels, and different nozzle configurations. A design condition was set by fixing the propulsion system exit radius at five meters and throat radius was varied to achieve a target thrust level. An adiabatic wall was assumed for the nozzle, and its length was assumed to be 80 percent of a 15 degree cone. The results conclude that although the performance of the NTR, based on infinite reaction rates, looks promising at low chamber pressures, finite rate chemical reactions will cause the actual performance to be considerably lower. Parameters which have a major influence on the delivered specific impulse value include the chamber temperature and the chamber pressures in the high thrust domain. Other parameters, such as 2-D and boundary layer effects, kinetic rates, and number of nozzles, affect the deliverable performance of an NTR nozzle to a lesser degree. For a single nozzle, maximum performance of 930 seconds and 1030 seconds occur at chamber temperatures of 2700 and 3100 K, respectively

  17. Noise from Aft Deck Exhaust Nozzles: Differences in Experimental Embodiments

    Science.gov (United States)

    Bridges, James E.

    2014-01-01

    Two embodiments of a rectangular nozzle on an aft deck are compared. In one embodiment the lower lip of the nozzle was extended with the sidewalls becoming triangles. In a second embodiment a rectangular nozzle was fitted with a surface that fit flush to the lower lip and extended outward from the sides of the nozzle, approximating a semi-infinite plane. For the purpose of scale-model testing, making the aft deck an integral part of the nozzle is possible for relatively short deck lengths, but a separate plate model is more flexible, accounts for the expanse of deck to the sides of the nozzle, and allows the nozzle to stand off from the deck. Both embodiments were tested and acoustic far-field results were compared. In both embodiments the extended deck introduces a new noise source, but the amplitude of the new source was dependent upon the span (cross-stream dimension) of the aft deck. The noise increased with deck length (streamwise dimension), and in the case of the beveled nozzle it increased with increasing aspect ratio. In previous studies of slot jets in wings it was noted that the increased noise from the extended aft deck appears as a dipole at the aft deck trailing edge, an acoustic source type with different dependence on velocity than jet mixing noise. The extraneous noise produced by the aft deck in the present studies also shows this behavior both in directivity and in velocity scaling.

  18. Numerical analysis of choked converging nozzle flows with surface ...

    Indian Academy of Sciences (India)

    Choked converging nozzle flow and heat transfer characteristics are numerically investigated by means of a recent computational model that integrates the axisymmetric continuity, state, momentum and energy equations. To predict the combined effects of nozzle geometry, friction and heat transfer rates, analyses are ...

  19. Design and Optimization of Aerospike nozzle using CFD

    Science.gov (United States)

    Naveen Kumar, K.; Gopalsamy, M.; Antony, Daniel; Krishnaraj, R.; Viswanadh, Chaparala B. V.

    2017-10-01

    New rocket designs are being adopted to increase the performance of the current satellite launch vehicles (SLVs). But, the aerospike (or plug) nozzle concept that has been under development since the 1950s is yet to be utilized on a launch platform. Due to its ability to adjust the environment by altering the outer jet boundary, the aerospike nozzle delivers better performance compared to present day bell nozzle. An aerospike nozzle is designed for 20 bar pressure ratio. In order to improve the performance of the aerospike nozzle for various conditions, optimization of the nozzle was carried out for some important design parameters and their performances were studied for cold flow conditions. Initially a model of an aerospike nozzle is created for certain parameters, then the optimization process is carried out for the nozzle (Truncated model & Base bleed model). Optimized model is designed by the software GAMBIT and the flow behaviour is analysed by the Computational Fluid Dynamics (CFD) software called FLUENT. Comparison also takes place between the full length and the optimized models.

  20. Numerical analysis of choked converging nozzle flows with surface ...

    Indian Academy of Sciences (India)

    Variation of discharge coefficients for sonic nozzles with flow geometry and Reynolds num- ber was reported by Paik et al (2000), who determined higher discharge coefficients with the increase of mass flow rate. Lear et al (1997) modelled dissipative effects of heat trans- fer on the exit kinetic energy and on nozzle efficiency ...

  1. Multi-orifice deposition nozzle for additive manufacturing

    Science.gov (United States)

    Lind, Randall F.; Post, Brian K.; Cini, Colin L.

    2017-11-21

    An additive manufacturing extrusion head includes a nozzle for accepting and depositing a heated material onto a work surface and/or part. The nozzle includes a valve body and an internal poppet body moveable between positions to permit deposition of at least two bead sizes of heated material onto a work surface and/or part.

  2. The Effect of Nozzle Trailing Edge Thickness on Jet Noise

    Science.gov (United States)

    Henderson, Brenda; Kinzie, Kevin; Haskin, Henry

    2004-01-01

    The effect of nozzle trailing edge thickness on broadband acoustic radiation and the production of tones is investigated for coannular nozzles. Experiments were performed for a core nozzle trailing edge thickness between 0.38 mm and 3.17 mm. The on-set of discrete tones was found to be predominantly affected by the velocity ratio, the ratio of the fan velocity to the core velocity, although some dependency on trailing edge thickness was also noted. For a core nozzle trailing edge thickness greater than or equal to 0.89 mm, tones were produced for velocity ratios between 0.91 and 1.61. For a constant nozzle trailing edge thickness, the frequency varied almost linearly with the core velocity. The Strouhal number based on the core velocity changed with nozzle trailing edge thickness and varied between 0.16 and 0.2 for the core nozzles used in the experiments. Increases in broadband noise with increasing trailing edge thickness were observed for tone producing and non-tone producing conditions. A variable thickness trailing edge (crenellated) nozzle resulted in no tonal production and a reduction of the broadband trailing edge noise relative to that of the corresponding constant thickness trailing edge.

  3. 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

  4. Thermographic Leak Detection of the Space Shuttle Main Engine Nozzle

    Science.gov (United States)

    Walker, James L.; Russell, Samuel S.

    1999-01-01

    The Space Shuttle Main Engines Nozzles consist of over one thousand tapered Inconel coolant tubes brazed to a stainless steel structural jacket. Liquid Hydrogen flows through the tubing, from the aft to forward end of the nozzle, under high pressure to maintain a thermal balance between the rocket exhaust and the nozzle wall. Three potential problems occur within the SSME nozzle coolant tubes as a result of manufacturing anomalies and the highly volatile service environment including poor or incomplete bonding of the tubes to the structural jacket, cold wall leaks and hot wall leaks. Of these conditions the identification of cold wall leaks has been the most problematic. The methods and results presented in this summary addresses the thermographic identification of cold wall "interstitial" leaks between the structural jacket and coolant tubes of the Space Shuttle Main Engines Nozzles.

  5. Study on steam pressure characteristics in various types of nozzles

    Science.gov (United States)

    Firman; Anshar, Muhammad

    2018-03-01

    Steam Jet Refrigeration (SJR) is one of the most widely applied technologies in the industry. The SJR system was utilizes residual steam from the steam generator and then flowed through the nozzle to a tank that was containing liquid. The nozzle converts the pressure energy into kinetic energy. Thus, it can evaporate the liquid briefly and release it to the condenser. The chilled water, was produced from the condenser, can be used to cool the product through a heat transfer process. This research aims to study the characteristics of vapor pressure in different types of nozzles using a simulation. The Simulation was performed using ANSYS FLUENT software for nozzle types such as convergent, convrgent-parallel, and convergent-divergent. The results of this study was presented the visualization of pressure in nozzles and was been validated with experiment data.

  6. Mechanical Behaviour of Lined Pipe

    NARCIS (Netherlands)

    Hilberink, A.

    2011-01-01

    Installing lined pipe by means of the reeling installation method seems to be an attractive combination, because it provides the opportunity of eliminating the demanding welds from the critical time offshore and instead preparing them onshore. However, reeling of lined pipe is not yet proven

  7. Residual stress determination in a dissimilar weld overlay pipe by neutron diffraction

    International Nuclear Information System (INIS)

    Woo, Wanchuck; Em, Vyacheslav; Hubbard, Camden R.; Lee, Ho-Jin; Park, Kwang Soo

    2011-01-01

    Highlights: → Determined residual stress distribution in a dissimilar weld overlay pipe. → Consists of a ferritic (SA508), austenitic (F316L) steels, Alloy 182 consumable. → Measured significant compression (-600 MPa) near the inner wall of overlay. → Validate integrity of the inner wall for the pressurized nozzle nuclear structure. - Abstract: Residual stresses were determined through the thickness of a dissimilar weld overlay pipe using neutron diffraction. The specimen has a complex joining structure consisting of a ferritic steel (SA508), austenitic steel (F316L), Ni-based consumable (Alloy 182), and overlay of Ni-base superalloy (Alloy 52M). It simulates pressurized nozzle components, which have been a critical issue under the severe crack condition of nuclear power reactors. Two neutron diffractometers with different spatial resolutions have been utilized on the identical specimen for comparison. The macroscopic 'stress-free' lattice spacing (d o ) was also obtained from both using a 2-mm width comb-like coupon. The results show significant changes in residual stresses from tension (300-400 MPa) to compression (-600 MPa) through the thickness of the dissimilar weld overlay pipe specimen.

  8. Residual stress determination in a dissimilar weld overlay pipe by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Wanchuck, E-mail: chuckwoo@kaeri.re.kr [Neutron Science Division, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Em, Vyacheslav [Neutron Science Division, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Hubbard, Camden R. [Materials Science and Technology Division, Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Lee, Ho-Jin [Nuclear Materials Research Center, Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Park, Kwang Soo [Corporate R and D Institute, Doosan Heavy Industries and Construction, Changwon 641-792 (Korea, Republic of)

    2011-10-15

    Highlights: {yields} Determined residual stress distribution in a dissimilar weld overlay pipe. {yields} Consists of a ferritic (SA508), austenitic (F316L) steels, Alloy 182 consumable. {yields} Measured significant compression (-600 MPa) near the inner wall of overlay. {yields} Validate integrity of the inner wall for the pressurized nozzle nuclear structure. - Abstract: Residual stresses were determined through the thickness of a dissimilar weld overlay pipe using neutron diffraction. The specimen has a complex joining structure consisting of a ferritic steel (SA508), austenitic steel (F316L), Ni-based consumable (Alloy 182), and overlay of Ni-base superalloy (Alloy 52M). It simulates pressurized nozzle components, which have been a critical issue under the severe crack condition of nuclear power reactors. Two neutron diffractometers with different spatial resolutions have been utilized on the identical specimen for comparison. The macroscopic 'stress-free' lattice spacing (d{sub o}) was also obtained from both using a 2-mm width comb-like coupon. The results show significant changes in residual stresses from tension (300-400 MPa) to compression (-600 MPa) through the thickness of the dissimilar weld overlay pipe specimen.

  9. Analysis, design and testing of high pressure waterjet nozzles

    Science.gov (United States)

    Mazzoleni, Andre P.

    1996-01-01

    The Hydroblast Research Cell at MSFC is both a research and a processing facility. The cell is used to investigate fundamental phenomena associated with waterjets as well as to clean hardware for various NASA and contractor projects. In the area of research, investigations are made regarding the use of high pressure waterjets to strip paint, grease, adhesive and thermal spray coatings from various substrates. Current industrial methods of cleaning often use ozone depleting chemicals (ODC) such as chlorinated solvents, and high pressure waterjet cleaning has proven to be a viable alternative. Standard methods of waterjet cleaning use hand held or robotically controlled nozzles. The nozzles used can be single-stream or multijet nozzles, and the multijet nozzles may be mounted in a rotating head or arranged in a fan-type shape. We consider in this paper the use of a rotating, multijet, high pressure water nozzle which is robotically controlled. This method enables rapid cleaning of a large area, but problems such as incomplete coverage (e.g. the formation of 'islands' of material not cleaned) and damage to the substrate from the waterjet have been observed. In addition, current stripping operations require the nozzle to be placed at a standoff distance of approximately 2 inches in order to achieve adequate performance. This close proximity of the nozzle to the target to be cleaned poses risks to the nozzle and the target in the event of robot error or the striking of unanticipated extrusions on the target surface as the nozzle sweeps past. Two key motivations of this research are to eliminate the formation of 'coating islands' and to increase the allowable standoff distance of the nozzle.

  10. Novel design for transparent high-pressure fuel injector nozzles

    Science.gov (United States)

    Falgout, Z.; Linne, M.

    2016-08-01

    The efficiency and emissions of internal combustion (IC) engines are closely tied to the formation of the combustible air-fuel mixture. Direct-injection engines have become more common due to their increased practical flexibility and efficiency, and sprays dominate mixture formation in these engines. Spray formation, or rather the transition from a cylindrical liquid jet to a field of isolated droplets, is not completely understood. However, it is known that nozzle orifice flow and cavitation have an important effect on the formation of fuel injector sprays, even if the exact details of this effect remain unknown. A number of studies in recent years have used injectors with optically transparent nozzles (OTN) to allow observation of the nozzle orifice flow. Our goal in this work is to design various OTN concepts that mimic the flow inside commercial injector nozzles, at realistic fuel pressures, and yet still allow access to the very near nozzle region of the spray so that interior flow structure can be correlated with primary breakup dynamics. This goal has not been achieved until now because interior structures can be very complex, and the most appropriate optical materials are brittle and easily fractured by realistic fuel pressures. An OTN design that achieves realistic injection pressures and grants visual access to the interior flow and spray formation will be explained in detail. The design uses an acrylic nozzle, which is ideal for imaging the interior flow. This nozzle is supported from the outside with sapphire clamps, which reduces tensile stresses in the nozzle and increases the nozzle's injection pressure capacity. An ensemble of nozzles were mechanically tested to prove this design concept.

