Water-hammer in the feed-water pipes for PWR steam generators

International Nuclear Information System (INIS)

Gonnet, Bernard; Leroy, Claude; Oullion, Jean; Yazidjian, J.-C.

1979-01-01

PWR boiler water feed pipes have been known for several years to be affected by violent water-hammer during start-ups and operation of the plant. In view of the varying results of corrective design modifications in America and Europe, FRAMATOME undertook an experimental research programme which resulted in the adoption of cruciform tubes on the feed-water distributor as the most reliable solution. Subsequent tests at Fessenheim I confirmed the effectiveness of this device [fr

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.)

Evaluation of cracking in steam generator feedwater piping in pressurized water reactor plants

International Nuclear Information System (INIS)

Goldberg, A.; Streit, R.D.

1981-05-01

Cracking in feedwater piping was detected near the inlet to steam generators in 15 pressurized water reactor plants. Sections with cracks from nine plants are examined with the objective of identifying the cracking mechanism and assessing various factors that might contribute to this cracking. Using transmission electron microscopy, fatigue striations are observed on replicas of cleaned crack surfaces. Calculations based on the observed striation spacings gave a cyclic stress value of 150 MPa (22 ksi) for one of the major cracks. The direction of crack propagation was invariably related to the piping surface and not to the piping axis. These two factors are consistent with the proposed concept of thermally induced, cyclic, tensile surface stresses and it is concluded that the overriding factor in the cracking problem was the presence of such undocumented cyclic loads

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

Reactor feedwater system

International Nuclear Information System (INIS)

Kagaya, Hiroyuki; Tominaga, Kenji.

1993-01-01

In a simplified water type reactor using a gravitationally dropping emergency core cooling system (ECCS), the present invention effectively prevents remaining high temperature water in feedwater pipelines from flowing into the reactor upon occurrence of abnormal events. That is, (1) upon LOCA, if a feedwater pipeline injection valve is closed, boiling under reduced pressure of the remaining high temperature water occurs in the feedwater pipelines, generated steams prevent the remaining high temperature water from flowing into the reactor. Accordingly, the reactor is depressurized rapidly. (2) The feedwater pipeline injection valve is closed and a bypassing valve is opened. Steams generated by boiling under reduced pressure of the remaining high temperature water in the feedwater pipelines are released to a condensator or a suppression pool passing through bypass pipelines. As a result, the remaining high temperature water is prevented from flowing into the reactor. Accordingly, the reactor is rapidly depressurized and cooled. It is possible to accelerate the depressurization of the reactor by the method described above. Further, load on the depressurization valve disposed to a main steam pipe can be reduced. (I.S.)

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.).

Feed water distribution pipe replacement at Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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.).

Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks. Phase report

International Nuclear Information System (INIS)

Besuner, P.M.; Caughey, W.R.

1976-11-01

The finite element (FE) and influence function (IF) methods are compared for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). These are two of the possible methods for determining stress intensity factors for nozzle corner cracks. The FE method is incorporated in a direct manner. The IF method is used to compute stress intensity factors only when the uncracked stress field (i.e., the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. Both the IF and FE methods are described in detail and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availablility of an accurate published result obtained from some recognized third method of solution

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.

Getting the most out of your new plant with a chordal ultrasonic feedwater flow measurement system

International Nuclear Information System (INIS)

Estrada, Herb; Hauser, Ernie

2007-01-01

The economic advantages of a chordal ultrasonic feedwater flow measurement system over conventional (flow nozzle-based) feedwater instrumentation are analyzed for new plants having ratings ranging from 1100 MWe to 1600 MWe. Specifically, each of the following topics is considered: The value of a 1.7% increase in the rating of the new plant, made possible by the reduced uncertainty in the determination of thermal power. The value of reduced startup time owing to enhanced steam supply water level control. The value of the reduced feedwater pumping power brought about by the elimination of flow nozzles. The value of the reduced calibration burden owing to the elimination of the feedwater flow differential pressure transmitters and resistance thermometers. The net difference in the acquisition costs of the ultrasonic system versus conventional feedwater flow instrumentation. The net savings in installation costs of the ultrasonic system vis-a-vis conventional feedwater flow instrumentation. The potential savings in outage time due to the reduced frequency of low steam supply water level trips (scrams) of the reactor. (author)

Numerical Analysis of Pelton Nozzle Jet Flow Behavior Considering Elbow Pipe

Science.gov (United States)

Chongji, Zeng; Yexiang, Xiao; Wei, Xu; Tao, Wu; Jin, Zhang; Zhengwei, Wang; Yongyao, Luo

2016-11-01

In Pelton turbine, the dispersion of cylindrical jet have a great influence on the energy interaction of jet and buckets. This paper simulated the internal flow of nozzle and the downstream free jet flow at 3 different needle strokes. The nozzle model consists of the elbow pipe and the needle rod which supported by 4 ribs. Homogenous model and SST k-ω model were adopted to simulate the unsteady two-phase jet flow. The development of free flow, including a contraction process followed by an expansion process, was analysed detailed as well as the influence of the nozzle geometry on the jet flow pattern. The increase of nozzle opening results in a more dispersion jet, which means a higher hydraulic loss. Upstream bend and ribs induce the secondary flow in the jet and decrease the jet concentration.

Ethanolamine properties and use for feedwater pH control: A pressurized water reactor case study

International Nuclear Information System (INIS)

Keeling, D.L.; Polidoroff, C.T.; Cortese, S.; Cushner, M.C.

1995-01-01

Ethanolamine (ETA) as a feedwater pH control additive has been recently used to minimize corrosion of secondary water components in the nuclear power industry pressurized water reactors (PWRs). The use of ETA is compared with ammonia. Relative volatility effects on various parts of the system are analyzed and chemistry changes are presented. Materials of construction and the use of existing plant equipment for ETA service are discussed. Properties of ETA as well as safety, storage and handling issues are compared with ammonia. Health d aquatic toxicity are reviewed. warnings, safety, handling guidelines, biodegradability an Diablo Canyon Power Plant used ammonia for pH control from 1985 until a change over to ETA in 1993/1994. Full flow condensate polishers that are required to protect the plant from saltwater cooling incursions limit the amount of pH additive. Iron levels in the secondary water systems are compared before and after changing to ETA and replacement of corrosion-susceptible piping. Iron reduction benefits are assessed along with other effects on the feedwater nozzles, low pressure turbine, polisher resin capacity and polisher regeneration system

Comparison of finite element and influence function methods for three-dimensional elastic analysis of boiling water reactor feedwater nozzle cracks

International Nuclear Information System (INIS)

Besuner, P.M.; Caughey, W.R.

1976-11-01

The paper compares the finite element (FE) and influence function (IF) methods for a three-dimensional elastic analysis of postulated circular-shaped surface cracks in the feedwater nozzle of a typical boiling water reactor (BWR). The FE method is incorporated in a direct manner. The nozzle and crack geometry and the complex loading are all included in the model which simulates the structural crack problem. The IF method is used to compute stress intensity factors only when the uncracked stress field (that is, the stress in the uncracked solid at the locus of the crack to be eventually considered) has been computed previously. The IF method evaluates correctly the disturbance of this uncracked stress field caused by the crack by utilizing a method of elastic superposition. Both the IF and FE methods are described in detail in the paper and are applied to several test cases chosen for their similarity to the nozzle crack problem and for the availability of an accurate published result obtained from some recognized third method of solution. Results are given which summarize both the accuracy and the direct computer costs of the two methods

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)

Behaviour of a pressure vessel nozzle with thermo-sleeve under thermal loading induced by stratified flow

International Nuclear Information System (INIS)

Kussmaul, K.; Mayinger, W.; Diem, H.; Katzenmeier, G.

1993-01-01

Startup at low reactor power may give rise to stratified flow conditions in pipes of boiling water and pressurized water reactors. Stratified flow regimes cause a steep temperature gradient between the cold and the hot fluid layer. This temperature gradient produces high axial stresses which, in the case of intermittent feeding of cold water and an appropriate number of repetitions, in principle may initiate cracking in the feedwater pipe and close to the feeding nozzle. Thermosleeves have been installed in a number of reactors to mitigate thermally induced stresses; they reduce the intensity of thermal transients by means of an insulating fluid annulus developing between the sleeve and the nozzle, in order to measure the temperature and stress gradients occurring in the region of the nozzle edge, the so-called TEMS experiments were carried out under realistic operating conditions, and with different cold water levels within the framework of German research activities in the field of reactor safety at the HDR test facility. The experiments served to simulate the physics phenomena by means of a FE-program and to verify the computational approach by comparisons of measurements and calculations

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)

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)

Altitude Performance Characteristics of Tail-pipe Burner with Variable-area Exhaust Nozzle

Science.gov (United States)

Jansen, Emmert T; Thorman, H Carl

1950-01-01

An investigation was conducted in the NACA Lewis altitude wind tunnel to determine effect of altitude and flight Mach number on performance of tail-pipe burner equipped with variable-area exhaust nozzle and installed on full-scale turbojet engine. At a given flight Mach number, with constant exhaust-gas and turbine-outlet temperatures, increasing altitude lowered the tail-pipe combustion efficiency and raised the specific fuel consumption while the augmented thrust ratio remained approximately constant. At a given altitude, increasing flight Mach number raised the combustion efficiency and augmented thrust ratio and lowered the specific fuel consumption.

Cold water injection nozzles

International Nuclear Information System (INIS)

Kura, Masaaki; Maeda, Masamitsu; Endo, Takio.

1979-01-01

Purpose: To inject cold water in a reactor without applying heat cycles to a reactor container and to the inner wall of a feedwater nozzle by securing a perforated plate at the outlet of the cold water injection nozzle. Constitution: A disc-like cap is secured to the final end of a return nozzle of a control rod drive. The cap prevents the flow of a high temperature water flowing downward in the reactor from entering into the nozzle. The cap is perforated with a plurality of bore holes for injecting cold water into the reactor. The cap is made to about 100 mm in thickness so that the cold water passing through the bore holes is heated by the heat conduction in the cap. Accordingly, the flow of high temperature water flowing downwardly in the reactor is inhibited by the cap from backward flowing into the nozzle. Moreover, the flow of the cold water in the nozzle is controlled and rectified when passed through the bore holes in the cap and then injected into the reactor. (Yoshino, Y.)

A connection of the steam generator feedwater section of WWER type nuclear power plants

International Nuclear Information System (INIS)

Matal, O.; Sadilek, J.

1989-01-01

In the feedwater piping of each steam generator, a plate for additional water pressure reduction is inserted before the first closing valve. During a steady water flow, the plate gives rise to a constant hydraulic resistance, bringing about steady reduction of the feedwater pressure; this also contributes to a stabilization of the feedwater flow rate into the steam generator. The control valve thus is stressed by minimal hydrodynamic forces. In this manner its load is decreased, its vibrations are damped, and the frequency of failures - and thereby the frequency of the nuclear power plant unit outages -is reduced. (J.P.). 1 fig

Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

International Nuclear Information System (INIS)

Wietelmann, F.

1978-01-01

In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

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)

Device for preventing cooling water from flowing out of reactor

International Nuclear Information System (INIS)

Chinen, Masanori; Kotani, Koichi; Murase, Michio.

1976-01-01

Object: To provide emergency cooling system, which can prevent cooling water bearing radioactivity from flowing to the outside of the reactor at the time of breakage of feedwater pipe, thus eliminating the possibility of exposure of the fuel rod to provide high reliability and also reducing the possibility of causing radioactive pollution. Structure: The device for preventing cooling water from flowing out from the reactor features a jet nozzle inserted in a feedwater pipe adjacent to the inlet or outlet thereof immediately before the reactor container. The nozzle outlet is provided in the vicinity of the reactor wall and in a direction opposite to the direction of out-flow, and water supplied from a high pressure pump is jetted from it. (Nakamura, S.)