  11. Adaptation of computer code ALMOD 3.4 for safety analyses of Westighouse type NPPs and calculation of main feedwater loss

    International Nuclear Information System (INIS)

    Kordis, I.; Jerele, A.; Brajak, F.

    1986-01-01

    The paper presents theoretical foundations of ALMOD 3.4 code and modification done in order to adjust the model to westinghouse type NPP. test cases for verification of added modules functioning were done and loss of main feedwater (FW) transient at nominal power was analysed. (author)

  12. Promethus Hot Leg Piping Concept

    International Nuclear Information System (INIS)

    AM Girbik; PA Dilorenzo

    2006-01-01

    The Naval Reactors Prime Contractor Team (NRPCT) recommended the development of a gas cooled reactor directly coupled to a Brayton energy conversion system as the Space Nuclear Power Plant (SNPP) for NASA's Project Prometheus. The section of piping between the reactor outlet and turbine inlet, designated as the hot leg piping, required unique design features to allow the use of a nickel superalloy rather than a refractory metal as the pressure boundary. The NRPCT evaluated a variety of hot leg piping concepts for performance relative to SNPP system parameters, manufacturability, material considerations, and comparison to past high temperature gas reactor (HTGR) practice. Manufacturability challenges and the impact of pressure drop and turbine entrance temperature reduction on cycle efficiency were discriminators between the piping concepts. This paper summarizes the NRPCT hot leg piping evaluation, presents the concept recommended, and summarizes developmental issues for the recommended concept

  13. Flexible ultrasonic pipe inspection apparatus

    Science.gov (United States)

    Jenkins, Charles F.; Howard, Boyd D.

    1998-01-01

    A flexible, modular ultrasonic pipe inspection apparatus, comprising a flexible, hollow shaft that carries a plurality of modules, including at least one rotatable ultrasonic transducer, a motor/gear unit, and a position/signal encoder. The modules are connected by flexible knuckle joints that allow each module of the apparatus to change its relative orientation with respect to a neighboring module, while the shaft protects electrical wiring from kinking or buckling while the apparatus moves around a tight corner. The apparatus is moved through a pipe by any suitable means, including a tether or drawstring attached to the nose or tail, differential hydraulic pressure, or a pipe pig. The rotational speed of the ultrasonic transducer and the forward velocity of the apparatus are coordinated so that the beam sweeps out the entire interior surface of the pipe, enabling the operator to accurately assess the condition of the pipe wall and determine whether or not leak-prone corrosion damage is present.

  14. Noise Prediction Module for Offset Stream Nozzles

    Science.gov (United States)

    Henderson, Brenda S.

    2011-01-01

    A Modern Design of Experiments (MDOE) analysis of data acquired for an offset stream technology was presented. The data acquisition and concept development were funded under a Supersonics NRA NNX07AC62A awarded to Dimitri Papamoschou at University of California, Irvine. The technology involved the introduction of airfoils in the fan stream of a bypass ratio (BPR) two nozzle system operated at transonic exhaust speeds. The vanes deflected the fan stream relative to the core stream and resulted in reduced sideline noise for polar angles in the peak jet noise direction. Noise prediction models were developed for a range of vane configurations. The models interface with an existing ANOPP module and can be used or future system level studies.

  15. Characterisation of inexpensive, simply shaped nozzles

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav

    2010-01-01

    Roč. 88, č. 11A (2010), s. 1433-1444 ISSN 0263-8762 R&D Projects: GA ČR GA101/07/1499; GA AV ČR IAA200760705 Institutional research plan: CEZ:AV0Z20760514 Keywords : nozzle * characteristic * separation of flow Subject RIV: BK - Fluid Dynamics Impact factor: 1.519, year: 2010 http://www.sciencedirect.com/science?_ob=MImg&_imagekey=B8JGF-4YPPRBF-3-2X&_cdi=43669&_user=640952&_pii=S0263876210001115&_origin=search&_coverDate=11%2F30%2F2010&_sk=999119988&view=c&wchp=dGLbVlW-zSkWb&md5=dbed1a6fea7702efd86e09264ff1a0e4&ie=/sdarticle.pdf

  16. Feedback mechanism for smart nozzles and nebulizers

    Science.gov (United States)

    Montaser, Akbar [Potomac, MD; Jorabchi, Kaveh [Arlington, VA; Kahen, Kaveh [Kleinburg, CA

    2009-01-27

    Nozzles and nebulizers able to produce aerosol with optimum and reproducible quality based on feedback information obtained using laser imaging techniques. Two laser-based imaging techniques based on particle image velocimetry (PTV) and optical patternation map and contrast size and velocity distributions for indirect and direct pneumatic nebulizations in plasma spectrometry. Two pulses from thin laser sheet with known time difference illuminate droplets flow field. Charge coupled device (CCL)) captures scattering of laser light from droplets, providing two instantaneous particle images. Pointwise cross-correlation of corresponding images yields two-dimensional velocity map of aerosol velocity field. For droplet size distribution studies, solution is doped with fluorescent dye and both laser induced florescence (LIF) and Mie scattering images are captured simultaneously by two CCDs with the same field of view. Ratio of LIF/Mie images provides relative droplet size information, then scaled by point calibration method via phase Doppler particle analyzer.

  17. Theoretical investigations of a viscous flow in rotational symmetry hollow jet nozzles with respect to a design of a flowing liquid metal target for a neutron spallation source

    International Nuclear Information System (INIS)

    Felsch, K.O.; Piesche, M.; Veith, W.

    1981-04-01

    The object of this theoretical study is the laminar and turbulent swirl free flow of a viscous incompressible medium in a rotation symmetric hollow jet nozzle whose geometrical configuration incorporates the technical conception of a molten metal target. Of interest is the construction of the nozzle in such a form that the wall boundaries reflect the natural frictional movement of the flow, i.e. the contours of the nozzle are trimmed by the interaction of the viscosity, momentum, gravity and surface tension forces. The mathematical treatment is based on an integral method. For laminar flow higher order polynomials were chosen and for turbulent flow the power of law of 1/7. As well as this the wall shear stresses in the turbulent flow region have to conform to the laws of pipe flow and in particular, to a modified form of Blasius' resistance law. The essential factors which are obtained from this study are the geometrical relationship between the average nozzle radius and the initial width of the fluid film, the exit angle and the Reynolds, Weber and Froude numbers as the characteristic geometric and physical flow parameters. (orig.) [de

  18. Waste pipe calculus extensions

    International Nuclear Information System (INIS)

    O'Connell, W.J.

    1979-01-01

    The waste pipe calculus provides a rapid method, using Laplace transforms, to calculate the transport of a pollutant such as nuclear waste, by a network of one-dimensional flow paths. The present note extends previous work as follows: (1) It provides an alternate approximation to the time-domain function (inverse Laplace transform) for the resulting transport. This algebraic approximation may be viewed as a simpler and more approximate model of the transport process. (2) It identifies two scalar quantities which may be used as summary consequence measures of the waste transport (or inversely, waste retention) system, and provides algebraic expressions for them. (3) It includes the effects of radioactive decay on the scalar quantity results, and further provides simplifying approximations for the cases of medium and long half-lives. This algebraic method can be used for quick approximate analyses of expected results, uncertainty and sensitivity, in evaluating selection and design choices for nuclear waste disposal systems

  19. Solar chemical heat pipe

    International Nuclear Information System (INIS)

    Levy, M.; Levitan, R.; Rosin, H.; Rubin, R.

    1991-08-01

    The performance of a solar chemical heat pipe was studied using CO 2 reforming of methane as a vehicle for storage and transport of solar energy. The endothermic reforming reaction was carried out in an Inconel reactor, packed with a Rh catalyst. The reactor was suspended in an insulated box receiver which was placed in the focal plane of the Schaeffer Solar Furnace of the Weizman Institute of Science. The exothermic methanation reaction was run in a 6-stage adiabatic reactor filled with the same Rh catalyst. Conversions of over 80% were achieved for both reactions. In the closed loop mode the products from the reformer and from the metanator were compressed into separate storage tanks. The two reactions were run either separately or 'on-line'. The complete process was repeated for over 60 cycles. The overall performance of the closed loop was quite satisfactory and scale-up work is in progress in the Solar Tower. (authors). 35 refs., 2 figs

  20. The linear VGA nozzle - a versatile tool for coal utilization

    Energy Technology Data Exchange (ETDEWEB)

    Walsh, W.A. Jr. [VGA Nozzle Co., Manchester, NH (United States)

    1993-12-31

    The newly available VGA nozzles provide significant improvements in a number of services and can advance the utilization of coal fuel in power plants and industrial processes. The nozzle designs for applications such as coal-water slurry combustion, flue gas cleaning, reburning and hot gas cooling, are described. VGA nozzles are patented as {open_quotes}Variable Gas Atomization,{close_quotes}. A conical configuration was first developed and successfully tested with heavy oil and coal-water mixture fuels at the Technical University of Nova Scotia. The test results showed the VGA nozzle to provide superior combustion characteristics at flow rates in excess of 1 gpm. The carbon burnout was complete, the total particulate emission was only 25% of that of competing nozzles, and there was a complete absence of wear of the nozzle tip and body components. A review is given of the 1980`s laboratory and field development/demonstration work and droplet particle size testing, previously reported at coal-slurry combustion and gas turbine conferences. Subsequently, a two phase S.B.I.R. (Small Business Innovative Research) program sponsored by DOE was recently completed by ADA Technologies, Inc., Englewood, CO, in which a production prototype linear VGA nozzle was developed for in-duct humidification of flue gases. As reported at the 1991 SO{sub 3} Control Symposium, December 3-6, Washington, DC, the nozzle achieves a 50% reduction in the energy consumption and lower capital, operating and maintenance costs. It is currently planned to market the linear VGA humidification nozzle as a cost-effective alternative to SO{sub 3} injection, for the conditioning of flue gas to achieve improved ESP performance.

  1. Temperature State of Noncooled Nozzle Adjutage of Liquid Rocket Engine

    Directory of Open Access Journals (Sweden)

    V. S. Zarubin

    2015-01-01

    Full Text Available The increasing specific impulse of the liquid rocket engine (LRE, which is designed to operate in space or in rarefied atmosphere, is directly related to the increasing speed of the combustion gases in the outlet section of the nozzle due to increasing nozzle expansion ratio. An intensity of the convective heat transfer of LRE combustion with the supersonic part of a nozzle shell in the first approximation is inversely proportional to the cross sectional area of gas dynamic path and reduces substantially as approaching to the outlet section of the nozzle.Therefore, in case of large nozzle expansion ratio the use of modern heat-resistant materials allows us to implement its outlet section as a thin-walled uncooled adjutage. This design solution results in reducing total weight of nozzle and decreasing overall preheat of LRE propellant used to cool the engine chamber. For a given diameter of the nozzle outlet section and pressure of combustion gases in this section, to make informed choices of permissible length for uncooled adjutage, it is necessary to have a reliable estimate of its thermal state on the steady-state LRE operation. A mathematical model of the nozzle shell heat transfer with the gas stream taking into account the heat energy transfer by convection and radiation, as well as by heat conduction along the generatrix of the shell enables this estimate.Quantitative analysis of given mathematical model showed that, because of the comparatively low pressure and temperature level of combustion gases, it is acceptable to ignore their own radiation and absorption capacity as compared with the convective heat intensity and the surface nozzle radiation. Thus, re-radiation of its internal surface portions is a factor of importance. Its taking into consideration is the main feature of the developed mathematical model.

  2. Aging and service wear of auxiliary feedwater pumps for PWR nuclear power plants. Volume 1. Operating experience and failure identification

    Energy Technology Data Exchange (ETDEWEB)

    Adams, M.L.; Makay, E.

    1986-07-01

    This report was produced under the Detection of Defects and Degradation Monitoring element of the Nuclear Plant Aging Research Program. Typical auxiliary feedwater pump (AUXFP) configurations are described in terms of configuration details, materials of construction, operating requirements, and modes of operation. AUXFP failure modes and causes due to aging and service wear are identified and explained, and measurable parameters (including functional indicators) for potential use in assessing operational readiness, establishing degradation trends, and detecting incipient failures are given. A series of measures to correct present deficiencies in surveillance, monitoring, and in-service testing practices is discussed. The main body of the report is supplemented by a number of relevant appendixes; in particular, a major appendix is included on engineering and design information useful to assess AUXFP operational readiness. 17 figs., 17 tabs.

  3. Trend and pattern analysis of failures of main feedwater system components in United States commercial nuclear power plants

    International Nuclear Information System (INIS)

    Gentillon, C.D.; Meachum, T.R.; Brady, B.M.

    1987-01-01

    The goal of the trend and pattern analysis of MFW (main feedwater) component failure data is to identify component attributes that are associated with relatively high incidences of failure. Manufacturer, valve type, and pump rotational speed are examples of component attributes under study; in addition, the pattern of failures among NPP units is studied. A series of statistical methods is applied to identify trends and patterns in failures and trends in occurrences in time with regard to these component attributes or variables. This process is followed by an engineering evaluation of the statistical results. In the remainder of this paper, the characteristics of the NPRDS that facilitate its use in reliability and risk studies are highlighted, the analysis methods are briefly described, and the lessons learned thus far for improving MFW system availability and reliability are summarized (orig./GL)

  4. Study by the disco method of critical components of a P.W.R. normal feedwater system

    International Nuclear Information System (INIS)

    Duchemin, B.; Villeneuve, M.J. de; Vallette, F.; Bruna, J.G.

    1983-03-01

    The DISCO (Determination of Importance Sensitivity of COmponents) method objectif is to rank the components of a system in order to obtain the most important ones versus availability. This method uses the fault tree description of the system and the cut set technique. It ranks the components by ordering the importances attributed to each one. The DISCO method was applied to the study of the 900 MWe P.W.R. normal feedwater system with insufficient flow in steam generator. In order to take account of operating experience several data banks were used and the results compared. This study allowed to determine the most critical component (the turbo-pumps) and to propose and quantify modifications of the system in order to improve its availability

  5. Application of the RCM (Reliability Centered Maintenance) approach to the optimization of emergency feedwater pump system maintenance

    International Nuclear Information System (INIS)

    Adamec, P.; Stvan, F.

    2001-12-01

    The major steps of the RCM analysis for the emergency feedwater pump system at the Dukovany NPP were as follows: Familiarization with the Dukovany maintenance process and strategy; Collection of information regarding data and information availability; system selection for the RCM study; implementation and use of suitable software; consulting the staff; FFA (Functional Failure Analysis) and FMEA (Failure Mode and Effect Analysis); Assessment of current periodical maintenance; and recommendations from the RCM analysis. The following Annexes are appended: A - Technological layouts; B - List of components; C - Information on functions; D - Critical components; E - FMEA; F - Consulting the staff (notes); G - Failure data (from RCM Workstation); H - Current periodical maintenance tasks; I - Selection of periodical maintenance tasks. (P.A.)

  6. Effusive atomic oven nozzle design using an aligned microcapillary array

    International Nuclear Information System (INIS)

    Senaratne, Ruwan; Rajagopal, Shankari V.; Geiger, Zachary A.; Fujiwara, Kurt M.; Lebedev, Vyacheslav; Weld, David M.

    2015-01-01

    We present a simple and inexpensive design for a multichannel effusive oven nozzle which provides improved atomic beam collimation and thus extended oven lifetimes. Using this design, we demonstrate an atomic lithium source suitable for trapped-atom experiments. At a nozzle temperature of 525 °C, the collimated atomic beam flux directly after the nozzle is 1.2 × 10 14 atoms/s with a peak beam intensity greater than 5.0 × 10 16 atoms/s/sr. This suggests an oven lifetime of several decades of continuous operation

  7. Fuel injector nozzle for an internal combustion engine

    Science.gov (United States)

    Cavanagh, Mark S.; Urven, Jr., Roger L.; Lawrence, Keith E.

    2008-11-04

    A direct injection fuel injector includes a nozzle tip having a plurality of passages allowing fluid communication between an inner nozzle tip surface portion and an outer nozzle tip surface portion and directly into a combustion chamber of an internal combustion engine. A first group of the passages have inner surface apertures located substantially in a first common plane. A second group of the passages have inner surface apertures located substantially in at least a second common plane substantially parallel to the first common plane. The second group has more passages than the first group.

  8. The fabrication of nozzles for nuclear components by welding

    International Nuclear Information System (INIS)

    Moraes, M.M.; Krausser, P.; Echeverria, J.A.V.

    1986-01-01

    A nozzle with medium outside diameter of 1000 mm and medium thickness of 150 mm composed integrally by deposited metal by submerged-arc (wire S3NiMo1, 0.5mm) was fabricated in NUCLEP. The nondestructive, mechanical, metallographic and chemical testing carried out in a test sample made by the same procedure and welding parameters, showed results according to specifications established for primary components for nuclear power plants, and the tests presented mechanical properties and tenacity better than similar nozzle samples. This nozzle is cheapest concerning to importations, in respecting to its forged similar. (M.C.K.) [pt

  9. Nuclear piping design, pipe support design and engineering during installation

    International Nuclear Information System (INIS)

    Podczerwinski, C.A.

    1983-01-01

    This paper discusses the computer-aided design of the piping and pipe supports of nuclear power plants. Hardware improvements have been made in the areas of man-machine communication, processing speed, and memory density. Topics considered include evolving design systems, application to current needs, safety-related small-bore piping and support design, as-built drawing review and reconciliation, large-bore pipe support design, snubber population reduction, and operating plant modifications. The improvements in man-machine communication hardware permit the designer to communicate with the computer in terms of pictures of elements of the design. The processing speed and memory density improvements enables the assembly of the design on the machine

  10. 49 CFR 192.55 - Steel pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Steel pipe. 192.55 Section 192.55 Transportation... BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS Materials § 192.55 Steel pipe. (a) New steel pipe is... in accordance with paragraph (c) or (d) of this section. (b) Used steel pipe is qualified for use...

  11. 49 CFR 192.59 - Plastic pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Plastic pipe. 192.59 Section 192.59 Transportation... BY PIPELINE: MINIMUM FEDERAL SAFETY STANDARDS Materials § 192.59 Plastic pipe. (a) New plastic pipe... specification; and (2) It is resistant to chemicals with which contact may be anticipated. (b) Used plastic pipe...

  12. 49 CFR 192.281 - Plastic pipe.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 3 2010-10-01 2010-10-01 false Plastic pipe. 192.281 Section 192.281... Plastic pipe. (a) General. A plastic pipe joint that is joined by solvent cement, adhesive, or heat fusion may not be disturbed until it has properly set. Plastic pipe may not be joined by a threaded joint or...

  13. Determination of the pipe stemming load

    International Nuclear Information System (INIS)

    Cowin, S.C.

    1979-01-01

    A mechanical model for the emplacement pipe system is developed. The model is then employed to determine the force applied to the surface collar of the emplacement pipe, the pipe-stemming load, and the stress along the emplacement pipe as a function of stemming height. These results are presented as integrals and a method for their numerical integration is given

  14. Large Eddy Simulation of Supercritical CO2 Through Bend Pipes

    Science.gov (United States)

    He, Xiaoliang; Apte, Sourabh; Dogan, Omer

    2017-11-01

    Supercritical Carbon Dioxide (sCO2) is investigated as working fluid for power generation in thermal solar, fossil energy and nuclear power plants at high pressures. Severe erosion has been observed in the sCO2 test loops, particularly in nozzles, turbine blades and pipe bends. It is hypothesized that complex flow features such as flow separation and property variations may lead to large oscillations in the wall shear stresses and result in material erosion. In this work, large eddy simulations are conducted at different Reynolds numbers (5000, 27,000 and 50,000) to investigate the effect of heat transfer in a 90 degree bend pipe with unit radius of curvature in order to identify the potential causes of the erosion. The simulation is first performed without heat transfer to validate the flow solver against available experimental and computational studies. Mean flow statistics, turbulent kinetic energy, shear stresses and wall force spectra are computed and compared with available experimental data. Formation of counter-rotating vortices, named Dean vortices, are observed. Secondary flow pattern and swirling-switching flow motions are identified and visualized. Effects of heat transfer on these flow phenomena are then investigated by applying a constant heat flux at the wall. DOE Fossil Energy Crosscutting Technology Research Program.

  15. Evaluation of fracture toughness of nuclear piping using real pipe and tensile compact pipe specimens

    International Nuclear Information System (INIS)

    Koo, J.M.; Park, S.; Seok, C.S.

    2013-01-01

    Highlights: • The tensile compact pipe (CP) specimen was proposed. • J-integral for the specimen was obtained by the plastic limit load analysis and FEA. • Fracture toughness tests by several types of specimens were performed and compared. • The constraint effects were considered by comparing Q-stresses for them. -- Abstract: The leak-before-break (LBB) concept is generally used to design the primary heat transport piping for a nuclear power plant. The LBB concept is based on the fracture resistance curve, which is obtained by J–R tests on various types of specimens. Fracture toughness data differ according to the various types of specimens. It has also been known that there is a difference in the constraint effect between real pipes and standard specimens, and LBB design using standard specimens is conservative. We propose a new type of specimen for J–R tests, a tensile compact pipe (CP) specimen, and perform fracture toughness tests on various types of specimens. We also perform constraint effect analysis on such specimens. The Q-stresses of the tensile CP specimens are lower than those of real pipes under 4-point bending, and are higher than those of elbow pipes. If the lever length of a tensile CP specimen is controlled, the specimen can simulate various stress conditions, and it is thought that the LBB design of piping in service can be performed using this specimen

  16. Evaluation of fracture toughness of nuclear piping using real pipe and tensile compact pipe specimens

    Energy Technology Data Exchange (ETDEWEB)

    Koo, J.M.; Park, S.; Seok, C.S., E-mail: seok@skku.edu

    2013-06-15

    Highlights: • The tensile compact pipe (CP) specimen was proposed. • J-integral for the specimen was obtained by the plastic limit load analysis and FEA. • Fracture toughness tests by several types of specimens were performed and compared. • The constraint effects were considered by comparing Q-stresses for them. -- Abstract: The leak-before-break (LBB) concept is generally used to design the primary heat transport piping for a nuclear power plant. The LBB concept is based on the fracture resistance curve, which is obtained by J–R tests on various types of specimens. Fracture toughness data differ according to the various types of specimens. It has also been known that there is a difference in the constraint effect between real pipes and standard specimens, and LBB design using standard specimens is conservative. We propose a new type of specimen for J–R tests, a tensile compact pipe (CP) specimen, and perform fracture toughness tests on various types of specimens. We also perform constraint effect analysis on such specimens. The Q-stresses of the tensile CP specimens are lower than those of real pipes under 4-point bending, and are higher than those of elbow pipes. If the lever length of a tensile CP specimen is controlled, the specimen can simulate various stress conditions, and it is thought that the LBB design of piping in service can be performed using this specimen.

  17. Selection of pipe repair methods.

    Science.gov (United States)

    2013-06-01

    The objective of this research is to provide pipeline operators with testing procedures and : results of the performance of composite pipe repair methods and ultimately, improve their : selection and installation, and reduce the risks associated with...

  18. Light pipes for LED measurements

    Science.gov (United States)

    Floyd, S. R.; Thomas, E. F., Jr.

    1976-01-01

    Light pipe directly couples LED optical output to single detector. Small area detector measures total optical output of diode. Technique eliminates thermal measurement problems and channels optical output to remote detector.

  19. Pulsating Heat Pipes, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Large radiator panels, based upon state of the art conventional heat pipes with attached fins for thermal load distribution and dissipation is the current baseline...

  20. Residual stress determination in an overlay dissimilar welded pipe by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Wan Chuck [ORNL; Em, Vyacheslav [Korea Atomic Energy Research Institute; Hubbard, Camden R [ORNL; Lee, Ho-Jin [Korea Atomic Energy Research Institute; Park, Kwang Soo [Doosan Heavy Industries & Construction

    2011-01-01

    Residual stresses were determined through the thickness of a dissimilar weld overlay pipe using neutron diffraction. The specimen has a complex joining structure consisting of a ferritic steel (SA508), austenitic steel (F316L), Ni-based consumable (Alloy 182), and overlay of Ni-base superalloy (Alloy 52M). It simulates pressurized nozzle components, which have been a critical issue under the severe crack condition of nuclear power reactors. Two neutron diffractometers with different spatial resolutions have been utilized on the identical specimen for comparison. The macroscopic 'stress-free' lattice spacing (d{sub o}) was also obtained from both using a 2-mm width comb-like coupon. The results show significant changes in residual stresses from tension (300-400 MPa) to compression (-600 MPa) through the thickness of the dissimilar weld overlay pipe specimen.

  1. Influence study of flow separation on the nozzle vibration response

    Directory of Open Access Journals (Sweden)

    Geng Li

    2016-06-01

    Full Text Available In the present paper, the vibration response difference of the upper stage nozzle with higher expansion ratio between ground and altitude simulation hot-firing test is analyzed. It indicates that the acceleration response of the nozzle under ground hot-firing test is much higher than that of the altitude condition. In order to find the essential reason, the experimental and numerical simulation studies of the flow separation are developed by using the test engine nozzle. The experimental data show that the nozzle internal flow occurred flow separation and the divergence cone internal wall pressure pulsation increased significantly downstream from the separation location. The numerical simulation and experimental results indicate that the increase of internal wall pressure and turbulence pulsating pressure are the substantial reason of vibration response increasing aggravatingly during the ground firing test.

  2. Optimal Thrust Vectoring for an Annular Aerospike Nozzle, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Recent success of an annular aerospike flight test by NASA Dryden has prompted keen interest in providing thrust vector capability to the annular aerospike nozzle...

  3. Altitude Compensating Nozzle Transonic Performance Flight Demonstration, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Altitude compensating nozzles continue to be of interest for use on future launch vehicle boosters and upper stages because of their higher mission average Isp and...

  4. Optimal Thrust Vectoring for an Annular Aerospike Nozzle Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Recent success of an annular aerospike flight test by NASA Dryden has prompted keen interest in providing thrust vector capability to the annular aerospike nozzle...

  5. Fabrication of Composite Combustion Chamber/Nozzle for Fastrac Engine

    Science.gov (United States)

    Lawrence, T.; Beshears, R.; Burlingame, S.; Peters, W.; Prince, M.; Suits, M.; Tillery, S.; Burns, L.; Kovach, M.; Roberts, K.

    2001-01-01

    The Fastrac Engine developed by the Marshall Space Flight Center for the X-34 vehicle began as a low cost engine development program for a small booster system. One of the key components to reducing the engine cost was the development of an inexpensive combustion chamber/nozzle. Fabrication of a regeneratively cooled thrust chamber and nozzle was considered too expensive and time consuming. In looking for an alternate design concept, the Space Shuttle's Reusable Solid Rocket Motor Project provided an extensive background with ablative composite materials in a combustion environment. An integral combustion chamber/nozzle was designed and fabricated with a silica/phenolic ablative liner and a carbon/epoxy structural overwrap. This paper describes the fabrication process and developmental hurdles overcome for the Fastrac engine one-piece composite combustion chamber/nozzle.

  6. Characterization of Plasmadynamics within a Small Magnetic Nozzle

    Data.gov (United States)

    National Aeronautics and Space Administration — This proposal presents an experimental and theoretical research project intended to develop a more refined model of the underlying physics of magnetic nozzles. The...