Investigation into sensitivity of Darlington boiler 2 feedwater flow calibration factor to boiler level control valve configuration

Energy Technology Data Exchange (ETDEWEB)

Coppens, D. [Darlington Nuclear Generating Station, Ontario Power Generation, Bowmanville, Ontario (Canada); Gurevich, Y. [Daystar Technologies Inc., Toronto, Ontario (Canada); Ton, V. [Inspection and Maintenance Services Div., Ontario Power Generation, Ajax, Ontario (Canada); Zobin, D. [AMEC NSS Ltd., Toronto, Ontario (Canada)

2009-07-01

The Ultrasonic Cross-Correlation Flow Meter (USCCFM) has been used for regular feedwater flow calibration at Darlington NGS since the early nineties. Typical measurement repeatability over the duration of a calibration run (normally several weeks long) is within {+-}0.2%. However, it was recently noticed that BO2 calibration factor experienced sudden changes of close to 1%. The paper will describe several different approaches used for identifying the reason for the observed effect. The investigation has revealed that changes in USCCFM readings are due to the complicated geometry of BO2 feedwater piping and that its accuracy can be as high as a fraction of percent if several readings are averaged around the pipe. (author)

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

New assessment of feed water piping in GKN I including optimisation of piping supports

International Nuclear Information System (INIS)

Zaiss, W.; Heil, C.; Baier, B.; Manke, A.

2003-01-01

The quality of nuclear power plant components and piping is specified according to the then current state of knowledge. In operation, the quality can be reduced by ageing phenomena, so in-service quality assessment is constantly required. The contribution discusses the individual aspects of reassessment and its technical procedure, using the example of a feedwater pipe in the GKN I containment. (orig.) [de

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)

Factors analysis of water hammer in FLOWMASTER for main feedwater systems 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 (NPP) is an important part in ensuring the cooling of a steam generator. It is the main pipe section where water hammers frequently occur. Studying the regulator patterns of water hammers in the main feedwater systems is significant to the stable operation of the system. This article focuses on a parametric study to avoid the consequences of water hammer effect in PWR by employing a general purpose fluid dynamic simulation software-FLOWMASTER. Through FLOWMASTER's transient calculating functions, a mathematical model is established with boundary conditions such as feedwater pumps, control valves, etc., calculations of water hammer pressure when feedwater pumps and control valves shut down, and simulations during instantaneous changes in water hammer pressure. Combining a plethora of engineering practical examples, this research verified the viability of calculating water hammer pressure through FLOWMASTER's transient functions and we found out that, increasing the periods of closure of control valves and feedwater pumps control water hammers effectively. We also found out that changing the intervals of closing signals to feedwater pumps and control valves aid to relieve hydraulic impact. This could be a guideline for practical engineering design and system optimization. (author)

Leak-before-break analysis of a dissimilar metal welded joint for connecting pipe-nozzle in nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Gong, N. [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Wang, G.Z., E-mail: gzwang@ecust.edu.cn [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China); Xuan, F.Z.; Tu, S.T. [MOE Key Laboratory of Pressurized System and Safety, School of Mechanical and Power Engineering, East China University of Science and Technology, Shanghai 200237 (China)

2013-02-15

Highlights: ► Leak-before-break (LBB) analysis for a dissimilar metal weld joint (DMWJ) is made. ► Pipe-nozzle geometry and inhomogeneous material property of DMWJ are incorporated. ► LBB behavior of a defect can be assessed by LBB assessment diagram and LBB curve. ► Feasibility region of LBB is enlarged with decreasing load and increasing J{sub R}. -- Abstract: This paper presents a leak-before-break (LBB) analysis for a dissimilar metal welded joint (DMWJ) connected the safe end to pipe-nozzle of a reactor pressure vessel of which is relevant to safety of nuclear power plant. Three-dimensional finite element analysis models were built for the DMWJ structure, and the initial inner circumferential surface cracks were postulated at the interface between A508 steel and buttering Alloy82. Based on the elastic–plastic fracture mechanics theory of J-integral, the crack growth stability was analyzed, and the pipe-nozzle geometry effect and inhomogeneous material properties of the DMWJ have been incorporated. Base on the analysis results, the LBB curves and LBB assessment diagrams were constructed for the DMWJ, and effects of applied bending moment loads and J-resistance curves of materials on LBB behavior were analyzed. The results show that the LBB behavior of a defect in the DMWJ under an upmost severe load can be assessed and predicted by plotting the defect size and its propagation path in the LBB assessment diagrams. With decreasing the maximum bending moment load and increasing the crack growth resistance of materials, the ligament instability lines shift upward and the critical crack length lines move to the right in the LBB assessment diagrams, which leads to enlargement of the feasibility region in the LBB behavior.

Altitude Performance Characteristics of Turbojet-engine Tail-pipe Burner with Variable-area Exhaust Nozzle Using Several Fuel Systems and Flame Holders

Science.gov (United States)

Johnson, Lavern A; Meyer, Carl L

1950-01-01

A tail-pipe burner with a variable-area exhaust nozzle was investigated. From five configurations a fuel-distribution system and a flame holder were selected. The best configuration was investigated over a range of altitudes and flight Mach numbers. For the best configuration, an increase in altitude lowered the augmented thrust ratio, exhaust-gas total temperature, and tail-pipe combustion efficiency, and raised the specific fuel consumption. An increase in flight Mach number raised the augmented thrust ratio but had no apparent effect on exhaust-gas total temperature, tail-pipe combustion efficiency, or specific fuel consumption.

Study on mixed analysis method for fatigue analysis of oblique safety injection nozzle on main piping

International Nuclear Information System (INIS)

Lu Xifeng; Zhang Yixiong; Ai Honglei; Wang Xinjun; He Feng

2014-01-01

The simplified analysis method and the detailed analysis method were used for the fatigue analysis of the nozzle on the main piping. Because the structure of the oblique safety injection nozzle is complex and some more severe transients are subjected. The results obtained are more penalized and cannot be validate when the simplified analysis method used for the fatigue analysis. It will be little conservative when the detailed analysis method used, but it is more complex and time-consuming and boring labor. To reduce the conservatism and save time, the mixed analysis method which combining the simplified analysis method with the detailed analysis method is used for the fatigue analysis. The heat transfer parameters between the fluid and the structure which used for analysis were obtained by heat transfer property experiment. The results show that the mixed analysis which heat transfer property is considered can reduce the conservatism effectively, and the mixed analysis method is a more effective and practical method used for the fatigue analysis of the oblique safety injection nozzle. (authors)

Piping benchmark problems for the ABB/CE System 80+ Standardized Plant

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1994-07-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for the ABB/Combustion Engineering System 80+ Standardized Plant, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the System 80+ standard design. It will be required that the combined license licensees demonstrate that their solution to these problems are in agreement with the benchmark problem set. The first System 80+ piping benchmark is a uniform support motion response spectrum solution for one section of the feedwater piping subjected to safe shutdown seismic loads. The second System 80+ piping benchmark is a time history solution for the feedwater piping subjected to the transient loading induced by a water hammer. The third System 80+ piping benchmark is a time history solution of the pressurizer surge line subjected to the accelerations induced by a main steam line pipe break. The System 80+ reactor is an advanced PWR type

Design and transient analyses of passive emergency feedwater system of CPR1000. Part 1. Air cooling condition

International Nuclear Information System (INIS)

Zhang Yapei; Qiu Suizheng; Su Guanghui; Tian Wenxi; Cao Jianhua; Lu Donghua; Fu Xiangang

2011-01-01

The steam generator secondary passive emergency feedwater system is a new design for traditional generation Ⅱ + reactor CPR1000. The passive emergency feedwater system is designed to supply water to the SG shell side and improve the safety and reliability of CPR1000 by completely or partially replacing traditional emergency water cooling system in the event of the feed line break (FLB) or loss of heat sink accident. The passive emergency feedwater system consists of steam generator (SG), heat exchanger (HX), air cooling tower, emergency makeup tank (EMT), and corresponding pipes and valves for air cooling condition. In order to improve the safety and reliability of CPR1000, the model of the primary loop system and the passive emergency feedwater system was developed to investigate residual heat removal capability of the passive emergency feedwater system and the transient characteristics of the primary loop system affected by the passive emergency feedwater system using RELAP5/MOD3.4. The transient characteristics of the primary loop system and the passive emergency feedwater system were calculated in the event of feed line break accident. Sensitivity studies of the passive emergency feedwater system were also conducted to investigate the response of the primary loop and the passive emergency feedwater system on the main parameters of the passive emergency feedwater system. The passive emergency feedwater system could supply water to the SG shell side from the EMT successfully. The calculation results showed that the passive emergency feedwater system could take away the decay heat from the primary loop effectively for air cooling condition, and that the single-phase and two-phase natural circulations were established in the primary loop and passive emergency feedwater system loop, respectively. (author)

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.).

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.)

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)

1998-12-31

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.).

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

International Nuclear Information System (INIS)

Fuller, R.; Harrell, J.

1996-01-01

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

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.

PSA effect analysis of a design modification of the auxiliary feedwater system for a Westinghouse type plant

International Nuclear Information System (INIS)

Bae, Yeon Kyoung; Lee, Eun Chan

2012-01-01

The auxiliary feedwater system is an important system used to mitigate most accidents considered in probabilistic safety assessment (PSA). The reference plant has produced electric power for about thirty years. Due to age related deterioration and lack of parts, a turbine driven auxiliary feedwater pump (TD AFWP), some valves, and piping of the auxiliary feedwater system should be replaced. This change includes relocation of some valves, installation of valves for maintenance of the steam generator, and a new cross tie line. According to the design change, the Final Safety Analysis Report (FSAR) has been revised. Therefore, this design modification affects the PSA. It is thus necessary to assess the improvement of plant safety. In this paper, the impact of the design change of the auxiliary feedwater system on the PSA is assessed. The results demonstrate that this modification considering the plant safety decreased the total CDF

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.)

Optimally growing boundary layer disturbances in a convergent nozzle preceded by a circular pipe

Science.gov (United States)

Uzun, Ali; Davis, Timothy B.; Alvi, Farrukh S.; Hussaini, M. Yousuff

2017-06-01

We report the findings from a theoretical analysis of optimally growing disturbances in an initially turbulent boundary layer. The motivation behind this study originates from the desire to generate organized structures in an initially turbulent boundary layer via excitation by disturbances that are tailored to be preferentially amplified. Such optimally growing disturbances are of interest for implementation in an active flow control strategy that is investigated for effective jet noise control. Details of the optimal perturbation theory implemented in this study are discussed. The relevant stability equations are derived using both the standard decomposition and the triple decomposition. The chosen test case geometry contains a convergent nozzle, which generates a Mach 0.9 round jet, preceded by a circular pipe. Optimally growing disturbances are introduced at various stations within the circular pipe section to facilitate disturbance energy amplification upstream of the favorable pressure gradient zone within the convergent nozzle, which has a stabilizing effect on disturbance growth. Effects of temporal frequency, disturbance input and output plane locations as well as separation distance between output and input planes are investigated. The results indicate that optimally growing disturbances appear in the form of longitudinal counter-rotating vortex pairs, whose size can be on the order of several times the input plane mean boundary layer thickness. The azimuthal wavenumber, which represents the number of counter-rotating vortex pairs, is found to generally decrease with increasing separation distance. Compared to the standard decomposition, the triple decomposition analysis generally predicts relatively lower azimuthal wavenumbers and significantly reduced energy amplification ratios for the optimal disturbances.

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

International Nuclear Information System (INIS)

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

1991-01-01

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

Aging and low-flow degradation of auxilary feedwater pumps

International Nuclear Information System (INIS)

Adams, M.L.

1992-01-01

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 related 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

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

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)

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.)