  7. An evaluation of nozzle afterbody code - AR02P

    Science.gov (United States)

    Guyton, F. C.

    1986-07-01

    A project was undertaken to develop a computational fluid dynamics (CFD) code for use in nozzle afterbody analysis. Objectives were to create a three-dimensional code capable of calculating afterbody flows with accuracy quantitatively close to the Navier-Stokes solutions, but which would use significantly fewer computer resources. The resulting program coupled an inverse boundary-layer routine with an Euler code and incorporated a jet plume. Calculations were made for the axisymmetric AGARD 15-deg boattail afterbody with variations in nozzle pressure ratio for Mach numbers 0.6 and 0.9, and compared with experimental results. The code predicted drag changes with NPR which showed the proper variations, but the code did not provide the accuracy required for typical nozzle afterbody analysis. (NPR = Nozzle total pressure to free stream static pressure ratio.)

  8. Fabrication, Cleaning, and Filtering of Microscopic Droplet Beam Nozzles

    Science.gov (United States)

    Warner, J.; Hunter, M.; Weierstall, U.; Spence, J. C. H.; Doak, R. B.

    2006-10-01

    Structure determination of proteins is a subject of intense current interest. Most relevant is a protein's native conformation, which generally requires it be immersed in water (if water-soluble) or a lipid jacket (if a membrane protein). Emerging schemes of serial protein diffraction propose to embed proteins in microscopic water droplets (membrane proteins encased in a detergent micelle) and pass these in vacuum through an x-ray or electron beam. Droplet diameters of tested, with and without sonication and both before and after the nozzle tip was formed. Flame burnishing was employed to smooth and clean the nozzles. In situ formation of silicate filter frits was investigated. Still, only about 30% of the 4 μm nozzles would run without clogging. An alternative to solid convergent nozzles will be described.

  9. Optimal Thrust Vectoring for an Annular Aerospike Nozzle, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Recent success of an annular aerospike flight test by NASA Dryden has prompted keen interest in providing thrust vector capability to the annular aerospike nozzle...

  10. Vibrational population distributions in nonequilibrium nozzle expansion flows

    Science.gov (United States)

    Watt, W. S.; Rich, J. W.

    1971-01-01

    Experimental measurements and theoretical calculations of the vibrational population distribution in nonequilibrium nozzle expansion flows of gas mixtures are reported. These studies were directed toward determining whether vibrational energy exchange pumping could lead to laser action on the vibrational bands of a diatomic molecule. Three different types of experiments were conducted. These showed (1) that vibrational energy was preferentially transferred from N2 to CO in supersonic nozzle flows containing these gases; (2) that under some conditions this vibrational energy exchange pumping mechanism created population inversions in the vibrational levels of CO; and (3) that at large expansion ratios the magnitude of these population inversions was sufficient to sustain lasing in the nozzle. A theoretical model was developed to calculate vibrational state population distributions in gas dynamic expansions of a mixture of diatomic gases. Although only isothermal calculations have been completed, these data indicate that population inversions are predicted for conditions similar to those obtained in the nozzle expansion flows.

  11. BOILING WATER REACTOR WITH FEED WATER INJECTION NOZZLES

    Science.gov (United States)

    Treshow, M.

    1963-04-30

    This patent covers the use of injection nozzles for pumping water into the lower ends of reactor fuel tubes in which water is converted directly to steam. Pumping water through fuel tubes of this type of boiling water reactor increases its power. The injection nozzles decrease the size of pump needed, because the pump handles only the water going through the nozzles, additional water being sucked into the tubes by the nozzles independently of the pump from the exterior body of water in which the fuel tubes are immersed. The resulting movement of exterior water along the tubes holds down steam formation, and thus maintains the moderator effectiveness, of the exterior body of water. (AEC)

  12. Fuel injection of coal slurry using vortex nozzles and valves

    Science.gov (United States)

    Holmes, Allen B.

    1989-01-01

    Injection of atomized coal slurry fuel into an engine combustion chamber is achieved at relatively low pressures by means of a vortex swirl nozzle. The outlet opening of the vortex nozzle is considerably larger than conventional nozzle outlets, thereby eliminating major sources of failure due to clogging by contaminants in the fuel. Control fluid, such as air, may be used to impart vorticity to the slurry and/or purge the nozzle of contaminants during the times between measured slurry charges. The measured slurry charges may be produced by a diaphragm pump or by vortex valves controlled by a separate control fluid. Fluidic circuitry, employing vortex valves to alternatively block and pass cool slurry fuel flow, is disclosed.

  13. Characterization of Accelerating Pipe Flow.

    Science.gov (United States)

    1988-03-01

    Journal of Chemical Engineering of Japan , vol...92, January 1970. 17. K. Kataoka, T. Kawabata, and K. Miki, "The Start-Up Response of Pipe Flow to a Step Change in Flow Rate," Journal of Chemical Engineering of Japan , vol...Pipe Flows," Journal of Chemical Engineering of Japan , vol. 9, no. 6, pp. 431-439, 1975. 19. E. van de Sande, A.P. Belde, B.J.G. Hamer, and W.

  14. 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

  15. Development of Intelligent pipe Locator

    Energy Technology Data Exchange (ETDEWEB)

    Miyamoto, Yukinori

    1987-09-10

    As the cities gets denser, development of a pipe locator, which can easily detect the position of the complicatedly buried underground pipes with high accuracy, is greatly demanded. A newly developed intelligent pipe locator detects a magnetic field generated by flowing an alternating-current through a buried pipe by means of a sensor unit placed on the ground and by computing and displaying with a micro-computer the position, depth, and the reliability of the detected result in only 3 seconds. Results of a half year field test since Sept. 1986 shows that the error was improved twice up to 5% from 10% of the conventional pipe locator. Operation time was also reduced down to one-fifth of the former method. By the practical use of this locator on the spot, one can expect the security improvements, reduction of frilling cost, and more efficient operation. In Japan recently, the number of conventional pipe locator is more than 5000 units in about 15 types. (10 figs, 7 tabs)

  16. Leaks in pipe networks

    Science.gov (United States)

    Pudar, Ranko S.; Liggett, James A.

    1992-01-01

    Leak detection in water-distribution systems can be accomplished by solving an inverse problem using measurements of pressure and/or flow. The problem is formulated with equivalent orifice areas of possible leaks as the unknowns. Minimization of the difference between measured and calculated heads produces a solution for the areas. The quality of the result depends on number and location of the measurements. A sensitivity matrix is key to deciding where to make measurements. Both location and magnitude of leaks are sensitive to the quantity and quality of pressure measurements and to how well the pipe friction parameters are known. The overdetermined problem (more measurements than suspected leaks) gives the best results, but some information can be derived from the underdetermined problem. The variance of leak areas, based on the quality of system characteristics and pressure data, indicates the likely accuracy of the results. The method will not substitute for more traditional leak surveys but can serve as a guide and supplement.

  17. Piping inspection round robin

    International Nuclear Information System (INIS)

    Heasler, P.G.; Doctor, S.R.

    1996-04-01

    The piping inspection round robin was conducted in 1981 at the Pacific Northwest National Laboratory (PNNL) to quantify the capability of ultrasonics for inservice inspection and to address some aspects of reliability for this type of nondestructive evaluation (NDE). The round robin measured the crack detection capabilities of seven field inspection teams who employed procedures that met or exceeded the 1977 edition through the 1978 addenda of the American Society of Mechanical Engineers (ASME) Section 11 Code requirements. Three different types of materials were employed in the study (cast stainless steel, clad ferritic, and wrought stainless steel), and two different types of flaws were implanted into the specimens (intergranular stress corrosion cracks (IGSCCs) and thermal fatigue cracks (TFCs)). When considering near-side inspection, far-side inspection, and false call rate, the overall performance was found to be best in clad ferritic, less effective in wrought stainless steel and the worst in cast stainless steel. Depth sizing performance showed little correlation with the true crack depths

  18. Reverse flow through a large scale multichannel nozzle

    International Nuclear Information System (INIS)

    Duignan, M.R.; Nash, C.A.

    1992-01-01

    A database was developed for the flow of water through a scaled nozzle of a Savannah River Site reactor inlet plenum. The water flow in the nozzle was such that it ranged from stratified to water solid conditions. Data on the entry of air into the nozzle and plenum as a function of water flow are of interest in loss-of-coolant studies. The scaled nozzle was 44 cm long, had an entrance diameter of 95 mm, an exit opening of 58 mm x 356 mm, and an exit hydraulic diameter approximately equal to that of the inlet. Within the nozzle were three flow-straightening vanes which divided the flow path into four channels. All data were taken at steady-state and isothermal (300 K ± 1.5 K) conditions. During the reverse flow of water through the nozzle the point at which air begins to enter was predicted within 90% by a critical weir-flow calculation. The point of air entry into the plenum itself was found to be a function of flow conditions

  19. Jet Noise Scaling in Dual Stream Nozzles

    Science.gov (United States)

    Khavaran, Abbas; Bridges, James

    2010-01-01

    Power spectral laws in dual stream jets are studied by considering such flows a superposition of appropriate single-stream coaxial jets. Noise generation in each mixing region is modeled using spectral power laws developed earlier for single stream jets as a function of jet temperature and observer angle. Similarity arguments indicate that jet noise in dual stream nozzles may be considered as a composite of four single stream jets representing primary/secondary, secondary/ambient, transition, and fully mixed zones. Frequency filter are designed to highlight spectral contribution from each jet. Predictions are provided at an area ratio of 2.0--bypass ratio from 0.80 to 3.40, and are compared with measurements within a wide range of velocity and temperature ratios. These models suggest that the low frequency noise in unheated jets is dominated by the fully mixed region at all velocity ratios, while the high frequency noise is dominated by the secondary when the velocity ratio is larger than 0.80. Transition and fully mixed jets equally dominate the low frequency noise in heated jets. At velocity ratios less than 0.50, the high frequency noise from primary/bypass becomes a significant contributing factor similar to that in the secondary/ambient jet.

  20. NEP heat pipe radiators. [Nuclear Electric Propulsion

    Science.gov (United States)

    Ernst, D. M.

    1979-01-01

    This paper covers improvements of heat pipe radiators for the thermionic NEP design. Liquid metal heat pipes are suitable as spacecraft radiator elements because of high thermal conductance, low mass and reliability, but the NEP thermionic system design was too large and difficult to fabricate. The current integral collector-radiator design consisting of several layers of thermionic converters, the annular-tangential collector heat pipe, the radiator heat pipe, and the transition zone designed to minimize the temperature difference between the collector heat pipe and radiator heat pipe are described. Finally, the design of micrometeoroid armor protection and the fabrication of the stainless steel annular heat pipe with a tangential arm are discussed, and it is concluded that the heat rejection system for the thermionic NEP system is well advanced, but the collector-radiator heat pipe transition and the 8 to 10 m radiator heat pipe with two bends require evaluation.

  1. An acoustic technique for tracing plastic pipe

    Energy Technology Data Exchange (ETDEWEB)

    Huebler, J.E.; Campbell, B.K. (Inst. of Gas Technology, Chicago, IL (United States)); Ching, G.K. (Southern California Gas Co., Los Angeles, CA (United States). Research Dept.)

    1993-08-01

    Many operation and maintenance activities performed by a gas distribution company require precise knowledge of the location of the gas main and/or service. These activities range from pipe location for the repair of a leak to the marking of pipe location as part of a one call system. Records provide one method of knowing the location of piping; however, these records are not always sufficiently accurate for field work. Thus, techniques for pipe location have always been an important need of the industry, and electromagnetic pipe locators have filled this need for years. Electromagnetic pipe locators, however, cannot find plastic pipe unless a tracer wire is buried next to or above the pipe. With the increased use of plastic pipe, a new technique for finding buried pipe is required. A successful acoustic plastic pipe locator could eliminate the use of tracer wire in new polyethylene pipe installations, thereby reducing pipe installation costs. Under sponsorship of the Southern California Gas Company, the Institute of Gas Technology (IGT) successfully demonstrated the proof-of-concept of an active acoustic plastic pipe location technique and is developing the technique into a practical field instrument.

  2. FEM Analysis and Measurement of Residual Stress by Neutron Diffraction on the Dissimilar Overlay Weld Pipe

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Soo; Lee, Ho Jin; Woo, Wan Chuck; Seong, Baek Seok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Byeon, Jin Gwi; Park, Kwang Soo; Jung, In Chul [Doosan Heavy Industries and Construction Co., Changwon (Korea, Republic of)

    2010-10-15

    Much research has been done to estimate the residual stress on a dissimilar metal weld. There are many methods to estimate the weld residual stress and FEM (Finite Element Method) is generally used due to the advantage of the parametric study. And the X-ray method and a Hole Drilling technique for an experimental method are also usually used. The aim of this paper is to develop the appropriate FEM model to estimate the residual stresses of the dissimilar overlay weld pipe. For this, firstly, the specimen of the dissimilar overlay weld pipe was manufactured. The SA 508 Gr3 nozzle, the SA 182 safe end and SA376 pipe were welded by the Alloy 182. And the overlay weld by the Alloy 52M was performed. The residual stress of this specimen was measured by using the Neutron Diffraction device in the HANARO (High-flux Advanced Neutron Application ReactOr) research reactor, KAERI (Korea Atomic Energy Research Institute). Secondly, FEM Model on the dissimilar overlay weld pipe was made and analyzed by the ABAQUS Code (ABAQUS, 2004). Thermal analysis and stress analysis were performed, and the residual stress was calculated. Thirdly, the results of the FEM analysis were compared with those of the experimental methods

  3. Three-dimensional analysis for liquid hydrogen in a cryogenic storage tank with heat pipe pump system

    Science.gov (United States)

    Ho, Son H.; Rahman, Muhammad M.

    2008-01-01

    This paper presents a study on fluid flow and heat transfer of liquid hydrogen in a zero boil-off cryogenic storage tank in a microgravity environment. The storage tank is equipped with an active cooling system consisting of a heat pipe and a pump-nozzle unit. The pump collects cryogen at its inlet and discharges it through its nozzle onto the evaporator section of the heat pipe in order to prevent the cryogen from boiling off due to the heat leaking through the tank wall from the surroundings. A three-dimensional (3-D) finite element model is employed in a set of numerical simulations to solve for velocity and temperature fields of liquid hydrogen in steady state. Complex structures of 3-D velocity and temperature distributions determined from the model are presented. Simulations with an axisymmetric model were also performed for comparison. Parametric study results from both models predict that as the speed of the cryogenic fluid discharged from the nozzle increases, the mean or bulk cryogenic fluid speed increases linearly and the maximum temperature within the cryogenic fluid decreases.

  4. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2

    International Nuclear Information System (INIS)

    Tijerina S, F.

    2008-01-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  5. LQG/LTR [linear quadratic Gaussian with loop transfer recovery] robust control system design for a low-pressure feedwater heater train

    International Nuclear Information System (INIS)

    Murphy, G.V.; Bailey, J.M.

    1990-01-01

    This paper uses the linear quadratic Gaussian with loop transfer recovery (LQG/LTR) control system design method to obtain a level control system for a low-pressure feedwater heater train. The control system performance and stability robustness are evaluated for a given set of system design specifications. The tools for analysis are the return ratio, return difference, and inverse return difference singular-valve plots for a loop break at the plant output. 3 refs., 7 figs., 2 tabs

  6. Analysis of water hammer phenomena - Application to deaerator-feedwater pump node

    International Nuclear Information System (INIS)

    Bigu, Melania; Tenescu, Mircea; Nita, Iulian Pavel

    2008-01-01

    The hydraulic hammer adverse effects are extensively presented in the literature available to those who operate and design installations in which this phenomenon occurs. There are specialized computational programs which evaluate diverse technical aspects which occur in this phenomenon. One must be noticed that not all the technical characteristics and not all effective operating modes which are treated in this paper are covered by existing computational programs. Moreover, even specialized developers of such programs recommend insistently that computational results offered by specialized programs to be verified by specialized technologists with experience in alternative theoretical computations in order to avoid any misinterpretation of results obtained by computational codes. After selective exposures of theoretical fundamentals of the problem there are presented a computational calculation obtained using the specialized calculation code PIPENET (Sunrise System Limited, Cambridge, Great Britain). The PIPENET calculation is compared with a standard computational calculation using theoretical correlations. An evaluation of the differences between those two computational methods is made in order to reveal the capabilities of the computational codes in solving the hydraulic hammering problems. In the first stage we obtained the elastic characteristics of the pipe where the phenomena of hydraulic hammering takes place. There are derivative descriptions of differential equations which describe the physical phenomena. In the second part we carried out a complete system analysis of water hammer effect due to a faulty closing of the four level control valves in steam generators. We compared the highest attended pressure with design pressure of the system. We observed that the design pressure is not overpassed. The analysis concluded that pumps' head protection is a very important parameter against overpressure in the feed water system. (authors)

  7. Fluid Structure Interaction in a Cold Flow Test and Transient CFD Analysis of Out-of-Round Nozzles

    Science.gov (United States)

    Ruf, Joseph; Brown, Andrew; McDaniels, David; Wang, Ten-See

    2010-01-01

    This viewgraph presentation describes two nozzle fluid flow interactions. They include: 1) Cold flow nozzle tests with fluid-structure interaction at nozzle separated flow; and 2) CFD analysis for nozzle flow and side loads of nozzle extensions with various out-of-round cases.

  8. 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.)

  9. 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

  10. Flexible ultrasonic pipe inspection apparatus

    Science.gov (United States)

    Jenkins, C.F.; Howard, B.D.

    1998-06-23

    A flexible, modular ultrasonic pipe inspection apparatus, comprises a flexible, hollow shaft that carries a plurality of modules, including at least one rotatable ultrasonic transducer, a motor/gear unit, and a position/signal encoder. The modules are connected by flexible knuckle joints that allow each module of the apparatus to change its relative orientation with respect to a neighboring module, while the shaft protects electrical wiring from kinking or buckling while the apparatus moves around a tight corner. The apparatus is moved through a pipe by any suitable means, including a tether or drawstring attached to the nose or tail, differential hydraulic pressure, or a pipe pig. The rotational speed of the ultrasonic transducer and the forward velocity of the apparatus are coordinated so that the beam sweeps out the entire interior surface of the pipe, enabling the operator to accurately assess the condition of the pipe wall and determine whether or not leak-prone corrosion damage is present. 7 figs.

  11. Flow induced vibrations of piping

    International Nuclear Information System (INIS)

    Gibert, R.J.; Axisa, F.

    1977-01-01

    This paper presents a method to take into account the inertial effects and the fluid compressibility 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-element 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 (acoustical sources or forces exciting directly the structure). Equations have been written in the hypothesis that acoustical wave lengths remain large compared to the diameter of the pipe. Indeed this is largely verified in almost practical cases. The method has been checked by an experiment performed on the GASCOGNE loop at D.E.M.T. The piping system under test consist 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 provide 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 [fr

  12. Feedwater flowrate estimation based on the two-step de-noising using the wavelet analysis and an autoassociative neural network

    International Nuclear Information System (INIS)

    Heo, Gyun Young; Choi, Seong Soo; Chang, Soon Heung

    1999-01-01

    This paper proposes an improved signal processing strategy for accurate feedwater flowrate estimation in nuclear power plants. It is generally known that ∼ 2% thermal power errors occur due to fouling phenomena in feedwater flowmeters. In the strategy proposed, the noises included in feedwater flowrate signal are classified into rapidly varying noises and gradually varying noises according to the characteristics in a frequency domain. The estimation precision is enhanced by introducing a low pass filter with the wavelet analysis against rapidly varying noises, and an autoassociative neural network which takes charge of the correction of only gradually varying noises. The modified multivariate stratification sampling using the concept of time stratification and MAXIMIN criteria is developed to overcome the shortcoming of a general random sampling. In addition the multi-stage robust training method is developed to increase the quality and reliability of training signals. Some validations using the simulated data from a micro-simulator were carried out. In the validation tests, the proposed methodology removed both rapidly varying noises and gradually varying noises respectively in each de-noising step, and 5.54 % root mean square errors of initial noisy signals were decreased to 0.674% after denoising. These results indicate that it is possible to estimate the reactor thermal power more elaborately by adopting this strategy. (author). 16 refs., 6 figs., 2 tabs

  13. Pipe Lines – External Corrosion

    Directory of Open Access Journals (Sweden)

    Dan Babor

    2008-01-01

    Full Text Available Two areas of corrosion occur in pipe lines: corrosion from the medium carried inside the pipes; corrosion attack upon the outside of the pipes (underground corrosion. Electrolytic processes are also involved in underground corrosion. Here the moisture content of the soil acts as an electrolyte, and the ions required to conduct the current are supplied by water-soluble salts (chlorides, sulfates, etc. present in the soil. The nature and amount of these soluble materials can vary within a wide range, which is seen from the varying electrical conductivity and pH (varies between 3 and 10. Therefore the characteristics of a soil will be an important factor in under-ground corrosion.

  14. Nuclear piping design - An overview

    International Nuclear Information System (INIS)

    Pattabiraman, J.; Neelwarne, A.

    1993-01-01

    Nuclear piping design is a continuously evolving process. Advances in analytical tools and the experience gained from the behaviour of structural systems under normal operating and extreme events like earthquake provide necessary inputs for refinement of design procedures/practices to achieve more economical and safe designs. Although, during last two decades considerable improvements in analytical tools were achieved, an overemphasis on providing conservatism in seismic design resulting in non optimal designs still remains. Present paper discusses aspects such as reliability and maintainability in the context of existing codal requirements of nuclear piping. The uncertainty associated with magnitude of seismic events, their damage potential in combination with other operational piping loads and lack of reliable data during 1970's were contributing factors for building conservatism in the seismic design rules/procedures. However, present status of research and experience on performance of above ground piping systems indicate that the damage potential of seismic event has been considerably overestimated and that overstiff designs have been adopted which have drawbacks in satisfying flexibility criteria for normal operations. The optimal design in the context of nuclear piping, therefore, should imply safe and trouble free operation throughout the life of plant with adequate margins provided for sustaining postulated seismic events. Though conformity with ASME code provides basic protection against piping failure, the issues related with reliability and maintainability, crucial for continued safe operation, can be best addressed through a rational design strategy. Such a design strategy should be based on careful evaluation of various loads related to their damage potential and impact on overall safety margins

  15. Drift-reducing nozzles and their biological efficacy.

    Science.gov (United States)

    Nuyttens, D; Dhoop, M; De Blauwer, V; Hermann, O; Hubrechts, W; Mestdagh, I; Dekeyser, D

    2009-01-01

    In 2007 and 2008, field trials were carried out with different standard and drift-reducing nozzles in sugar beet, maize, chicory, Belgian endive (all herbicide applications), wheat (fungicide application) and potatoes (Haulm killing herbicide application). The effect of nozzle type (standard flat fan, low-drift flat fan, air injection), nozzle size (ISO 02, 03 and 04) and application volume on the biological efficacy was investigated. All applications were done using a plot sprayer with volume rates ranging from 160 to 320 l.ha(-1) at recommended dose rates with commonly used (mix of) plant protection products. For each crop, the experiments included four replicates in a randomized block design. Depending on the type of application, the efficacy was measured in terms of weed control, disease and yield level, percentage dead leaf and stem, etc. In a previous research, drift and droplet characteristics of the different techniques were measured. In general no important effect of application technique on biological efficacy was observed for the tested herbicide and fungicide applications within the interval of volume rates and droplet size tested. Drift-reducing nozzles performed similar as conventional nozzles under good spraying conditions and using a correct spray application technique.

  16. Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

    International Nuclear Information System (INIS)

    Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae

    2016-01-01

    A Containment Filtered Venting System (CFVS) is the most interested device to mitigate a threat against containment integrity under the severe accident of nuclear power plant by venting with the filtration of the fission products. FNC technology and partners have been developed the self-priming scrubbing nozzle used for the CFVS which is based on the venturi effect. The thermal-hydraulic performances such as passive scrubbing water suction as well as pressure drop across the nozzle have been tested under various thermal-hydraulic conditions. The two types of test section have been built for testing the thermal-hydraulic performance of the self-priming scrubbing nozzle. Through the visualization loop, the liquid suction performance through the slit, pressure drop across the nozzle are measured. The passive water suction flow through the suction slit at the throat is important parameter to define the scrubbing performance of the self-priming scrubbing nozzle. The water suction flow is increased with the increase of the overhead water level at the same inlet gas flow. It is not so much changed with the change of inlet gas flow at the overhead water level.

  17. System for installing a steam generator nozzle dam

    International Nuclear Information System (INIS)

    McDonald, F.X.; Weisel, E.M.; Schukei, G.E.

    1991-01-01

    This patent describes a system for installing a nozzle dam in a nuclear steam generator having a head including a head internal surface, a manway penetrating the head, and a nozzle penetrating the head. It comprises a manipulator adapted to be passed through the manway and having one end adapted to be attached remotely to the head internal surface and a free end including a clamp member; nozzle dam segments, each segment sized to pass through the manway and having means thereon for engaging at least one other segment, the segments when fully engaged to each other forming a dam subassembly sized to pass into and seat against the nozzle; and means for controlling the manipulator while the one end is attached to the head internal surface, such that the clamp member grasps and supports one of the dam segments within the head until the subassembly is formed within the head, and then translates the dam subassembly within the head until the dam subassembly seats within the nozzle

  18. High performance flexible heat pipes

    Science.gov (United States)

    Shaubach, R. M.; Gernert, N. J.

    1985-01-01

    A Phase I SBIR NASA program for developing and demonstrating high-performance flexible heat pipes for use in the thermal management of spacecraft is examined. The program combines several technologies such as flexible screen arteries and high-performance circumferential distribution wicks within an envelope which is flexible in the adiabatic heat transport zone. The first six months of work during which the Phase I contract goal were met, are described. Consideration is given to the heat-pipe performance requirements. A preliminary evaluation shows that the power requirement for Phase II of the program is 30.5 kilowatt meters at an operating temperature from 0 to 100 C.

  19. Seismic design of piping systems

    International Nuclear Information System (INIS)

    Anglaret, G.; Beguin, J.L.

    1986-01-01

    This paper deals with the method used in France for the PWR nuclear plants to derive locations and types of supports of auxiliary and secondary piping systems taking earthquake in account. The successive steps of design are described, then the seismic computation method and its particular conditions of applications for piping are presented. The different types of support (and especially seismic ones) are described and also their conditions of installation. The method used to compare functional tests results and computation results in order to control models is mentioned. Some experiments realised on site or in laboratory, in order to validate models and methods, are presented [fr

  20. Laboratory exercises on oscillation modes of pipes

    Science.gov (United States)

    Haeberli, Willy

    2009-03-01

    This paper describes an improved lab setup to study the vibrations of air columns in pipes. Features of the setup include transparent pipes which reveal the position of a movable microphone inside the pipe; excitation of pipe modes with a miniature microphone placed to allow access to the microphone stem for open, closed, or conical pipes; and sound insulation to avoid interference between different setups in a student lab. The suggested experiments on the modes of open, closed, and conical pipes, the transient response of a pipe, and the effect of pipe diameter are suitable for introductory physics laboratories, including laboratories for nonscience majors and music students, and for more advanced undergraduate laboratories. For honors students or for advanced laboratory exercises, the quantitative relation between the resonance width and damping time constant is of interest.