Assessment of a potential rapid condensation induced water hammer in a passive auxiliary feedwater system

Energy Technology Data Exchange (ETDEWEB)

Jo, Jong Chull; Shin, Byung Soo; Do, Kyu Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Moody, Frederick J. [General Electric (Retired), CA (United States)

2012-10-15

A passive auxiliary feedwater system (PAFS) which is incorporated in the APR+ system is a kind of closed natural circulation loop. The PAFS has no operating functions during normal plant operation, but it has a dedicated safety function of the residual heat removal following initiating events, including the unlikely event of the most limiting single failure occurring coincident with a loss of offsite power, when the feedwater system becomes inoperable or unavailable. Even in the unlikely event of a station blackout, the isolation valves can be opened either by DC power or manual operation and then the PAFS can also provide adequate condensate to the steam generator (SG). The PAFS piping in the vicinity of each of the two SGs is designed to minimize the potential for destructive water hammer during start up operation by setting the stroke time for full close or full open of the condensate isolation valves upon receipt of a passive auxiliary feedwater actuation signal. The temperature of the stagnant condensate water and its surrounding tubes and piping during the reactor normal operation modes may fall to the ambient temperature. A possible concern is the introduction of saturated steam into the PAFS recirculation pipe downstream of the PCHX in the beginning of the PAFS operation. Although the steam introduction rate is expected to be slow, a rapid condensation rate is expected due to the initial cold surrounding temperature in the pipe, which could result in a localized pressure reduction and the propagation of decompression and velocity disturbances into the condensate water leg, which might cause the sudden closure of check valves and associated water hammer. Thus, it is requisite for the licensing review of the PAFS design to confirm if destructive water hammers will not be produced due to such rapid condensation induced decompressions in the system. This paper addresses an assessment of the potential local decompressions which could result from the steam

Dynamics of the nozzle valve with regard to the properties of the piping system

Directory of Open Access Journals (Sweden)

Klas Roman

2018-01-01

Full Text Available It is obvious that the main function of the nozzle valve is to shut off the stream of fluid in the piping system. The response rate of the valve to the decreasing or reversing flow in the system will then depend on the valve properties and equally on the properties of the piping system. The interaction of these two elements is also important for the origin of pressure pulsations in the system. While the pressure pulsations were the cause for design of this particular valve it should be noted that the general design of the valve for any pipeline system is not possible. The valve cannot properly work under all circumstances and operating conditions. With respect to this, the dynamic properties of the valve will be assessed on the basis of the valve equation of motion and the pipeline model. An adequate response of the whole system can be obtained by combining both approaches. The valve equations of motion are also complemented by CFD simulations, which enable to capture the movement of the valve disc with respect to flow rate.

Reactor feedwater device

International Nuclear Information System (INIS)

Igarashi, Noboru.

1986-01-01

Purpose: To suppress soluble radioactive corrosion products in a feedwater device. Method: In a light water cooled nuclear reactor, an iron injection system is connected to feedwater pipeways and the iron concentration in the feedwater or reactor coolant is adjusted between twice and ten times of the nickel concentration. When the nickel/iron ratio in the reactor coolant or feedwater goes nearer to 1/2, iron ions are injected together with iron particles to the reactor coolant to suppress the leaching of stainless steels, decrease the nickel in water and increase the iron concentration. As a result, it is possible to suppress the intrusion of nickel as one of parent nuclide of radioactive nuclides. Further, since the iron particles intruded into the reactor constitute nuclei for capturing the radioactive nuclides to reduce the soluble radioactive corrosion products, the radioactive nuclides deposited uniformly to the inside of the pipeways in each of the coolant circuits can be reduced. (Kawakami, Y.)

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)

Feedwater control system in nuclear power plants

International Nuclear Information System (INIS)

Masuyama, Hideo.

1981-01-01

Purpose: To enable switching operation for feedwater systems in a short time and with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of the fluctuations in the flow rate in the feedwater system during manual operation. Constitution: In a BWR type nuclear power plant having a plurality of feedwater systems to a nuclear reactor, a feedwater control system is constituted with a reactor water level controller, a M/A switcher for switching either of automatic flow rate demand signals or manual flow rate set signals from the reactor level controller to apply flow rate demand signals for each of the feedwater systems, a calculation device for calculating the flow rate set signals in the feedwater systems during manual operation and an adder for subtracting the flow rate set signals in the manual feedwater system calculated in the calculating device from the automatic flow rate demand signals for the feedwater systems during automatic operation. This enables rapid switching for the feedwater systems with no fluctuations in the reactor water level by increasing or decreasing the flow rate in the feedwater systems during automatic operation by the amount of fluctuations in the flow rate in the feedwater systems during manual operation and compensating the effects in upon manual and automatic switching by the M/A switcher. (Seki, T.)

Estimation of residual stress distribution for pressurizer nozzle of Kori nuclear power plant considering safe end

Energy Technology Data Exchange (ETDEWEB)

Song, Tae Kwang; Bae, Hong Yeol; Chun, Yun Bae; Oh, Chang Young; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Lee, Kyoung Soo; Park, Chi Yong [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

2008-08-15

In nuclear power plants, ferritic low alloy steel nozzle was connected with austenitic stainless steel piping system through alloy 82/182 butt weld. Accurate estimation of residual stress for weldment is important in the sense that alloy 82/182 is susceptible to stress corrosion cracking. There are many results which predict residual stress distribution for alloy 82/182 weld between nozzle and pipe. However, nozzle and piping system usually connected through safe end which has short length. In this paper, residual stress distribution for pressurizer nozzle of Kori nuclear power plant was predicted using FE analysis, which considered safe end. As a result, existing residual stress profile was redistributed and residual stress of inner surface was decreased specially. It means that safe end should be considered to reduce conservatism when estimating the piping system.

Monitor for reactor feedwater systems

International Nuclear Information System (INIS)

Takizawa, Yoji; Tomizawa, Teruaki

1983-01-01

Purpose: To improve the reliability of operator's procedures upon occurrence of the feedwater system abnormality in a BWR type reactor by presenting the operation with effective information to avoid such abnormality. Constitution: A feedwater temperature at the reactor inlet of a reactor feedwater system measured by a temperature detector and a predetermined value for the feedwater temperature at the reactor inlet determined depending on the reactor conditions are inputted to a start-up system. The start-up system outputs a start-up signal when the difference between the inputted values exceeds a predetermined value. Then, the start-up signal is inputted to a display device where information required for the operator is displayed in the device. Thus, the information required for the operator is rapidly provided upon abnormality of the feedwater system to thereby improve the reliability of the operator's procedures. (Moriyama, K.)

Transient simulation of feedwater vaporization during a DBA LOP/LOCA using RELAP5/MOD3.1

International Nuclear Information System (INIS)

Harrell, J.R.; Fuller, R.W.

1996-01-01

The current design and testing requirements for the feedwater check valves (FWCVs) at the Grand Gulf Nuclear Station (GGNS) 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. The original design and testing 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. Given this condition, the appropriate testing criteria would be based on air with a relatively tight allowable limit. 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 leakage flow exists from the reactor vessel to the condenser through the feedwater piping during the reactor vessel blowdown phase. 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

Feedwater system in a nuclear power plant

International Nuclear Information System (INIS)

Shimizu, Tadayuki.

1975-01-01

Object: To improve the control property of a steam turbine for a feedwater pump and plant operation characteristics where water is supplied at a low rate. Structure: In a nuclear power plant where feedwater pumps of the reactor are driven by a steam turbine, the main feedwater duct on the discharge side of the feedwater pumps is provided with a cut-off valve and is connected parallel with a bypass duct having a pressure compensated flow control valve. With this arrangement, at the time when the rate of feedwater is high the cut-off valve is open so that water supplied from the feedwater pumps driven by the steam turbine is supplied through the main feedwater duct to the reactor while in case when the rate of feedwater is low the flow control valve is opened to let the water be supplied through the bypass duct. (Kamimura, M.)

Effects of Inner Surface Roughness and Asymmetric Pipe Flow on Accuracy of Profile Factor for Ultrasonic Flow Meter

International Nuclear Information System (INIS)

Michitsugu Mori; Kenichi Tezuka; Yasushi Takeda

2006-01-01

Flow profile factors (PFs), which adjust measurements to real flow rates, also strongly depend on flow profiles. To determine profile factors for actual power plants, manufactures of flowmeters usually conduct factory calibration tests under ambient flow conditions. Indeed, flow measurements with high accuracy for reactor feedwater require them to conduct calibration tests under real conditions, such as liquid conditions and piping layouts. On the contrary, as nuclear power plants are highly aging, readings of flowmeters for reactor feedwater systems drift due to the changes of flow profiles. The causes of those deviations are affected by the change of wall roughness of inner surface of pipings. We have conducted experiments to quantify the effects of flow patterns on the PFs due to pipe roughness and asymmetric flow, and the results of our experiments have shown the effects of elbows and pipe inner roughness, which strongly affect to the creation of the flow patterns. Those changes of flow patterns lead to large errors in measurements with transit time (time-of-flight: TOF) ultrasonic flow meters. In those experiments, changes of pipe roughness result in the changes of PFs with certain errors. Therefore, we must take into account those effects in order to measure the flow rates of feedwater with better accuracy in actual power plants. (authors)

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.)

Protection method and protection device for liquid supply channel for nuclear reactor

International Nuclear Information System (INIS)

Sato, Masanori; Fujimoto, Sachiko

1998-01-01

In the present invention, thermal stresses exerted on feedwater pipelines and a feedwater nozzle portion in a LBR type reactor are reduced to improve integrity of a reactor, suppress addition of facilities and reliably reduce thermal stresses by a simple structure. A connection pipe channel is formed between the upper side of a horizontal portion of a pipeline for feeding water to a reactor pressure vessel and a vertical portion of the feedwater pipeline. A transferring pump is disposed in the midway of the connection pipe channel for sucking supernatant water in the horizontal portion and rendering it to join with water in the vertical portion. When supply of water is stagnated, high temperature water in the horizontal portion is transferred by the action of the transfer pump to the low temperature water in the vertical portion to join them. With such procedures, the water supplied to the horizontal portion in the feedwater pipeline is flown to suppress occurrence of heat-stratification phenomenon thereby enabling to reduce the temperature difference. (T.M.)

Calibration of nozzle for air mass flow measurement

Science.gov (United States)

Uher, Jan; Kanta, Lukáš

2017-09-01

The effort to make calibration measurement of mass flow through a nozzle was not satisfying. Traversing across the pipe radius with Pitot probe was done. The presence of overshoot behind the bend in the pipe was found. The overshoot led to an asymmetric velocity profile.

Method of preventing sodium from flowing when pipes of a fast breeder reactor are injured

International Nuclear Information System (INIS)

Nakai, Yasushi; Yamagishi, Yoshiaki; Koga, Tomonari.

1975-01-01

Object: To inject high pressure sodium into an inlet nozzle portion when fluid pressure in the inlet nozzle portion of a core cooling pipe on the inlet side is in an abnormal condition, to thereby quickly and positively prevent the flow of sodium in a high pressure chamber in a reactor vessel, when pipes are injured. Structure: When the core cooling pipe on the inlet side is injured and as a consequence the pressure gage detects an abnormal condition of fluid pressure in the inlet nozzle, the valve is opened to allow high pressure sodium to inject into the inlet nozzle through a high pressure sodium supply pipe, thereby blocking a back-flow of sodium in the high pressure chamber into the core cooling pipe. (Kamimura, M.)

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.

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)

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...

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

Condensation driven water hammer studies for feed water distribution pipe

International Nuclear Information System (INIS)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Longvinov, S.A.; Trunov, N.B.; Sitnik, Yu.K.

1997-01-01

Special T-shaped feedwater distribution pipes were installed in steam generators at the Loviisa (Finland) and Rovno (Russia) nuclear power plants. The new shape was tested in an extensive testing programme. Since the tubes frequently suffer from corrosion damage, large-scale water hammer experiments were performed on a model facility in 1996. The main objectives of the water hammer experiments were to find out the prevailing parameters leading to water hammers, as well as the sensitivity of hammering to boundary conditions. A water hammer may occur when the mass flow rate into the steam generator exceeds 6 kg/s and the temperature difference between steam generator and feedwater exceeds 100 degC. Visual experiments and stress analyses of the pipe were also carried out. The weakest part, the T-joint, may hold against such water hammers only for a limited time of the order of few minutes. (M.D.)