  1. Corrosion of Spiral Rib Aluminized Pipe

    Science.gov (United States)

    2012-08-01

    Large diameter, corrugated steel pipes are a common sight in the culverts that run alongside many Florida roads. Spiral-ribbed aluminized pipe (SRAP) has been widely specified by the Florida Department of Transportation (FDOT) for runoff drainage. Th...

  2. Corrosion of Spiral Rib Aluminized Pipe : [Summary

    Science.gov (United States)

    2012-01-01

    Large diameter, corrugated steel pipes are a common sight in the culverts that run alongside many Florida roads. Spiral-ribbed aluminized pipe (SRAP) has been widely specified by the Florida Department of Transportation (FDOT) for runoff drainage. Th...

  3. Review of nuclear piping seismic design requirements

    International Nuclear Information System (INIS)

    Slagis, G.C.; Moore, S.E.

    1994-01-01

    Modern-day nuclear plant piping systems are designed with a large number of seismic supports and snubbers that may be detrimental to plant reliability. Experimental tests have demonstrated the inherent ruggedness of ductile steel piping for seismic loading. Present methods to predict seismic loads on piping are based on linear-elastic analysis methods with low damping. These methods overpredict the seismic response of ductile steel pipe. Section III of the ASME Boiler and Pressure Vessel Code stresses limits for piping systems that are based on considerations of static loads and hence are overly conservative. Appropriate stress limits for seismic loads on piping should be incorporated into the code to allow more flexible piping designs. The existing requirements and methods for seismic design of piping systems, including inherent conservations, are explained to provide a technical foundation for modifications to those requirements. 30 refs., 5 figs., 3 tabs

  4. 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

  5. Performance evaluation of buried pipe installation.

    Science.gov (United States)

    2010-05-01

    The purpose of this study is to determine the effects of geometric and mechanical parameters characterizing the soil structure interaction developed in a buried pipe installation located under roads/highways. The drainage pipes or culverts installed ...

  6. The analysis of the functional role of man and machine in the control of a notional auxiliary feedwater system

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Codazzi, A.; Decortis, F.

    1991-01-01

    We will describe here the simulation of a moderately complex plant, i.e. the Auxiliary Feedwater System (AFWS) of a nuclear power plant, which has been developed for interacting with a cognitive model of operator in a simulation framework of man-machine system studies as well as with an external operator for verifying and validating the hypotheses of the theoretical model by experimental studies. In order to develop such simulation, which must be very flexible for satisfying the needs of interaction with an operator as well as with a cognitive model, a number of special conditions have been respected: the model of functional behaviour of the system has been extended to include the logic of control mechanisms, i.e. components, indicators and actuators; the control tasks for a number of sequences has been developed; the robustness of physical model has been tested in whole possible configuration of the plant; and finally, the interface of the simulation with the model for dynamic failures of components has also been granted. In this paper, these aspects of the deterministic model of the AFWS will be firstly presented in detail. Then, the interface of the plant simulation with an external user or with the cognitive model of the operator will be described focusing on the analysis of the control task. Finally, we will attempt to integrate our approach in an overall framework of taxonomy for studying human actions in complex work context

  7. Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes, E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

  8. LOFT/LP-FW-1, Loss of Fluid Test, PWR Response to Loss-of-Feedwater Transient

    International Nuclear Information System (INIS)

    1989-01-01

    1 - Description of test facility: The LOFT Integral Test Facility is a scale model of a LPWR. The intent of the facility is to model the nuclear, thermal-hydraulic phenomena which would take place in a LPWR during a LOCA. The general philosophy in scaling coolant volumes and flow areas in LOFT was to use the ratio of the LOFT core [50 MW(t)] to a typical LPWR core [3000 MW(t)]. For some components, this factor is not applied; however, it is used as extensively as practical. In general, components used in LOFT are similar in design to those of a LPWR. Because of scaling and component design, the LOFT LOCA is expected to closely model a LPWR LOCA. 2 - Description of test: The first OECD LOFT experiment was conducted on February 20, 1983. It was designed to evaluate the generic PWR system response during a complete loss-of-feedwater transient. The objective of the experiment was to investigate the performance of primary 'feed and bleed' using a 'bleed' from the PORV and 'feed' from the HPIS to provide decay heat removal and system pressure reduction while maintaining the primary coolant inventory. 3 - Experimental limitations or shortcomings: Short core and steam generator, excessive core bypass, other scaling compromises, and lack of adequate measurements in certain areas

  9. Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

  10. Parallel island genetic algorithm applied to a nuclear power plant auxiliary feedwater system surveillance tests policy optimization

    International Nuclear Information System (INIS)

    Pereira, Claudio M.N.A.; Lapa, Celso M.F.

    2003-01-01

    In this work, we focus the application of an Island Genetic Algorithm (IGA), a coarse-grained parallel genetic algorithm (PGA) model, to a Nuclear Power Plant (NPP) Auxiliary Feedwater System (AFWS) surveillance tests policy optimization. Here, the main objective is to outline, by means of comparisons, the advantages of the IGA over the simple (non-parallel) genetic algorithm (GA), which has been successfully applied in the solution of such kind of problem. The goal of the optimization is to maximize the system's average availability for a given period of time, considering realistic features such as: i) aging effects on standby components during the tests; ii) revealing failures in the tests implies on corrective maintenance, increasing outage times; iii) components have distinct test parameters (outage time, aging factors, etc.) and iv) tests are not necessarily periodic. In our experiments, which were made in a cluster comprised by 8 1-GHz personal computers, we could clearly observe gains not only in the computational time, which reduced linearly with the number of computers, but in the optimization outcome

  11. Optimization of the pumping ring in a mechanical seal with an integrated cooler for feed-water pumps

    International Nuclear Information System (INIS)

    Buchdahl, D.; Martin, R.; Gueret, G.; Blanc, M.

    1994-07-01

    To simplify maintenance, E.D.F. along with its collaborators undertook the study of mechanical seal with integrated cooler used in feed-water pumps in the nuclear power plants. The cooler, integrated to the pump acts as a thermal barrier as well as a cooler of the mechanical seal. The water circulation in the cooler is assumed by an integrated pumping ring in the rotary part of the mechanical seal, with a matching screw thread in the pumping case. This assembly of mechanical seal/integrated cooler is tested in a test loop at the EDF/DER Laboratory. All working conditions are similar to that at site. Tests with different configurations of the rotor/stator profiles are performed, i.e.; different lengths and types of threading. Hydraulic performances and the global thermal balance of this assembly are studied. Our basic aim during these tests is to optimize the hydraulic performance of the pumping ring so as to best cool the mechanical seal faces. The different results obtained and the conclusions drawn during these tests are presented. (authors). 7 figs., 3 refs

  12. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  13. Analysis of Municipal Pipe Network Franchise Institution

    Science.gov (United States)

    Yong, Sun; Haichuan, Tian; Feng, Xu; Huixia, Zhou

    Franchise institution of municipal pipe network has some particularity due to the characteristic of itself. According to the exposition of Chinese municipal pipe network industry franchise institution, the article investigates the necessity of implementing municipal pipe network franchise institution in China, the role of government in the process and so on. And this offers support for the successful implementation of municipal pipe network franchise institution in China.

  14. 46 CFR 45.133 - Air pipes.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Air pipes. 45.133 Section 45.133 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) LOAD LINES GREAT LAKES LOAD LINES Conditions of Assignment § 45.133 Air pipes. (a) Where an air pipe to any tank extends above the freeboard or superstructure deck— (1) The exposed part of the air pipe must...

  15. Seismic analysis of nuclear piping system

    International Nuclear Information System (INIS)

    Shrivastava, S.K.; Pillai, K.R.V.; Nandakumar, S.

    1975-01-01

    To illustrate seismic analysis of nuclear power plant piping, a simple piping system running between two floors of the reactor building is assumed. Reactor building floor response is derived from time-history method. El Centre earthquake (1940) accelerogram is used for time-history analysis. The piping system is analysed as multimass lumped system. Behaviour of the pipe during the said earthquake is discussed. (author)

  16. 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

  17. A finite element approach for predicting nozzle admittances

    Science.gov (United States)

    Sigman, R. K.; Zinn, B. T.

    1983-01-01

    A finite element method is used to predict the admittances of axisymmetric nozzles. It is assumed that the flow in the nozzle is isentropic and the disturbances are small so that linear analyses apply. An approximate, two dimensional compressible model is used to describe the steady flow in the nozzle. The propagation of acoustic disturbances is governed by the complete linear wave equation. The differential form of the acoustic equation is transformed to an integral equation by using Galerkin's method, and Green's theorem is applied so that the acoustic boundary conditions can be introduced through the boundary residuals. The boundary conditions are described for both straight and curved sonic lines. A two dimensional FEM with linear elements is used to solve the acoustic equation. A one dimensional FEM is also used to solve the reduced equation of Crocco, and the solution verifies the sufficiency of the boundary residual formulation. Comparison between computed admittances and experimental data is shown to be quite good.

  18. Theoretical determination of nozzle admittances using a finite element approach

    Science.gov (United States)

    Sigman, R. K.; Zinn, B. T.

    1980-01-01

    A finite element method is used to predict the admittances of axisymmetric nozzles. It is assumed that the flow in the nozzle is isentropic and irrotational, and the disturbances are small so that linear analyses apply. An approximate, two dimensional compressible model is used to describe the steady flow in the nozzle. The propagation of acoustic disturbances is governed by the complete linear wave equation. The differential form of the acoustic equation is transformed to an integral equation using Galerkin's method, and Green's theorem is applied so that the acoustic boundary conditions can be introduced through the boundary residuals. A two-dimensional FEM using linear elements is used to solve the acoustic equation. A one dimensional FEM is also used to solve the reduced equation of Crocco, and the solution verifies the sufficiency of the boundary residual formulation. Comparison between computed admittances and experimental data is shown to be quite good.

  19. Ice Control with Brine Spread with Nozzles on Highways

    DEFF Research Database (Denmark)

    Bolet, Lars; Fonnesbech, Jens Kristian

    2010-01-01

    During the years 1996-2006, the former county of Funen, Denmark, gradually replaced pre-wetted salt with brine spread with nozzles as anti-icing agent in all her ice control activities. The replacement related to 1000 kilometres of highways. Jeopardizing neither road safety nor traffic flow...... spreading on a highway with traffic. A total of 800 spots were measured for residual salt for every spreader. The measurements and the spread pattern for brine spreading with nozzles were so precisely, that we learned: “When there is moisture, water or ice on the road, we need to take into account...... that the salt will run from the high level of the road to the lower level”. In the test the salt moved 1 meter in 3 hours. The knowledge gained from the measurements in the county of Funen - brine spread with nozzles, spreading salt to high level of the road and using GPS controlled spreading – was implemented...

  20. Top-nozzle mounted replacement guide pin assemblies

    International Nuclear Information System (INIS)

    Gilmore, C.B.; Andrews, W.H.

    1993-01-01

    A replacement guide pin assembly is provided for aligning a nuclear fuel assembly with an upper core plate of a nuclear reactor core. The guide pin assembly includes a guide pin body having a radially expandable base insertable within a hole in the top nozzle, a ferrule insertable within the guide pin base and capable of imparting a radially and outwardly directed force on the expandable base to expand it within the hole of the top nozzle and thereby secure the guide pin body to the top nozzle in response to a predetermined displacement of the ferrule relative to the guide pin body along its longitudinal axis, and a lock screw interfitted with the ferrule and threaded into the guide pin body so as to produce the predetermined displacement of the ferrule. (author)

  1. Effect of nozzle geometry for swirl type twin-fluid water mist nozzle on the spray characteristic

    International Nuclear Information System (INIS)

    Yoon, Soon Hyun; Kim, Do Yeon; Kim, Dong Keon; Kim, Bong Hwan

    2011-01-01

    Experimental investigations on the atomization characteristics of twin-fluid water mist nozzle were conducted using particle image velocimetry (PIV) system and particle motion analysis system (PMAS). The twin-fluid water mist nozzles with swirlers designed two types of swirl angles such as 0 .deg. , 90 .deg. and three different size nozzle hole diameters such as 0.5mm, 1mm, 1.5mm were employed. The experiments were carried out by the injection pressure of water and air divided into 1bar, 2bar respectively. The droplet size of the spray was measured using PMAS. The velocity and turbulence intensity were measured using PIV. The velocity, turbulence intensity and SMD distributions of the sprays were measured along the centerline and radial direction. As the experimental results, swirl angle controlled to droplet sizes. It was found that SMD distribution decreases with the increase of swirl angle. The developed twin-fluid water mist nozzle was satisfied to the criteria of NFPA 750, Class 1. It was proven that the developed nozzle under low pressures could be applied to fire protection system

  2. Nuclear class 1 piping stress analysis

    International Nuclear Information System (INIS)

    Lucas, J.C.R.; Maneschy, J.E.; Mariano, L.A.; Tamura, M.

    1981-01-01

    A nuclear class 1 piping stress analysis, according to the ASME code, is presented. The TRHEAT computer code has been used to determine the piping wall thermal gradient. The Nupipe computer code was employed for the piping stress analysis. Computer results were compared with the allowable criteria from the ASME code. (Author) [pt

  3. 49 CFR 236.712 - Brake pipe.

    Science.gov (United States)

    2010-10-01

    ... OF SIGNAL AND TRAIN CONTROL SYSTEMS, DEVICES, AND APPLIANCES Definitions § 236.712 Brake pipe. A pipe running from the engineman's brake valve through the train, used for the transmission of air under... 49 Transportation 4 2010-10-01 2010-10-01 false Brake pipe. 236.712 Section 236.712 Transportation...