Tracer test method and process data reconciliation based on VDI 2048. Comparison of two methods for highly accurate determination of feedwater massflow at NPP Beznau

International Nuclear Information System (INIS)

Hungerbuehler, T.; Langenstein, M.

2007-01-01

The feedwater mass flow is the key measured variable used to determine the thermal reactor output in a nuclear power plant. Usually this parameter is recorded via venturi nozzles of orifice plates. The problem with both principles of measurement, however, is that an accuracy of below 1% cannot be reached. In order to make more accurate statements about the feedwater amounts recirculated in the water-steam cycle, tracer measurements that offer an accuracy of up to 0.2% are used. In the NPP Beznau both methods have been used in parallel to determine the feedwater flow rates in 2004 (unit 1) and 2005 (unit 2). Comparison of the results shows that a high level of agreement is obtained between the results of the reconciliation and the results of the tracer measurements. As a result of the findings of this comparison, a high level of acceptance of process data reconciliation based on VDI 2048 was achieved. (orig.)

Applications of equivalent linearization approaches to nonlinear piping systems

International Nuclear Information System (INIS)

Park, Y.; Hofmayer, C.; Chokshi, N.

1997-01-01

The piping systems in nuclear power plants, even with conventional snubber supports, are highly complex nonlinear structures under severe earthquake loadings mainly due to various mechanical gaps in support structures. Some type of nonlinear analysis is necessary to accurately predict the piping responses under earthquake loadings. The application of equivalent linearization approaches (ELA) to seismic analyses of nonlinear piping systems is presented. Two types of ELA's are studied; i.e., one based on the response spectrum method and the other based on the linear random vibration theory. The test results of main steam and feedwater piping systems supported by snubbers and energy absorbers are used to evaluate the numerical accuracy and limitations

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

Water inlet and steam outlet pipes fitted one inside the other for nuclear reactors

International Nuclear Information System (INIS)

Mc Donald, B.N.

1976-01-01

A description is given of a combined exhaust nozzle and intake pipe system to support a heat exchanger inside a nuclear reactor pressure vessel. It comprises a generally cylindrical part on the exhaust nozzle, the cylindrical part having an inside passage, a flange around the passage and provided with means to secure the exhaust nozzle to the reactor pressure vessel so as to make it fluidtight. The cylindrical part has an aperture inside to take the intake pipe inside the passage so as to enable the intake pipe to project into the heat exchanger. A collar made on the heat exchanger projects from the heat exchanger to the cylindrical nozzle component to establish communication with the inside passage for the fluid [fr

Development of a multi-path ultrasonic flow meter for the application to feedwater flow measurement in nuclear power plants

International Nuclear Information System (INIS)

Jong, J. C.; Ha, J. H.; Kim, Y. H.; Jang, W. H.; Park, K. S.; Park, M. S.; Park, M. H.

2002-01-01

In this work, we propose a method to measure the feedwater flow using multi-path ultrasonic flow meter (UFM). Since the UFM measures a path velocity at which the ultrasonic wave is propagated, the flow profile may be important to convey the path velocity to the velocity averaged over the entire cross section of the flowing medium. The conventional UFM has used the smooth-wall circular pipe model presented by Nikurades. However, this model covers a lower range which is less than 3.2 million while the Reynolds number of the feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we feedwater flow in operating nuclear power plants (NPPs) is about 20 million. Therefore, we proposed the non-linear correlation model that combines the ratio between the DP output and proposed the non-linear correlation model that combines the ratio between the DP output and UFM output. Experiments were performed using both computer simulation and newly constructed NPPs' test data. The uncertainty analysis result shows that the proposed method has reasonably lower uncertainty than conventional UFM

Feedwater processing method in a boiling water reactor

Energy Technology Data Exchange (ETDEWEB)

Izumitani, M; Tanno, K

1976-09-06

The purpose of the invention is to decrease a quantity of corrosion products moving from the feedwater system to the core. Water formed into vapor after heated in a reactor is fed to the turbine through a main steam line to drive a generator to return it to liquid-state water in a condenser. The water is then again cycled into the reactor via the condensate pump, desalting unit, low pressure feedwater heater, medium pressure feedwater heater, and high pressure feedwater heater. The reactor water is recycled by a recycling pump. At this time, the reactor water recycled by the recycling pump is partially poured into a middle point between the desalting unit and the low pressure feedwater heater through a reducing valve or the like. With the structure described above, the quantity of the corrosion products from the feedwater system may be decreased by the function of a large quantity of active oxygen contained in the reactor water.

A study for a guide chart of lower and upper boundary regions to avoid the condensation-induced water hammer in a long horizontal pipe

International Nuclear Information System (INIS)

Lee, Byung Jin

1995-02-01

Effects of the key system parameters such as the pipe length, the pipe diameter, the feedwater temperature and the system pressure on the critical flow rates of both the upper and the lower boundaries have been examined for long horizontal pipes. The upper and lower critical flow rates are sensitive to the pipe diameter, the pipe length and the system pressure, but not to the feedwater temperature over the practical operating ranges. Guide charts of the CIWH region boundary have been developed to be used in the system design and operation to predict the operating conditions vulnerable to the CIWH. The charts illustrate a series of the operating ranges bounded by the lower and the upper limiting curves where the water hammer is very likely to occur. A design and operational procedure has also been provided to help the designer and the operator to avoid the CIWH

San Onofre/Zion auxiliary feedwater system seismic fault tree modeling

International Nuclear Information System (INIS)

Najafi, B.; Eide, S.

1982-02-01

As part of the study for the seismic evaluation of the San Onofre Unit 1 Auxiliary Feedwater System (AFWS), a fault tree model was developed capable of handling the effect of structural failure of the plant (in the event of an earthquake) on the availability of the AFWS. A compatible fault tree model was developed for the Zion Unit 1 AFWS in order to compare the results of the two systems. It was concluded that if a single failure of the San Onofre Unit 1 AFWS is to be prevented, some weight existing, locally operated locked open manual valves have to be used for isolation of a rupture in specific parts of the AFWS pipings

Preliminary design of RDE feedwater pump impeller

International Nuclear Information System (INIS)

Sri Sudadiyo

2018-01-01

Nowadays, pumps are being widely used in the thermal power generation including nuclear power plants. Reaktor Daya Experimental (RDE) is a proposed nuclear reactor concept for the type of nuclear power plant in Indonesia. This RDE has thermal power 10 MW th , and uses a feedwater pump within its steam cycle. The performance of feedwater pump depends on size and geometry of impeller model, such as the number of blades and the blade angle. The purpose of this study is to perform a preliminary design on an impeller of feedwater pump for RDE and to simulate its performance characteristics. The Fortran code is used as an aid in data calculation in order to rapidly compute the blade shape of feedwater pump impeller, particularly for a RDE case. The calculations analyses is solved by utilizing empirical correlations, which are related to size and geometry of a pump impeller model, while performance characteristics analysis is done based on velocity triangle diagram. The effect of leakage, pass through the impeller due to the required clearances between the feedwater pump impeller and the volute channel, is also considered. Comparison between the feedwater pump of HTR-10 and of RDE shows similarity in the trend line of curve shape. These characteristics curves will be very useful for the values prediction of performance of a RDE feedwater pump. Preliminary design of feedwater pump provides the size and geometry of impeller blade model with 5-blades, inlet angle 14.5 degrees, exit angle 25 degrees, inside diameter 81.3 mm, exit diameter 275.2 mm, thickness 4.7 mm, and height 14.1 mm. In addition, the optimal values of performance characteristics were obtained when flow capacity was 4.8 kg/s, fluid head was 29.1 m, shaft mechanical power was 2.64 kW, and efficiency was 52 % at rotational speed 1750 rpm. (author)

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)

Structural evaluation of IEA-R1 primary system pump nozzles

International Nuclear Information System (INIS)

Fainer, Gerson; Faloppa, Altair A.; Oliveira, Carlos A. de; Mattar Neto, Miguel

2017-01-01

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)

ESBWR power maneuvering via feedwater temperature control

International Nuclear Information System (INIS)

Saha, P.; Marquino, W.; Tucker, L. J.

2008-01-01

The ESBWR is a Generation III+ Boiling Water Reactor (BWR) driven by natural circulation. For a given geometry/hardware, system pressure, downcomer water level and feedwater temperature, the core flow rate in the ESBWR is only a function of reactor power, controlled through the control blade movement. In order to provide operational flexibility, another method of core-wide or global power maneuvering via feedwater temperature control has been developed. This is independent of power maneuvering via control blade movement, and it lowers the linear heat generation rate (LHGR) changes near the tip of control blades, which improves fuel reliability. All required stability, anticipated operational occurrences (AOOs), infrequent events, special events including anticipated transients without scram (ATWS), and loss-of-coolant accident (LOCA) analyses have been performed for the 4500 MWt ESBWR. Based on the results of these analyses at 'high', nominal and 'low' feedwater temperatures, a safe Power - Feedwater Temperature operating domain has been developed. This paper summarizes the results of these analyses and presents the ESBWR Power - Feedwater Temperature operating domain or map. (authors)

MAXIMUM AIR SUCTION INTO HORIZONTAL OPEN ENDED CYLINDRICAL LOUVERED PIPE

Directory of Open Access Journals (Sweden)

SAMEER RANJAN SAHU

2017-02-01

Full Text Available The main approach behind the present numerical investigation is to estimate the mass flow rate of air sucked into a horizontal open-ended louvered pipe from the surrounding atmosphere. The present numerical investigation has been performed by solving the conservation equations for mass, momentum and energy along with two equation based k-ɛ model for a louvered horizontal cylindrical pipe by finite volume method. It has been found from the numerical investigation that mass suction rate of air into the pipe increases with increase in louvered opening area and the number of nozzles used. Keeping other parameters fixed, for a given mass flow rate there exists an optimum protrusion of nozzle for highest mass suction into the pipe. It was also found from the numerical investigation that increasing the pipe diameter the suction mass flow rate of air was increased.

A Basic Study on the Ejection of ICI Nozzle under Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Cho, Jong Rae; Bae, Ji Hoon; Bang, Kwang Hyun [Korea Maritime and Ocean University, Busan (Korea, Republic of); Park, Jong Woong [Dongguk University, Gyeongju (Korea, Republic of)

2016-05-15

Nozzle injection should be blocked because it affect to the environment if its melting core exposes outside. The purpose of this study is to carry out the thermos mechanical analysis due to debris relocation under severe accidents and to predict the nozzle ejection calculated considering the contact between the nozzle and lower head, and the supports of pipe cables. As a result of analyzing process of severe accidents, there was melting reaction between nozzle and the lower head. In this situation, we might predict the non-uniform contact region of nozzle hole of lower head and nozzle outside, delaying ejection of nozzles. But after melting, the average remaining length of the nozzle was 120mm and the maximum vertical displacement of lower nozzle near the weld is 3.3mm so there would be no nozzle this model, because the cable supports restrains the vertical displacement of nozzle.

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

Lead corrosion and transport in simulated secondary feedwater

International Nuclear Information System (INIS)

McGarvey, G.B.; Ross, K.J.; McDougall, T.E.; Turner, C.W.

1998-01-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

Lead corrosion and transport in simulated secondary feedwater

Energy Technology Data Exchange (ETDEWEB)

McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

1998-07-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of,lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is due to low levels of lead although few if any failures have been experimentally linked to lead when sub-parts per billion levels are present in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the three principal corrosion products in the secondary feedwater; magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values less than that of the feedwater (9-10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces following different treatment conditions will be used to propose a model for tile transport and probable fate of lead in the secondary feedwater system. (author)

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

Nuclear power plants

International Nuclear Information System (INIS)

Usui, Eizo.