  4. 46 CFR 197.336 - Pressure piping.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 7 2010-10-01 2010-10-01 false Pressure piping. 197.336 Section 197.336 Shipping COAST... GENERAL PROVISIONS Commercial Diving Operations Equipment § 197.336 Pressure piping. Piping systems that... the pressure boundaries as set forth in § 197.462. ...

  5. Smoking water pipe is injurious to lungs

    DEFF Research Database (Denmark)

    Sivapalan, Pradeesh; Ringbæk, Thomas; Lange, Peter

    2014-01-01

    This review describes the pulmonary consequences of water pipe smoking. Smoking water pipe affects the lung function negatively, is significantly associated with chronic obstructive pulmonary disease and increases the risk of lung infections. Case reports suggest that regular smokers of water pipe...

  6. The locating ways of laying pipe manipulator

    Science.gov (United States)

    Wang, Dan; Li, Bin; Lei, DongLiang

    2010-01-01

    The laying pipe manipulator is a new equipment to lay concrete pipe. This kind of manipulator makes the work of laying pipes mechanized and automated. We report here a new laying pipe manipulator. The manipulator has 5 free degrees, and is driven by the hydraulic system. In the paper, one critical question of manipulator is studied: the locating ways of the manipulator to lay concrete pipe. During the process of laying concrete pipe, how to locate the manipulator is realized by the locating system of manipulator. The locating system consists of photoelectric target, laser producer, and computer. According to different construction condition, one or two or three photoelectric targets can be used. During the process of laying concrete pipe, if the interface of pipes are jointed together, and the other segment of pipe deviates from the pipe way, one target can be used, if the angle that the manipulator rotates around the holding pipe's axes is 0°, two targets can be used, three targets can be used at any site. In the paper, according to each locating way, the theory analysis is done. And the mathematical models of the manipulator moving from original position to goal position are obtained by different locating way. And the locating experiment was done. According to the experiment result, the work principle and mathematical models of different locating way was turned out to be well adopted for requirement, the mathematical model of different locating way supplies the basic control theory for the manipulator to lay and joint concrete pipe automatically.

  7. Development of Flow and Heat Transfer Models for the Carbon Fiber Rope in Nozzle Joints of the Space Shuttle Reusable Solid Rocket Motor

    Science.gov (United States)

    Wang, Q.; Ewing, M. E.; Mathias, E. C.; Heman, J.; Smith, C.; McCool, Alex (Technical Monitor)

    2001-01-01

    Methodologies have been developed for modeling both gas dynamics and heat transfer inside the carbon fiber rope (CFR) for applications in the space shuttle reusable solid rocket motor joints. Specifically, the CFR is modeled using an equivalent rectangular duct with a cross-section area, friction factor and heat transfer coefficient such that this duct has the same amount of mass flow rate, pressure drop, and heat transfer rate as the CFR. An equation for the friction factor is derived based on the Darcy-Forschheimer law and the heat transfer coefficient is obtained from pipe flow correlations. The pressure, temperature and velocity of the gas inside the CFR are calculated using the one-dimensional Navier-Stokes equations. Various subscale tests, both cold flow and hot flow, have been carried out to validate and refine this CFR model. In particular, the following three types of testing were used: (1) cold flow in a RSRM nozzle-to-case joint geometry, (2) cold flow in a RSRM nozzle joint No. 2 geometry, and (3) hot flow in a RSRM nozzle joint environment simulator. The predicted pressure and temperature history are compared with experimental measurements. The effects of various input parameters for the model are discussed in detail.

  8. RSRM Nozzle-to-Case Joint J-leg Development

    Science.gov (United States)

    Albrechtsen, Kevin U.; Eddy, Norman F.; Ewing, Mark E.; McGuire, John R.

    2003-01-01

    Since the beginning of the Space Shuttle Reusable Solid Rocket Motor (RSRM) program, nozzle-to-case joint polysulfide adhesive gas paths have occurred on several flight motors. These gas paths have allowed hot motor gases to reach the wiper O-ring. Even though these motors continue to fly safely with this condition, a desire was to reduce such occurrences. The RSRM currently uses a J-leg joint configuration on case field joints and igniter inner and outer joints. The J-leg joint configuration has been successfully demonstrated on numerous RSRM flight and static test motors, eliminating hot gas intrusion to the critical O-ring seals on these joints. Using the proven technology demonstrated on the case field joints and igniter joints, a nozzle-to-case joint J-leg design was developed for implementation on RSRM flight motors. This configuration provides an interference fit with nozzle fixed housing phenolics at assembly, with a series of pressurization gaps incorporated outboard of the joint mating surface to aid in joint pressurization and to eliminate any circumferential flow in this region. The joint insulation is bonded to the nozzle phenolics using the same pressure sensitive adhesive used in the case field joints and igniter joints. An enhancement to the nozzle-to-case joint J-leg configuration is the implementation of a carbon rope thermal barrier. The thermal barrier is located downstream of the joint bondline and is positioned within the joint in a manner where any hot gas intrusion into the joint passes through the thermal barrier, reducing gas temperatures to a level that would not affect O-rings downstream of the thermal barrier. This paper discusses the processes used in reaching a final nozzle-to-case joint J-leg design, provides structural and thermal results in support of the design, and identifies fabrication techniques and demonstrations used in arriving at the final configuration.

  9. PPOOLEX experiments with two parallel blowdown pipes

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-01-15

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

  10. Pipe Leak Detection Technology Development

    Science.gov (United States)

    The U. S. Environmental Protection Agency (EPA) has determined that one of the nation’s biggest infrastructural needs is the replacement or rehabilitation of the water distribution and transmission systems. The institution of more effective pipe leak detection technology will im...

  11. Integrity conception for pipe systems

    International Nuclear Information System (INIS)

    Bartonicek, J.; Zaiss, W.; Schoeckle, F.

    2004-01-01

    There are safety-relevant mechanical components or pipe systems in which fracture must be excluded. The procedural specifications for ensuring this were developed in the early eighties at MPA-Stuttgart and updated in the mid-nineties to include ageing phenomena. The regulations are contained in KTA 3201.1 through KTA 3201.4 (orig.) [de

  12. Automatic welding machine for piping

    International Nuclear Information System (INIS)

    Yoshida, Kazuhiro; Koyama, Takaichi; Iizuka, Tomio; Ito, Yoshitoshi; Takami, Katsumi.

    1978-01-01

    A remotely controlled automatic special welding machine for piping was developed. This machine is utilized for long distance pipe lines, chemical plants, thermal power generating plants and nuclear power plants effectively from the viewpoint of good quality control, reduction of labor and good controllability. The function of this welding machine is to inspect the shape and dimensions of edge preparation before welding work by the sense of touch, to detect the temperature of melt pool, inspect the bead form by the sense of touch, and check the welding state by ITV during welding work, and to grind the bead surface and inspect the weld metal by ultrasonic test automatically after welding work. The construction of this welding system, the main specification of the apparatus, the welding procedure in detail, the electrical source of this welding machine, the cooling system, the structure and handling of guide ring, the central control system and the operating characteristics are explained. The working procedure and the effect by using this welding machine, and the application to nuclear power plants and the other industrial field are outlined. The HIDIC 08 is used as the controlling computer. This welding machine is useful for welding SUS piping as well as carbon steel piping. (Nakai, Y.)

  13. Residual stress in polyethylene pipes

    Czech Academy of Sciences Publication Activity Database

    Poduška, Jan; Hutař, Pavel; Kučera, J.; Frank, A.; Sadílek, J.; Pinter, G.; Náhlík, Luboš

    2016-01-01

    Roč. 54, SEP (2016), s. 288-295 ISSN 0142-9418 R&D Projects: GA MŠk LM2015069; GA MŠk(CZ) LQ1601 Institutional support: RVO:68081723 Keywords : polyethylene pipe * residual stress * ring slitting method * lifetime estimation Subject RIV: JL - Materials Fatigue, Friction Mechanics Impact factor: 2.464, year: 2016

  14. Spinning pipe gas lens revisited

    CSIR Research Space (South Africa)

    Mafusire, C

    2008-01-01

    Full Text Available , there is little information on optical phase aberrations and no study to date on the propagation parameters of the laser beam, but has rather remained rooted in the domain of ray optics. Researchers revisit the spinning pipe gas lens in this paper with new...

  15. Shelf life extension for the lot AAE nozzle severance LSCs

    Science.gov (United States)

    Cook, M.

    1990-01-01

    Shelf life extension tests for the remaining lot AAE linear shaped charges for redesigned solid rocket motor nozzle aft exit cone severance were completed in the small motor conditioning and firing bay, T-11. Five linear shaped charge test articles were thermally conditioned and detonated, demonstrating proper end-to-end charge propagation. Penetration depth requirements were exceeded. Results indicate that there was no degradation in performance due to aging or the linear shaped charge curving process. It is recommended that the shelf life of the lot AAE nozzle severance linear shaped charges be extended through January 1992.

  16. Numerical study on drop formation through a micro nozzle

    International Nuclear Information System (INIS)

    Kim, Sung Il; Son, Gi Hun

    2005-01-01

    The drop ejection process from a micro nozzle is investigated by numerically solving the conservation equations for mass and momentum. The liquid-gas interface is tracked by a level set method which is extended for two-fluid flows with irregular solid boundaries. Based on the numerical results, the liquid jet breaking and droplet formation behavior is found to depend strongly on the pulse type of forcing pressure and the contact angle at the gas-liquid-solid interline. The negative pressure forcing can be used to control the formation of satelite droplets. Also, various nozzle shapes are tested to investigate their effect on droplet formation

  17. 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

  18. Computational Simulation on a Coaxial Substream Powder Feeding Laval Nozzle of Cold Spraying

    Directory of Open Access Journals (Sweden)

    Guosheng HUANG

    2014-09-01

    Full Text Available In this paper, a substream coaxial powder feeding nozzle was investigated for use in cold spraying. The relationship between nozzle structure and gas flow, the acceleration behavior of copper particles were examined by computational simulation method. Also, one of the nozzle was used to spray copper coating on steel substrate. The simulation results indicate that: the velocity of gas at the center of the nozzle is lower than that of the conventional nozzle. Powders are well restrained near the central line of the nozzle, no collision occurred between the nozzle wall and the powders. This type of nozzle with expansion 3.25 can successfully deposit copper coating on steel substrate, the copper coating has low porosity about 3.1 % – 3.8 % and high bonding strength about 23.5 MPa – 26.8 MPa. DOI: http://dx.doi.org/10.5755/j01.ms.20.3.4244

  19. Gas flows in radial micro-nozzles with pseudo-shocks

    Science.gov (United States)

    Kiselev, S. P.; Kiselev, V. P.; Zaikovskii, V. N.

    2017-12-01

    In the present paper, results of an experimental and numerical study of supersonic gas flows in radial micro-nozzles are reported. A distinguishing feature of such flows is the fact that two factors, the nozzle divergence and the wall friction force, exert a substantial influence on the flow structure. Under the action of the wall friction force, in the micro-nozzle there forms a pseudo-shock that separates the supersonic from subsonic flow region. The position of the pseudo-shock can be evaluated from the condition of flow blockage in the nozzle exit section. A detailed qualitative and quantitative analysis of gas flows in radial micro-nozzles is given. It is shown that the gas flow in a micro-nozzle is defined by the complicated structure of the boundary layer in the micro-nozzle, this structure being dependent on the width-to-radius ratio of the nozzle and its inlet-to-outlet pressure ratio.

  20. DURACON - Variable Emissivity Broadband Coatings for Liquid Propellant Rocket Nozzles Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The need exists for a fast drying, robust, low gloss, black, high emissivity coating that can be applied easily on aircraft rocket nozzles and nozzle extensions....

  1. Experimental and Numerical Study of Nozzle Plume Impingement on Spacecraft Surfaces

    National Research Council Canada - National Science Library

    Ketsdever, A. D; Lilly, T. C; Gimelshein, S. F; Alexeenko, A. A

    2005-01-01

    ...) nozzle plume impinging on a simulated spacecraft surface. The nozzle flow impingement is investigated experimentally using a nano-Newton resolution force balance and numerically using the Direct Simulation Monte Carlo (DSMC...

  2. Nozzle Plume Impingement on Spacecraft Surfaces: Effects of Surface Roughness (POSTPRINT)

    National Research Council Canada - National Science Library

    Ngalande, C; Killingsworth, M; Lilly, T; Gimelshein, S; Ketsdever, A

    2005-01-01

    ...) nozzle plume impinging on simulated spacecraft surfaces. The nozzle flow impingement is investigated experimentally using a nano-Newton resolution force balance and numerically using the Direct Simulation Monte Carlo (DSMC...

  3. 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)

  4. Analysis and design of optimized truncated scarfed nozzles subject to external flow effects

    Science.gov (United States)

    Shyne, Rickey J.; Keith, Theo G., Jr.

    1990-01-01

    Rao's method for computing optimum thrust nozzles is modified to study the effects of external flow on the performance of a class of exhaust nozzles. Members of this class are termed scarfed nozzles. These are two-dimensional, nonsymmetric nozzles with a flat lower wall. The lower wall (the cowl) is truncated in order to save weight. Results from a parametric investigation are presented to show the effects of the external flowfield on performance.