1981-01-01

Purpose: To prevent boiling of saturated water in the drain tank of a humidity separator by charging cooling water in the drain tank upon power decrease of a turbine. Constitution: Saturated water is separated from high pressure turbine exhausts in a humidity separator and stored in a drain tank. The saturated water in the drain tank is controlled to a constant level and the excess water is sent to a condensator and a feedwater heater. A cooling water feed pipe is branched as a cooling water feed pipe from the exhaust side of a reactor feedwater pump and connected by way of a closing valve to a spray nozzle provided in the drain tank. While the closing valve is usually closed to keep the water level constant in the drain tank, the closing valve is opened upon sudden decrease in the turbine power to charge the condensates by way of the cooling water feed pipe to the drain tank. Thus, the saturated water is mixed with the dondensates and the temperature is lowered to prevent boiling of the saturated water. (Kawakami, Y.)

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

Next-generation nozzle check valve significantly reduces operating costs

Energy Technology Data Exchange (ETDEWEB)

Roorda, O. [SMX International, Toronto, ON (Canada)

2009-01-15

Check valves perform an important function in preventing reverse flow and protecting plant and mechanical equipment. However, the variety of different types of valves and extreme differences in performance even within one type can change maintenance requirements and life cycle costs, amounting to millions of dollars over the typical 15-year design life of piping components. A next-generation non-slam nozzle check valve which prevents return flow has greatly reduced operating costs by protecting the mechanical equipment in a piping system. This article described the check valve varieties such as the swing check valve, a dual-plate check valve, and nozzle check valves. Advancements in optimized design of a non-slam nozzle check valve were also discussed, with particular reference to computer flow modelling such as computational fluid dynamics; computer stress modelling such as finite element analysis; and flow testing (using rapid prototype development and flow loop testing), both to improve dynamic performance and reduce hydraulic losses. The benefits of maximized dynamic performance and minimized pressure loss from the new designed valve were also outlined. It was concluded that this latest non-slam nozzle check valve design has potential applications in natural gas, liquefied natural gas, and oil pipelines, including subsea applications, as well as refineries, and petrochemical plants among others, and is suitable for horizontal and vertical installation. The result of this next-generation nozzle check valve design is not only superior performance, and effective protection of mechanical equipment but also minimized life cycle costs. 1 fig.

Minimum throttling feedwater control in VVER-1000 and PWR NPPs

International Nuclear Information System (INIS)

Symkin, B.E.; Thaulez, F.

2004-01-01

This paper presents an approach for the design and implementation of advanced digital control systems that use a minimum-throttling algorithm for the feedwater control. The minimum-throttling algorithm for the feedwater control, i.e. for the control of steam generators level and of the feedwater pumps speed, is applicable for NPPs with variable speed feedwater pumps. It operates in such a way that the feedwater control valve in the most loaded loop is wide open, steam generator level in this loop being controlled by the feedwater pumps speed, while the feedwater control valves in the other loops are slightly throttling under the action of their control system, to accommodate the slight loop imbalances. This has the advantage of minimizing the valve pressure losses hence minimizing the feedwater pumps power consumption and increasing the net MWe. The benefit has been evaluated for specific plants as being roughly 0.7 and 2.4 MW. The minimum throttling mode has the further advantages of lowering the actuator efforts with potential positive impact in actuator life and of minimizing the feedwater pipelines vibrations. The minimum throttling mode of operation has been developed by the Ukrainian company LvivORGRES. It has been applied with great deal of success on several VVER-1000 NPPs, six units of Zaporizhzha in Ukraine plus, with participation of Westinghouse, Kozloduy 5 and 6 in Bulgaria and South Ukraine 1 to 3 in Ukraine. The concept operates with both ON-OFF valves and true control valves. A study, jointly conducted by Westinghouse and LvivORGRES, is ongoing to demonstrate the applicability of the concept to PWRs having variable speed feedwater pumps and having, or installing, digital feedwater control, standalone or as part of a global digital control system. The implementation of the algorithm at VVER-1000 plants provided both safety improvement and direct commercial benefits. The minimum-throttling algorithm will similarly increase the performance of PWRs. The

Lead corrosion and transport in simulated secondary feedwater

Energy Technology Data Exchange (ETDEWEB)

McGarvey, G.B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Ross, K.J.; McDougall, T.E. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Turner, C.W

1999-07-01

The ubiquitous presence of lead at trace levels in secondary feedwater is a concern to all operators of steam generators and has prompted laboratory studies of its interaction with Inconel 600, Inconel 690, Monel 400 and Incoloy 800. Acute exposures of steam generator alloys to high levels of lead in the laboratory and in the field have accelerated the degradation of these alloys. There is some disagreement over the role of lead when the exposure is to chronic levels. It has been proposed that most of the present degradation of steam generator tubes is caused by low levels of lead although few, if any, failures have been experimentally linked to lead when it is present in sub-parts per billion in the feedwater. One reason for the difficulty in assigning the role of the lead is related to its possible immobilization on the surfaces of corrosion products or iron oxide films in the feedwater system. We have measured lead adsorption profiles on the 3 principal corrosion products in the secondary feedwater: magnetite, lepidocrocite and hematite. In all cases, essentially complete adsorption of the lead is achieved at pH values that are lower than the pH of the feedwater (9 to 10). If lead is maintained in this adsorbed state, it may be more chemically benign than lead that is free to dissolve in the feedwater and subsequently adsorb on steam generator tube surfaces. In this paper, we report on lead adsorption onto simulated corrosion products under simulated feedwater conditions and propose a physical model for the transport and fate of lead under operating conditions. The nature of lead adsorption onto the surfaces of different corrosion products will be discussed. The desorption behaviour of lead from iron oxide surfaces after different treatment conditions will be used to propose a model for the transport and probable fate of lead in the secondary feedwater system. (author)

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

An experimental investigation on the pressure characteristics of high speed self-resonating pulsed waterjets influenced by feeding pipe diameter

Energy Technology Data Exchange (ETDEWEB)

Li, Dong; Kang, Dong; Ding, Xiao Long; Wang, Xiao Huan; Fang, Zhen Long [School of Power and Mechanical Engineering, Wuhan University, Hubei Province (China)

2016-11-15

The destructive power of a continuous waterjet issuing from a nozzle can be greatly enhanced by generating self-resonance in the nozzle assembly to produce a Self-resonating pulsed waterjet (SRPW). To further improve the performance of SRPW, effects of feeding pipe diameter on the pressure characteristics were experimentally investigated by measuring and analyzing the axial pressure oscillation peaks and amplitudes. Four organ-pipe nozzles of different chamber lengths and three feeding pipes of different diameters were employed. Results show that feeding pipe diameter cannot change the feature of SRPW of having an optimum standoff distance, but it slightly changes the oscillating frequency of the jet. It is also found that feeding pipe diameter significantly affects the magnitudes of pressure oscillation peak and amplitude, largely depending on the pump pressure and standoff distance. The enhancement or attenuation of the pressure oscillation peak and amplitude can be differently affected by the same feeding pipe diameter.

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)

Pipe and hose decontamination apparatus

International Nuclear Information System (INIS)

Fowler, D.E.

1985-01-01

A pipe and hose decontamination apparatus is disclosed using freshly filtered high pressure Freon solvent in an integrated closed loop to remove radioactive particles or other contaminants from items having a long cylindrical geometry such as hoses, pipes, cables and the like. The pipe and hose decontamination apparatus comprises a chamber capable of accomodating a long cylindrical work piece to be decontaminated. The chamber has a downward sloped bottom draining to a solvent holding tank. An entrance zone, a cleaning zone and an exit drying zone are defined within the chamber by removable partitions having slotted rubber gaskets in their centers. The entrance and exit drying zones contain a horizontally mounted cylindrical housing which supports in combination a plurality of slotted rubber gaskets and circular brushes to initiate mechanical decontamination. Solvent is delivered at high pressure to a spray ring located in the cleaning zone having a plurality of nozzles surrounding the work piece. The solvent drains into a solvent holding tank located below the nozzles and means are provided for circulating the solvent to and from a solvent cleaning, distilling and filter unit

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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.

Aiding operator performance at low power feedwater control

International Nuclear Information System (INIS)

Woods, D.D.

1986-01-01

Control of the feedwater system during low power operations (approximately 2% to 30% power) is a difficult task where poor performance (excessive trips) has a high cost to utilities. This paper describes several efforts in the human factors aspects of this task that are underway to improve feedwater control. A variety of knowledge acquisition techniques have been used to understand the details of what makes feedwater control at low power difficult and what knowledge and skill distinguishes expert operators at this task from less experienced ones. The results indicate that there are multiple factors that contribute to task difficulty

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

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

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.

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

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

Critical flashing flows in nozzles with subcooled inlet conditions

International Nuclear Information System (INIS)

Abuaf, N.; Jones, O.C. Jr.; Wu, B.J.C.

1983-01-01

Examination of a large number of experiments dealing with flashing flows in converging and converging-diverging nozzles reveals that knowledge of the flashing inception point is the key to the prediction of critical flow rates. An extension of the static flashing inception correlation of Jones [16] and Alamgir and Lienhard [17] to flowing systems has allowed the determination of the location of flashing inception in nozzle flows with subcooled inlet conditions. It is shown that in all the experiments examined with subcooled inlet regardless of the degree of inlet subcooling, flashing inception invariably occurred very close to the throat. A correlation is given to predict flashing inception in both pipes and nozzles which matches all data available, but is lacking verification in intermediate nozzle geometries where turbulence may be important. A consequence of this behavior is that the critical mass flux may be correlated to the pressure difference between the nozzle inlet and flashing inception, through a single phase liquid discharge coefficient and an accurate prediction of the flashing inception pressure at the throat. Comparison with the available experiments indicate that the predicted mass fluxes are within 5 percent of the measurements

Secondary coolant circuit operation tests: steam generator feedwater supply

International Nuclear Information System (INIS)

Beroux, M.

1985-01-01

No one important accident occurred during the start-up tests of the 1300MWe P4 series, concerning the feedwater system of steam generators (SG). This communication comments on some incidents, that the tests allowed to detect very soon and which had no consequences on the operation of units: 1) Water hammer in feedwater tubes, and incidents met in the emergency steam generator water supply circuit. The technological differences between SG 900 and 1300 are pointed out, and the measures taken to prevent this problem are presented. 2) Incidents met on the emergency feedwater supply circuit of steam generators; mechanical or functional modifications involved by these incidents [fr

Dependence of steam generator vibrations on feedwater pressure

International Nuclear Information System (INIS)

Sadilek, J.

1989-01-01

Vibration sensors are attached to the bottom of the steam generator jacket between the input and output primary circuit collectors. The effective vibration value is recorded daily. Several times higher vibrations were observed at irregular intervals; their causes were sought, and the relation between the steam generator vibrations measured at the bottom of its vessel and the feedwater pressure was established. The source of the vibrations was found to be in the feedwater tract of the steam generator. The feedwater tract is described and its hydraulic characteristics are given. Vibrations were measured on the S02 valve. It is concluded that vibrations can be eliminated by reducing the water pressure before the control valves and by replacing the control valves with ones with more suitable control characteristics. (E.J.). 3 figs., 1 tab., 3 refs

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.

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

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

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

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)

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

Energy Technology Data Exchange (ETDEWEB)

Chen, Che-Hao; Shih, Chunkuan [National Tsing Hua Univ., Taiwan (China). Inst. of Nuclear Engineering and Science; Wang, Jong-Rong; Lin, Hao-Tzu [Atomic Energy Council, Taiwan (China). Inst. of Nuclear Energy Research

2013-07-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

Trace analysis of auxiliary feedwater capacity for Maanshan PWR loss-of-normal-feedwater transient

International Nuclear Information System (INIS)

Chen, Che-Hao; Shih, Chunkuan; Wang, Jong-Rong; Lin, Hao-Tzu

2013-01-01

Maanshan nuclear power plant is a Westinghouse PWR of Taiwan Power Company (Taipower, TPC). A few years ago, TPC has made many assessments in order to uprate the power of Maanshan NPP. The assessments include NSSS (Nuclear Steam Supply System) parameters calculation, uncertainty acceptance, integrity of pressure vessel, reliability of auxiliary systems, and transient analyses, etc. Since the Fukushima Daiichi accident happened, it is necessary to consider transients with multiple-failure. Base on the analysis, we further study the auxiliary feedwater capability for Loss-of-Normal-Feedwater (LONF) transient. LONF is the limiting transient of non-turbine trip initiated event for ATWS (Anticipated Transient Without Scram) which results in a reduction in capability of the secondary system to remove the heat generated in the reactor core. If the turbine fails to trip immediately, the secondary water inventory will decrease significantly before the actuation of auxiliary feedwater (AFW) system. The heat removal from the primary side decreases, and this leads to increases of primary coolant temperature and pressure. The water level of pressurizer also increases subsequently. The heat removal through the relief valves and the auxiliary feedwater is not sufficient to fully cope with the heat generation from primary side. The pressurizer will be filled with water finally, and the RCS pressure might rise above the set point of relief valves for water discharge. RCS pressure depends on steam generator inventory, primary coolant temperature, negative reactivity feedback, and core power, etc. The RCS pressure may reach its peak after core power reduction. According to ASME Code Level C service limit criteria, the Reactor Coolant System (RCS) pressure must be under 22.06 MPa. The USNRC is developing an advanced thermal hydraulic code named TRACE for nuclear power plant safety analysis. The development of TRACE is based on TRAC and integrating with RELAP5 and other programs. SNAP

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.)

Loss-of-feedwater transients in PWRs

International Nuclear Information System (INIS)

Burns, R.D. III.

1980-01-01

Recent severe accident sequence analysis (SASA) work in LASL's Multifault Accident Analysis Section has focused on loss-of-feedwater (LOFW) transients at a 4-loop Westinghouse nuclear power reactor. In all transients studied, the initiator was loss of main feedwater and reactor coolant pump (RCP) trip, caused by temporary loss of off-site power. Subsequent automatic actions included reactor scram, closure of the main steam isolation valves, and initiation of auxiliary feedwater (AFW) flow. TRAC-PD2 calculations were designed to study the consequences of AFW delivery rates below the minimum specified in the emergency operating procedures (EOPs) for the reference 4-loop plant. Six types of LOFW scenarios have been studied, including (1) zero AFW availability (nominal case), (2) initially zero AFW but full recovery after 2 h, (3) zero AFW with steam generator (SG) atmospheric relief valve (ARV) malfunction, (4) zero AFW with high pressure charging flow initiated after 2 h, and (5) zero AFW with delay in reactor scram. Additional cases were considered to study the effects of uncertainties in pressurizer heater/spray operation, operator manual initiation of high pressure charging flow, reactor initial conditions, and RCP and power coastdown characteristics. Nominal case results, rationale for selections of other cases, and lessons learned are summarized

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

Feedwater heater performance evaluation using the heat exchanger workstation

International Nuclear Information System (INIS)

Ranganathan, K.M.; Singh, G.P.; Tsou, J.L.

1995-01-01

A Heat Exchanger Workstation (HEW) has been developed to monitor the condition of heat exchanging equipment power plants. HEW enables engineers to analyze thermal performance and failure events for power plant feedwater heaters. The software provides tools for heat balance calculation and performance analysis. It also contains an expert system that enables performance enhancement. The Operation and Maintenance (O ampersand M) reference module on CD-ROM for HEW will be available by the end of 1995. Future developments of HEW would result in Condenser Expert System (CONES) and Balance of Plant Expert System (BOPES). HEW consists of five tightly integrated applications: A Database system for heat exchanger data storage, a Diagrammer system for creating plant heat exchanger schematics and data display, a Performance Analyst system for analyzing and predicting heat exchanger performance, a Performance Advisor expert system for expertise on improving heat exchanger performance and a Water Calculator system for computing properties of steam and water. In this paper an analysis of a feedwater heater which has been off-line is used to demonstrate how HEW can analyze the performance of the feedwater heater train and provide an economic justification for either replacing or repairing the feedwater heater

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

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%

Trace analysis of loss of feedwater flow event in Lungmen ABWR

International Nuclear Information System (INIS)

Wang Jongrong; Lin Haotzu; Wang Weichen; Yang Shuming; Shih Chunkuan

2009-01-01

TRACE (TRAC/RELAP Advanced Computational Engine) model of Lungmen Nuclear Power Plant was used to analyze the Loss of Feedwater Flow transient as defined in Lungmen FSAR Chapter 15. The results were compared with those from FSAR and RETRAN02. Lungmen TRACE model will have two models: In model A, vessel is divided into 11 axial levels, 4 radial rings and 1 azimuthal sectors; In model B, vessel is divided into 11 axial levels, 4 radial rings, and 6 azimuthal sectors. The above models include feedwater control system, narrow range water level control system, and wide range water level control system. The loss of feedwater flow (LOFW) transient began with the trip of two operating feedwater pumps either from the pump mechanical/electric failure, or the operator human error, or high water level signal. Feedwater flow was assumed to descend to 0 in 5 seconds and led to the decrease of reactor water level. At L3 low water level setpoint, the system actuated reactor scram signal and RIP trip signal for RIPs not connected to the M/G set. At L2 low-low water level setpoint, the system would trip the other six RIPs. This paper compares those important thermal parameters at steady state, such as the dome pressure and temperature of reactor vessel, steam flow, feedwater flow, core flow, and RIP flow, etc.. It also compares system parameters under transient conditions, such as core thermal power, core flow, steam flow, feedwater flow, Narrow Range Water Level (NRWL), Wide Range Water Level (WRWL) and RIP flow, etc.. It was concluded that the steady state and transient results of TRACE calculations are in good agreement with those from RETRAN02. In summary, our studies concluded that Lungmen TRACE model is correct and accurate enough for future safety analysis applications. (author)

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.)

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

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

International Nuclear Information System (INIS)

Seeberger, E.K.

2000-01-01

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

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)

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)

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

Boiling water reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Inoue, Kotaro; Ishida, Masayoshi.

1975-01-01

Object: To connect a feedwater pipe to a recycling pipe line, the recycling pipe line being made smaller in diameter, thereby minimizing loss of coolant resulting from rupture of the pipe and improving safety against trouble of coolant loss. Structure: A feedwater pipe is directly connected to a recycling pipe line before a booster pump, and a mixture of recycling water and feedwater is increased in pressure by the booster pump, after which it is introduced into a jet pump in the form of water for driving the jet pump to suck surrounding water causing it to be flown into the core. In accordance with the abovementioned structure, since the flow of feedwater can be used as a part of water for driving the jet pump, the flow within the recycling pipe line may be decreased so that the recycling pipe line can be made smaller in diameter to reduce the flow of coolant in the reactor, which flows out when the pipe is ruptured. (Furukawa, Y.)

Manual for investigation and correction of feedwater heater failures

International Nuclear Information System (INIS)

Bell, R.J.; Diaz-Tous, I.A.; Bartz, J.A.

1993-01-01

The Electric Power Research Institute (EPRI) has sponsored the development of a recently published manual which is designed to assist utility personnel in identifying and correcting closed feedwater heater problems. The main portion of the manual describes common failure modes, probable means of identifying root causes and appropriate corrective actions. These include materials selection, fabrication practices, design, normal/abnormal operation and maintenance. The manual appendices include various data, intended to aid those involved in monitoring and condition assessment of feedwater heaters. This paper contains a detailed overview of the manual content and suggested means for its efficient use by utility engineers and operations and maintenance personnel who are charged with the responsibilities of performing investigations to identify the root cause(s) of closed feedwater problems/failures and to provide appropriate corrective actions. 4 refs., 3 figs., 2 tabs

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

Thermalhydraulic study of a stratified flow in a piping elbow (Application to the model Coufast)

International Nuclear Information System (INIS)

Peniguel, C.; Stephan, J.M.

1992-11-01

In PWR's, mechanical damages (cracks) have been detected at the internal faces of steam generator feedwater piping and also in dead legs, when thermal stratification occurs. To gain some understanding on these issues, experimental and numerical programs have been set up at EDF. This paper reports a thermalhydraulic study of an elbow geometry under operating conditions leading to the establishment of a stable stratified flow. Results obtained with ESTET (a three dimensional finite differences-finite volume code solving the averaged Navier-Stokes equations) and comparisons with experimental data obtained on COUFAST (an analytical mock up, scale 1 of a French 900-MW PWR steam generator pipe elbow) are shown

Heat pipes to reduce engine exhaust emissions

Science.gov (United States)

Schultz, D. F. (Inventor)

1984-01-01

A fuel combustor is presented that consists of an elongated casing with an air inlet conduit portion at one end, and having an opposite exit end. An elongated heat pipe is mounted longitudinally in the casing and is offset from and extends alongside the combustion space. The heat pipe is in heat transmitting relationship with the air intake conduit for heating incoming air. A guide conduit structure is provided for conveying the heated air from the intake conduit into the combustion space. A fuel discharge nozzle is provided to inject fuel into the combustion space. A fuel conduit from a fuel supply source has a portion engaged in heat transfer relationship of the heat pipe for preheating the fuel. The downstream end of the heat pipe is in heat transfer relationship with the casing and is located adjacent to the downstream end of the combustion space. The offset position of the heat pipe relative to the combustion space minimizes the quenching effect of the heat pipe on the gaseous products of combustion, as well as reducing coking of the fuel on the heat pipe, thereby improving the efficiency of the combustor.

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

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

Nozzle airfoil having movable nozzle ribs

Science.gov (United States)

Yu, Yufeng Phillip; Itzel, Gary Michael

2002-01-01

A nozzle vane or airfoil structure is provided in which the nozzle ribs are connected to the side walls of the vane or airfoil in such a way that the ribs provide the requisite mechanical support between the concave side and convex side of the airfoil but are not locked in the radial direction of the assembly, longitudinally of the airfoil. The ribs may be bi-cast onto a preformed airfoil side wall structure or fastened to the airfoil by an interlocking slide connection and/or welding. By attaching the nozzle ribs to the nozzle airfoil metal in such a way that allows play longitudinally of the airfoil, the temperature difference induced radial thermal stresses at the nozzle airfoil/rib joint area are reduced while maintaining proper mechanical support of the nozzle side walls.

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.)

Review of the Shearon Harris Unit 1 auxiliary feedwater system reliability analysis

International Nuclear Information System (INIS)

Fresco, A.; Youngblood, R.; Papazoglou, I.A.

1986-02-01

This report presents the results of a review of the Auxiliary Feedwater System Reliability Analysis for the Shearon Harris Nuclear Power Plant (SHNPP) Unit 1. The objective of this report is to estimate the probability that the Auxiliary Feedwater System will fail to perform its mission for each of three different initiators: (1) loss of main feedwater with offsite power available, (2) loss of offsite power, (3) loss of all ac power except vital instrumentation and control 125-V dc/120-V ac power. The scope, methodology, and failure data are prescribed by NUREG-0611 for other Westinghouse plants

Nozzle evaluation for Project W-314

International Nuclear Information System (INIS)

Galbraith, J.D.

1998-01-01

Revisions to the waste transfer system piping to be implemented by Project W-314 will eliminate the need to access a majority of interfarm jumper connections associated with specific process pits. Additionally, connections that formerly facilitated waste transfers from the Plutonium-Uranium Extraction (PUREX) Plant are no longer required. This document identified unneeded process pit jumper connections, describes former designated routing, denotes current status (i.e., open or blanked), and recommends appropriate disposition for all. Blanking of identified nozzles should be accomplished by Project W-314 upon installation of jumpers and acceptance by Tank Waste Remediation System (TWRS) Tank Farm Operations

Loss of feedwater heater analysis for the South Texas Project

International Nuclear Information System (INIS)

Joyce, K.C.; Johnson, M.R.; Albury, C.R.