  5. Practical aspects of acoustic plastic pipe location

    Energy Technology Data Exchange (ETDEWEB)

    Huebler, J.E.; Campbell, B.K. [Institute of Gas Technology, Chicago, IL (United States); Ching, G.K. [Southern California Gas Co. (United States)

    1993-12-31

    Many gas distribution company operation and maintenance activities require precise knowledge of the location of buried plastic piping. Plastic pipe cannot be located if the tracer wire is gone or was never installed. Under sponsorship of the Southern California Gas Company, IGT successfully demonstrated an acoustic plastic pipe location technique and is developing the technique into a practical field instrument an acoustic signal is injected directly into the gas at a service. The acoustic signal travels in the gas in the pipes, not in the pipe wall. As the acoustic wave travels along the pipe, some of the sound radiates from the pipe through the soil to the surface of the ground. An array of sensors on the surface of the ground perpendicular to the pipe detects the acoustic signal, thereby locating the Pipe. Two different acoustic measurements are used. The first measurement locates the pipe to within {plus_minus} 3-ft. Then the second technique determines the location of the pipe to within {plus_minus} 6-in.

  6. Underground pipe inspection device and method

    Energy Technology Data Exchange (ETDEWEB)

    Germata, Daniel Thomas [Wadsworth, IL

    2009-02-24

    A method and apparatus for inspecting the walls of an underground pipe from inside the pipe in which an inspection apparatus having a circular planar platform having a plurality of lever arms having one end pivotably attached to one side of the platform, having a pipe inspection device connected to an opposite end, and having a system for pivoting the lever arms is inserted into the underground pipe, with the inspection apparatus oriented with the planar platform disposed perpendicular to the pipe axis. The plurality of lever arms are pivoted toward the inside wall of the pipe, contacting the inside wall with each inspection device as the apparatus is conveyed along a length of the underground pipe.

  7. Effects of dimensional size and surface roughness on service performance for a micro Laval nozzle

    Science.gov (United States)

    Cai, Yukui; Liu, Zhanqiang; Shi, Zhenyu

    2017-05-01

    Nozzles with large and small dimensions are widely used in various industries. The main objective of this research is to investigate the effects of dimensional size and surface roughness on the service performance of a micro Laval nozzle. The variation of nozzle service performance from the conventional macro to micro scale is presented in this paper. This shows that the dimensional nozzle size has a serious effect on the nozzle gas flow friction. With the decrease of nozzle size, the velocity performance and thrust performance deteriorate. The micro nozzle performance has less sensitivity to the variation of surface roughness than the large scale nozzle does. Surface quality improvement and burr prevention technologies are proposed to reduce the friction effect on the micro nozzle performance. A novel process is then developed to control and depress the burr generation during micro nozzle machining. The polymethyl-methacrylate as a coating material is coated on the rough machined surface before finish machining. Finally, the micro nozzle with a throat diameter of 1 mm is machined successfully. Thrust test results show that the implement and application of this machining process benefit the service performance improvement of the micro nozzle.

  8. Effects of dimensional size and surface roughness on service performance for a micro Laval nozzle

    International Nuclear Information System (INIS)

    Cai, Yukui; Liu, Zhanqiang; Shi, Zhenyu

    2017-01-01

    Nozzles with large and small dimensions are widely used in various industries. The main objective of this research is to investigate the effects of dimensional size and surface roughness on the service performance of a micro Laval nozzle. The variation of nozzle service performance from the conventional macro to micro scale is presented in this paper. This shows that the dimensional nozzle size has a serious effect on the nozzle gas flow friction. With the decrease of nozzle size, the velocity performance and thrust performance deteriorate. The micro nozzle performance has less sensitivity to the variation of surface roughness than the large scale nozzle does. Surface quality improvement and burr prevention technologies are proposed to reduce the friction effect on the micro nozzle performance. A novel process is then developed to control and depress the burr generation during micro nozzle machining. The polymethyl-methacrylate as a coating material is coated on the rough machined surface before finish machining. Finally, the micro nozzle with a throat diameter of 1 mm is machined successfully. Thrust test results show that the implement and application of this machining process benefit the service performance improvement of the micro nozzle. (paper)

  9. Within-band spray distribution of nozzles used for herbaceous plant control

    Science.gov (United States)

    James H. Miller

    1994-01-01

    Abstract. Described are the spray patterns of nozzles setup for banded herbaceous plant control treatments. Spraying Systems Company nozzles. were tested, but similar nozzles are available from other manufacturers. Desirable traits were considered to be as follows: an even distribution pattern, low volume, low height, large droplets, and a single...

  10. Inelastic pipe elements for analysis of pipe whip

    International Nuclear Information System (INIS)

    Powell, H.

    1977-01-01

    Two alternative assumptions for the effects of moment interaction following yielding of a pipe are compared. The piping system must usually be divided into short finite elements, in order to account for wave propagation through the piping. Where short elements are used, it is accurate and convenient to use a lumped plasticity finite element model, the pipe being represented by three-dimensional beam-column elements in which yielding is assumed to be concentrated in generalized plastic hinges at the element ends. It is also convenient to assume that the generalized moment-rotation relationship at a hinge is elastic-perfectly-plastic, and to account for strain hardening using the well-known parallel element procedure. With this assumption, the task of monitoring hinge behavior is simplified, yet completely arbitrary moment-rotation relationships can be constructed. The interaction relationship defining the combinations of bending and torsional moments which produce yield at a plastic hinge can easily be determined. Classical plasticity theory adopts the normality criterion, in which post-yield deformations are divided into components normal and tangential to the yield surface. The normal components are then assumed to be plastic, producing no change in moment, and the tangential rotations to be elastic, producing moment change in accordance with the element elastic stiffness. An alternative, simpler assumption is that post-yield rotations are entirely plastic. With this assumption, the moments at the hinge remain unchanged, as in a 'rusty' hinge. The elasto-plastic element stiffness for this model does not change continuously during the response analysis, so that the computation is simpler, more economical, but less accurate

  11. Theoretical determination of nozzle admittances using a finite element method

    Science.gov (United States)

    Sigman, R. K.; Zinn, B. T.

    1979-01-01

    A finite element method (FEM) is used to predict the admittances of axisymmetric nozzles. The flow in the nozzle is assumed to be isentropic and the disturbances are assumed to be small so that linear analyses apply. An approximate two dimensional compressible flow model is used to describe the steady flow in the nozzle. The propagation of acoustic disturbances is governed by the complete linear acoustic wave equation. This partial differential wave equation is transformed to an integral equation using Galerkin's method and Green's theorem is applied so that the acoustic boundary conditions can be introduced through the boundary residuals. A two dimensional finite element method using linear triangular elements is used to solve the integral acoustic equation. A one dimensional FEM is used to solve the reduced nozzle acoustic equation developed by Crocco and the solution is used to verify the sufficiency of the boundary residual formation. It is shown that agreement between predicted values of the admittance and experimental data is quite good.

  12. SHINE Tritium Nozzle Design: Activity 6, Task 1 Report

    Energy Technology Data Exchange (ETDEWEB)

    Okhuysen, Brett S. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Pulliam, Elias Noel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-05

    In FY14, we studied the qualitative and quantitative behavior of a SHINE/PNL tritium nozzle under varying operating conditions. The result is an understanding of the nozzle’s performance in terms of important flow features that manifest themselves under different parametric profiles. In FY15, we will consider nozzle design with a focus on nozzle geometry and integration. From FY14 work, we will understand how the SHINE/PNL nozzle behaves under different operating scenarios. The first task for FY15 is to evaluate the FY14 model as a predictor of the actual flow. Considering different geometries is more time-intensive than parameter studies, therefore we recommend considering any relevant flow features that were not included in the FY14 model. In the absence of experimental data, it is particularly important to consider any sources of heat in the domain or boundary conditions that may affect the flow and incorporate these into the simulation if they are significant. Additionally, any geometric features of the beamline segment should be added to the model such as the orifice plate. The FY14 model works with hydrogen. An improvement that can be made for FY15 is to develop CFD properties for tritium and incorporate those properties into the new models.

  13. Numerical analysis of choked converging nozzle flows with surface ...

    Indian Academy of Sciences (India)

    (5a–b). One-dimensional momentum and energy equations (6) and (7) are applied to each differential cell in the nozzle, where the nodal properties such as P,U and Cp are interrelated with the contributions of cellular variants like Ff ,I,dq and . Equation (7) represents the conserva- tion of mechanical and thermal energies ...

  14. Development of rapid mixing fuel nozzle for premixed combustion

    Energy Technology Data Exchange (ETDEWEB)

    Katsuki, Masashi; Chung, Jin Do; Kim, Jang Woo; Hwang, Seung Min [Hoseo University, Asan (Korea, Republic of); Kim, Seung Mo [Pusan National University, Busan (Korea, Republic of); Ahn, Chul Ju [Osaka University, Osaka (Japan)

    2009-03-15

    Combustion in high-preheat and low oxygen concentration atmosphere is one of the attractive measures to reduce nitric oxide emission as well as greenhouse gases from combustion devices, and it is expected to be a key technology for the industrial applications in heating devices and furnaces. Before proceeding to the practical applications, we need to elucidate combustion characteristics of non-premixed and premixed flames in high-preheat and low oxygen concentration conditions from scientific point of view. For the purpose, we have developed a special mixing nozzle to create a homogeneous mixture of fuel and air by rapid mixing, and applied this rapidmixing nozzle to a Bunsen-type burner to observe combustion characteristics of the rapid-mixture. As a result, the combustion of rapid-mixture exhibited the same flame structure and combustion characteristics as the perfectly prepared premixed flame, even though the mixing time of the rapid-mixing nozzle was extremely short as a few milliseconds. Therefore, the rapid-mixing nozzle in this paper can be used to create preheated premixed flames as far as the mixing time is shorter than the ignition delay time of the fuel

  15. Hypersonic Wind Tunnel Nozzle Survivability for T&E

    Science.gov (United States)

    2007-03-01

    used to melt the electrode while a second electron beam was used to control the rate of solidification in the mold . Gas bubbles tend to come to the...38 4.4 Ni -Coated Cu - Back-Side-Cooled Arc-Heater Nozzles .............................................45 5.0 SUMMARY/CONCLUSIONS...25 25. Principal Stress Distribution for Direction 1

  16. Nonlinear indirect combustion noise for compact supercritical nozzle flows

    Science.gov (United States)

    Huet, M.

    2016-07-01

    In this paper, indirect combustion noise generated by the acceleration of entropy perturbations through a supercritical nozzle is investigated in the nonlinear regime and in the low-frequency limit (quasi-static hypothesis). This work completes the study of Huet and Giauque (Journal of Fluid Mechanics 733 (2013) 268-301) for nonlinear noise generation in nozzle flows without shock and particularly focuses on shocked flow regimes. It is based on the analytical model of Marble and Candel for compact nozzles (Journal of Sound and Vibration 55 (1977) 225-243), initially developed for excitations in the linear regime and rederived here for nonlinear perturbations. Full nonlinear analytical solutions are provided in the absence of shock as well as second-order analytical expressions when a shock is present in the diffuser. An analytical evaluation of the shock displacement inside the nozzle caused by the forcing is proposed and maximum possible forcings to avoid unchoke and 'over-choke' are discussed. The accuracy of the second-order model and the nonlinear contributions to the generated waves are then addressed. This model is found to be very accurate for the generated entropy wave with negligible nonlinear contributions. Nonlinearities are more visible, but still limited, for the downstream acoustic wave for large inlet Mach numbers. Analytical developments are validated thanks to comparisons with numerical simulations.

  17. The separation nozzle process for uranium isotope enrichment

    International Nuclear Information System (INIS)

    Becker, E.W.

    1977-01-01

    The paper covers the most important steps in the technological development and the future prospects of the separation nozzle process. In this process uranium isotope separation is brought about by the mass dependence of the centrifugal forces in a curved flow of a UF 6 /H 2 mixture. Due to the large excess in hydrogen, the high ratio of UF 6 flow velocity to thermal velocity required for an effective isotope separation is obtained at relatively low expansion ratios and, accordingly, with relatively low gas-dynamic losses. As the optimum Reynolds number of the curved jet is comparatively low, and as a high absolute pressure is essential for economic reasons, the characteristic dimensions of the nozzle systems are made as small as possible. For commercial application in the near future, systems involving mechanical jet deflection have been developed. Promising results were, however, also obtained with separation nozzle systems generating a streamline curvature by the interaction of opposed jets. Most of the development work has been done at the Nuclear Research Centre, Karlsruhe. Since 1970 the STEAG company (FRG) has been involved in the commercial implementation of the process. Two industrial-scale separative stages were tested successfully. This work constitutes the basis of planning of a separation nozzle demonstration plant to be built in Brazil. (author)

  18. Construction of a pulsed nozzle fourier transform microwave ...

    Indian Academy of Sciences (India)

    Administrator

    Construction of a pulsed nozzle fourier transform microwave spectrometer to study the lithium bond. A P TIWARI 1, B J MUKKADA 1, E ARUNAN 1 and P C MATHIAS 2. 1Department of Inorganic and Physical Chemistry, Indian Institute of. Science, Bangalore 560 012, India. 2Sophisticated Instruments Facility, Indian Institute ...

  19. Ayame/PAM-D apogee kick motor nozzle failure analysis

    Science.gov (United States)

    1981-01-01

    The failure of two communication satellites during firing sequence were examined. The correlation/comparison of the circumstances of the Ayame incidents and the failure of the STAR 48 (DM-2) motor are reviewed. The massive nozzle failure of the AKM to determine the impact on spacecraft performance is examined. It is recommended that a closer watch is kept on systems techniques,

  20. The jet nozzle process for uranium 235 isotopic enrichment

    International Nuclear Information System (INIS)

    Jordan, I.; Umeda, K.; Brown, A.E.P.

    1979-01-01

    A general survey of the isotopic enrichment of Uranium - 235, principally by jet nozzle process, is made. Theoretical treatment of a single stage and cascade of separation stages of the above process with its development in Germany until 1976 is presented [pt