1987-01-01

The results of the steady state and transient analyses of the low pressure feedwater heater train for the South Texas Nuclear Project are presented. The South Texas Project consists of two 1250 MW Westinghouse PWR units. This analysis was performed using the Modular Modeling System (MMS) simulation code. The model presented will be incorporated into the secondary side model in support of the plant training simulator and the analysis of secondary side transients. Results of this analysis are considered preliminary until benchmarked against actual plant data. A model description of the feedwater heater train from the condensate pumps to the deaerator is presented. The methodology used to develop the model is also discussed. Results of the steady state run are presented, and a transient, the loss of extraction steam to feedwater heater 15A, is examined

Feedwater connection repair and modification at GKN

Energy Technology Data Exchange (ETDEWEB)

Witteman, C; Klees, J E

1985-03-01

From January to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing.

Feedwater connection repair and modification at GKN

International Nuclear Information System (INIS)

Witteman, C.; Klees, J.E.

1985-01-01

From Jan. to March 1983 the feedwater connection of GKN was repaired using a boring lathe, spark machining and semi-automatic welding. Nondestructive examination was performed by ultrasonic and eddy-current testing

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.

Crack initiation through vibration fatigue of small-diameter pipes

International Nuclear Information System (INIS)

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

2002-01-01

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

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.

Grain size control method for the nozzles of AP1000 primary coolant pipes

Energy Technology Data Exchange (ETDEWEB)

Wang, Shenglong [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Sun, Yanhui [Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Yang, Bin, E-mail: byang@ustb.edu.cn [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China); Collaborative Innovation Center of Steel Technology, University of Science & Technology Beijing, Beijing 100083 (China); Zhang, Mingxian [State Key Laboratory for Advanced Metals and Materials, University of Science & Technology Beijing, Beijing 100083 (China)

2017-04-01

Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

Grain size control method for the nozzles of AP1000 primary coolant pipes

International Nuclear Information System (INIS)

Wang, Shenglong; Sun, Yanhui; Yang, Bin; Zhang, Mingxian

2017-01-01

Highlights: • Design a new forging technology for AP1000 primary coolant pipe. • Method combining FEM and scale-down experiments is adopted. • The grain size and distribution in simulation and experiment are consistent. • Get optimal forging parameters for production guiding. - Abstract: AP1000 primary coolant pipe is made of 316LN austenitic stainless steel. It is a large special-shaped pipe manufactured by integral forging technology. Owing to non-uniform temperature and deformation during forging, coarse grains often occur in the boss sections of the pipe especially in the nozzles’ parts. In the present study, a new forging technology was proposed to control the grain size. The finite element method was used to optimize the forging speed and friction coefficient, then the scale-down experiments were performed for comparison. The forging speed is suggested to be less than 20 mm/s, and effective lubricants should be used to decrease the friction coefficient. The errors of the grain size between the experiment and simulation are less than 20%.

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

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

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

International Nuclear Information System (INIS)

Rahardjo, H.P.

2011-01-01

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. (author)

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.))

Thermal Performance and Operation Limit of Heat Pipe Containing Neutron Absorber

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Jeong, Yeong Shin; Kim, In Guk; Bang, In Choel [UNIST, Ulsan (Korea, Republic of)

2015-05-15

Recently, passive safety systems are under development to ensure the core cooling in accidents involving impossible depressurization such as station blackout (SBO). Hydraulic control rod drive mechanisms, passive auxiliary feedwater system (PAFS), Passive autocatalystic recombiner (PAR), and so on are types of passive safety systems to enhance the safety of nuclear power plants. Heat pipe is used in various engineering fields due to its advantages in terms of easy fabrication, high heat transfer rate, and passive heat transfer. Also, the various concepts associated with safety system and heat transfer using the heat pipe were developed in nuclear engineering field.. Thus, our group suggested the hybrid control rod which combines the functions of existing control rod and heat pipe. If there is significant temperature difference between active core and condenser, the hybrid control rod can shutdown the nuclear fission reaction and remove the decay heat from the core to ultimate heat sink. The unique characteristic of the hybrid control rod is the presence of neutron absorber inside the heat pipe. Many previous researchers studied the effect of parameters on the thermal performance of heat pipe. However, the effect of neutron absorber on the thermal performance of heat pipe has not been investigated. Thus, the annular heat pipe which contains B{sub 4}C pellet in the normal heat pipe was prepared and the thermal performance of the annular heat pipe was studied in this study. Hybrid control rod concept was developed as a passive safety system of nuclear power plant to ensure the safety of the reactor at accident condition. The hybrid control rod must contain the neutron absorber for the function as a control rod. So, the effect of neutron absorber on the thermal performance of heat pipe was experimentally investigated in this study. Temperature distributions at evaporator section of annular heat pipe were lower than normal heat pipe due to the larger volume occupied by

A probabilistic evaluation of the Shearon Harris Nuclear Power Plant auxiliary feedwater isolation system

International Nuclear Information System (INIS)

Anoba, R.C.

1989-01-01

This paper reports on a fault tree approach that was used to evaluate the safety significance of modifying the Shearon Harris Auxiliary Feedwater Isolation System. The design modification was a result of on-site reviews which identified a single failure in the Auxiliary Feedwater Isolation circuitry

Altitude Performance Characteristics of Tail-pipe Burner with Convergingconical Burner Section on J47 Turbojet Engine

Science.gov (United States)

Prince, William R; Mcaulay, John E

1950-01-01

An investigation of turbojet-engine thrust augmentation by means of tail-pipe burning was conducted in the NACA Lewis altitude wind tunnel. Performance data were obtained with a tail-pipe burner having a converging conical burner section installed on an axial-flow-compressor type turbojet engine over a range of simulated flight conditions and tail-pipe fuel-air ratios with a fixed-area exhaust nozzle. A maximum tail-pipe combustion efficiency of 0.86 was obtained at an altitude of 15,000 feet and a flight Mach number of 0.23. Tail-pipe burner operation was possible up to an altitude of 45,000 feet at a flight Mach number of 0.23.

Evaluation of examination techniques for ferritic stainless steel feedwater heater tubing

International Nuclear Information System (INIS)

Nugent, M.J.; Catapano, M.C.

1995-01-01

Ferritic stainless steel has been finding increased application in utility plant feedwater heaters due to good strength and corrosion resistance and absence of potential copper contamination of feedwater system. Ferritic stainless steel is highly magnetic and is generally not inspectable using conventional eddy current testing techniques. A variety of techniques have been developed for inspection of this tubing material used in typical heat exchanger applications. Through a project funded by the Empire State Electric Energy Research Corporation (ESEERCO), the evaluation of data generated by four present state of the art NDE testing techniques were evaluated on a controlled mock-up of the heater tubing with service related defects. The primary objective was to determine the strengths and limitations of each method. The testing of two in service feedwater heaters at the Consolidated Edison Company of New York, Inc. (Con Edison's) Arthur Kill Generating Station also allowed further evaluations based on actual field conditions

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

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)

Signal validation and failure correction algorithms for PWR steam generator feedwater control

International Nuclear Information System (INIS)

Nasrallah, C.N.; Graham, K.F.

1986-01-01

A critical contributor to the reliability of a nuclear power plant is the reliability of the control systems which maintain plant operating parameters within desired limits. The most difficult system to control in a PWR nuclear power plant and the one which causes the most reactor trips is the control of the feedwater flow to the steam generators. The level in the steam generator must be held within relatively narrow limits, with reactor trips set for both too high and too low a level. The steam generator level is inherently unstable in that it is an open integrator of feedwater flow steam flow mismatch. The steam generator feedwater control system relies on sensed variables in order to generate the appropriate feedwater valve control signal. In current systems, each of these sensed variables comes from a single sensor which may be a separate control sensor or one of the redundant protection sensors that is manually selected by the operator. In case this single signal is false, either due to sensor malfunction or due to a test signal being substituted during periodic test and maintenance, the control system will generate a wrong control signal to the feedwater control valve. This will initiate a steam generator level upset. The solution to this problem is for the control system to sense a given variable with more than one redundant sensor. Normally there are three or four sensors for each variable monitored by the reactor protection system. The techniques discussed allow the control system to compare these redundant sensor signals and generate a validated signal for each measured variable that is insensitive to false signals

Aging assessment of auxiliary feedwater systems

International Nuclear Information System (INIS)

Casada, D.A.

1989-01-01

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 study has reviewed historical failure experience and current monitoring practices for the AFW System. This paper provides an overview of the study approach and results. 7 figs

Development of an ultrasonic flow and temperature measurement system for pressurized water reactors

International Nuclear Information System (INIS)

James, R.W.; Lubnow, T.; Baumgart, G.; Ravetti, D.

1996-01-01

In U.S. nuclear plants, primary coolant flow and reactor thermal power are calculated from a measurement of feedwater flow to the steam generator combined with knowledge of steam generator heat transfer characteristics nd measurement of hot leg temperature by resistance temperature detectors (RTDs). The calculation of plant thermal output is complicated by an indirect measurement of primary coolant mass flow rate and thermal streaming in the region where hot leg temperature is typically measured. Uncertainty in the thermal output calculation results from uncertainties in steam generator characteristics, in the hot leg temperature due to thermal streaming, and in fouling of venturi nozzles used for feedwater flow measurement. This in turn leads to operation of power plants ar lower levels of efficiency. The Electric Power Research Institute (EPRI) has on ongoing project to develop a prototype system to directly measure primary coolant flow rate and bulk average temperature using ultrasonic transducers externally mounted on the pipe. The topic of this paper is a summary of the project experience in developing this system. The technology being developed in this project is based in part upon previously existing ultrasonic feedwater flow measurement technology developed by MPR Associates and Caldon, Inc EPRI is a non-profit company performing research for U.S. and international electric power utilities. (authors)

BOA II: Asbestos Pipe-Insulation Removal Robot System. Innovative Technology Summary Report

International Nuclear Information System (INIS)

None

2001-01-01

The objective of this task is to develop and demonstrate a mechanical, asbestos-removal system that can be remotely operated without a containment area. The technology, known as BOA, consists of a pipe-crawler removal head and a boom vehicle system with dual robots. BOA's removal head can be remotely placed on the outside of the pipe and can crawl along the pipe, removing lagging and insulation. The lagging and insulation is cut using a hybrid endmill water-jet cutter and then diced into 2-inch cube sections of ACM. These ACM sections are then removed from the pipe using a set of blasting fan- spray nozzles, vacuumed off through a vacuum hose, and bagged. Careful attention to vacuum and entrapment air flow ensures that the system can operate without a containment area while meeting local and federal standards for fiber count

Excessive heat removal due to feedwater system malfunction

International Nuclear Information System (INIS)

Beader, D.; Peterlin, G.

1986-01-01

Excessive heat removal transient of the Krsko Nuclear Power Plant, caused by steam generators feedwater system malfunctions was simulated by RELAP5/MOD1 computer code. The results are increase of power and reactor scram caused by high-high steam generator level. (author)

Analysis of containment parameters during the main steam line break with the failure of the feedwater control valves

International Nuclear Information System (INIS)

Fabjan, L.; Petelin, S.; Mavko, B.; Gortnar, O.; Tiselj, I.

1992-01-01

U.S. Nuclear Regulatory Commission (NRC) information notice 91-69: 'Errors in Main Steam Line Break Analyses for Determining Containment Parameters' shows the possibility of an accident which could lead to beyond design containment pressure and temperature. Such accident would be caused by the continuation of feedwater flow following a main stream line break (MSLB) inside the containment. Krsko power plant already experienced problems with main feedwater control valves. For that reason, analysis of MSLB has been performed taking into account continuous feedwater addition scenario and different containment safety systems capabilities availability. Steam and water released into the containment during MSLB was calculated using RELAP5/MOD2 computer code. The containment response to MSLB was calculated using CONTEMPT-LT/028 computer code. The results indicated that the continuous feedwater flow following a MSLB could lead to beyond design containment pressure. The peak pressure and temperature depend on isolation time for main- and auxiliary-feedwater supply. In the case of low boron concentration injection, the core recriticality is characteristic for this type of accidents. It was concluded that the presented analysis of MSLB with continuous feedwater addition scenario is the worst case for containment design

Altitude Compensating Nozzle

Science.gov (United States)

Ruf, Joseph H.; Jones, Daniel

2015-01-01

The dual-bell nozzle (fig. 1) is an altitude-compensating nozzle that has an inner contour consisting of two overlapped bells. At low altitudes, the dual-bell nozzle operates in mode 1, only utilizing the smaller, first bell of the nozzle. In mode 1, the nozzle flow separates from the wall at the inflection point between the two bell contours. As the vehicle reaches higher altitudes, the dual-bell nozzle flow transitions to mode 2, to flow full into the second, larger bell. This dual-mode operation allows near optimal expansion at two altitudes, enabling a higher mission average specific impulse (Isp) relative to that of a conventional, single-bell nozzle. Dual-bell nozzles have been studied analytically and subscale nozzle tests have been completed.1 This higher mission averaged Isp can provide up to a 5% increase2 in payload to orbit for existing launch vehicles. The next important step for the dual-bell nozzle is to confirm its potential in a relevant flight environment. Toward this end, NASA Marshall Space Flight Center (MSFC) and Armstrong Flight Research Center (AFRC) have been working to develop a subscale, hot-fire, dual-bell nozzle test article for flight testing on AFRC's F15-D flight test bed (figs. 2 and 3). Flight test data demonstrating a dual-bell ability to control the mode transition and result in a sufficient increase in a rocket's mission averaged Isp should help convince the launch service providers that the dual-bell nozzle would provide a return on the required investment to bring a dual-bell into flight operation. The Game Changing Department provided 0.2 FTE to ER42 for this effort in 2014.

Contributions of Modranska potrubni a.s. to the safety improvement of piping systems and valves of NPS type VVER 440 and VVER 1000

International Nuclear Information System (INIS)

Slach, J.

2004-01-01

The following activities are described: (i) Installation of pipe whip restraints on piping for high pressure and temperature steam and feed piping; (ii) Installation of air receivers for quick-acting valves with air actuator on VVER 440 units at the Jaslovske Bohunice V2 NPP; (iii) Replacement of the technical water distribution system material in the reactor hall of the Temelin VVER 1000 units; Installation of measuring nozzles on main steam piping DN 600 at the Temelin VVER 1000 units. (P.A.)

An estimation of reactor thermal power uncertainty using UFM-based feedwater flow rate in nuclear power plants

International Nuclear Information System (INIS)

Byung Ryul Jung; Ho Cheol Jang; Byung Jin Lee; Se Jin Baik; Woo Hyun Jang

2005-01-01

Most of Pressurized Water Reactors (PWRs) utilize the venturi meters (VMs) to measure the feedwater (FW) flow rate to the steam generator in the calorimetric measurement, which is used in the reactor thermal power (RTP) estimation. However, measurement drifts have been experienced due to some anomalies on the venturi meter (generally called the venturi meter fouling). The VM's fouling tends to increase the measured pressure drop across the meter, which results in indication of increased feedwater flow rate. Finally, the reactor thermal power is overestimated and the actual reactor power is to be reduced to remain within the regulatory limits. To overcome this VM's fouling problem, the Ultrasonic Flow Meter (UFM) has recently been gaining attention in the measurement of the feedwater flow rate. This paper presents the applicability of a UFM based feedwater flow rate in the estimation of reactor thermal power uncertainty. The FW and RTP uncertainties are compared in terms of sensitivities between the VM- and UFM-based feedwater flow rates. Data from typical Optimized Power Reactor 1000 (OPR1000) plants are used to estimate the uncertainty. (authors)

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

Laguna Verde annulus pressurization loads evaluation

International Nuclear Information System (INIS)

Castaneda, M. A.; Cruz, M. A.; Cardenas, J. B.; Vargas, A.; Cruz, H. J.; Mercado, J. J.

2010-10-01

Annulus pressurization, jet impingement, pipe whip restraint and jet thrust are phenomena related to postulated pipe ruptures. A postulated pipe rupture at the weld between recirculation, or feedwater piping and a reactor nozzle safe end, will lead to a high flow rate of flashing water/steam mixture into the annulus between the reactor pressure vessel and the biological shield wall. The total effect of the vessel and pipe inventory blowdown from the break being postulated must be accounted for in the evaluation. A recirculation line break will give rise to an angular dependent short term pressure differential around the vessel, followed by a longer term pressure buildup in the annulus. A recirculation line postulated rupture may not produce worst case conditions and reference to time intervals for only the recirculation break should be treated superficially. A postulated rupture of the feedwater piping may produce the extreme case for determining: 1) the shield wall and reactor vessel to pedestal interactions, 2) loading on the reactor vessel internals, or 3) responses for the balance of piping attached to the vessel. Recently it was identified a potential issue regarding the criteria used to determine which cases were evaluated for Annulus Pressurization (A P) loads for new loads plants. The original A P loads methodology in the late 1970 and early 1980 years separated the mass/energy release calculation from the structural response calculation based on the implicit assumption that the maximum overall mass/energy release will result in maximizing the structural response and corresponding stresses on the reactor pressure vessel, internals, and containment structures. This process did not consider the dynamic response in the primary and secondary safety related structures, components and equipment. Consequently, the A P loads used as input for design adequacy evaluations of Nuclear Steam Supply System safety related components for new loads plants might have

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)

Auxiliary feedwater system aging study

International Nuclear Information System (INIS)

Kueck, J.D.

1992-01-01

The Phase 1 Auxiliary Feedwater (AFW) System Aging Study, NUREG/CR-5404 V1, focused on how and to what extent the various AFW system component types fail, how the failures have been and can be detected, and on the value of current testing requirements and practices. This follow-on study, which will be provided in full in NUREG/CR-5404 V2, provides a closure to the Phase 1 Study. For each of the component types and for the various sources of component failure identified in the Phase 1 Study, the methods of failure detection were designated and tabulated and the following findings became evident: Instrumentation and Control (I and C) related failures dominated the group of failures that were detected during demand conditions; many of the potential failure sources not detectable by the current monitoring practices were related to the I and C portion of the system; some component failure modes are actually aggravated by conventional test methods; and several important system functions did not undergo any function verification test. The goal of this follow-on study was to categorize and evaluate the deficiencies in testing identified by Phase 1 and to make specific recommendations for corrective action. In addition, this study presents discussions of alternate, state-of-the-art test methods, and provides a proposed Auxiliary Feedwater Pump test at normal operating pressure which should do much to verify system operability while eliminating degradation

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)

Automatic control device for the reduction of reactor power

International Nuclear Information System (INIS)

Sumida, Susumu; Mizuno, Hiroshi.

1982-01-01

Purpose: To early detect troubles in condensate pipeways and feedwater pipeways of BWR-type reactor. Constitution: Detectors are provided to a condensate pipe, a condensator, a low pressure condensate pump, a condensate desalting device and a high pressure condensate pump for constituting condensate pipeways, as well as to a feedwater heater, a feedwater pipe and a feedwater pump for constituting feedwater pipeways. Each of the detectors is connected by way of a lead wire to an abnormal detection and processing device. The abnormal detection and processing device, which are connected to a recycling control device, monitor the input from the detector and sends a control signal to the recycling control system upon calculation of a trouble signal from the detector. (Sekiya, K.)

Heat transfer pipe shielding device for heat exchanger

International Nuclear Information System (INIS)

Hanawa, Jun.

1991-01-01

The front face and the rear face of a frame that surrounds the circumference of the water chamber body of a multi-tube heat exchanger are covered by a rotational shielding plate. A slit is radially formed to the shielding plate for the insertion of a probe or cleaner to the heat transfer pipe and a deflector is disposed on the side opposite to the slit. The end of the heat transfer pipe to be inspected is exposed to the outer side by way of the slit by the rotation of the shielding plate, and the probe or cleaner is inserted in the heat transfer pipe to conduct an eddy current injury monitoring test or cleaning. The inside of the water chamber and the heat transfer pipe is exhausted by a ventilation nozzle disposed to the frame. Accordingly, a shielding effect upon inspection and cleaning can be obtained and, in addition, inspection and exhaustion at the cleaning position can be conducted easily. Since the operation for attachment and detachment is easy, the effect of reducing radiation dose per unit can be obtained by the shortening of the operation time. (N.H.)

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.

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)

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.

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)

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

International Nuclear Information System (INIS)

Rabello, Emerson G.; Silva, Luiz L.; Gomes, Paulo T.V.

2011-01-01

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)

Numerical and experimental analysis of the vibratory behavior of a nuclear power plant piping system excitated by a pump

International Nuclear Information System (INIS)

Vatin, E.; Guillou, J.; Tephany, F.; Trollat, C.

1993-08-01

This paper presents a study on the dynamic response of piping systems installed in the French 1300 MWe Nuclear Power Plants. Variations in pressure are generated by a multi-staged centrifugal pump mounted on the piping system and provide a dynamic excitation of the pipe. This type of dynamic loading has led to nozzle cracks for some of the pipes, whereas, for other installations, it has not be found detrimental. This study presents an explanation of the different dynamic behavior observed at the various plants. To this end, a numerical model, calibrated with on-site measurements, is impleted. (authors). 8 figs., 1 tab., 5 refs

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

Fuzzy Logic Approach to Diagnosis of Feedwater Heater Performance Degradation

International Nuclear Information System (INIS)

Kang, Yeon Kwan; Kim, Hyeon Min; Heo, Gyun Young; Sang, Seok Yoon

2014-01-01

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

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

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

Digital feedwater and recirculation flow control for GPUN Oyster Creek

International Nuclear Information System (INIS)

Burjorjee, D.; Gan, B.

1992-01-01

This paper describes the digital system for feedwater and recirculation control that GPU Nuclear will be installing at Oyster Creek during its next outage - expected circa December 1992. The replacement was motivated by considerations of reliability and obsolescence - the analog equipment was aging and reaching the end of its useful life. The new system uses Atomic Energy of Canada Ltd.'s software platform running on dual, redundant, industrial-grade 386 computers with opto-isolated field input/output (I/O) accessed through a parallel bus. The feedwater controller controls three main feed regulating valves, two low flow regulating valves, and two block valves. The recirculation controller drives the five scoop positioners of the hydraulic couplers. The system also drives contacts that lock up the actuators on detecting an open circuit in their current loops

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)

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)

Open channel steam generator feedwater system

International Nuclear Information System (INIS)

Kim, R.F.; Min-Hsiung Hu.

1985-01-01

A steam generator which utilizes a primary fluid to vaporize a secondary fluid is provided with an open flow channel and elevated discharge nozzle for the introduction of secondary fluid. The discharge nozzle is positioned above a portion of the inlet line such that the secondary fluid passes through a vertical section of inlet line prior to its discharge into the open channel. (author)

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. For each type of loading Ksub(I) values are given for cracks at 15 nozzle locations and for six crack depths. Reasonable agreement is noted between calculated and previously published pressure-induced Ksub(I) values. Comparisons are made to determine the effect on Ksub(I) of crack location, thermal stress, and residual stress as compared to pressure stress. For the thermal transient it is shown that Ksub(I) for small crack depths is maximized early in the transient while Ksub(I) for large cracks is maximized later, under steady state conditions. Ksub(I) computations should, therefore, be made for several transient time points and the maximum Ksub(I) for a given crack depth should be used for design analysis. It is concluded that the effects on Ksub(I) of location, thermal stresses, and residual stresses are significant and generally too complex to evalute without advanced numerical procedures. The utilized combination of finite element analysis of the uncracked structure and three-dimensional influence function analysis of the cracked structure is demonstrated

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.)

A novel feedwater system for the RETRAN model of the Palo Verde nuclear generating station

International Nuclear Information System (INIS)

Secker, P.A.; Webb, J.R.

1988-01-01

This paper presents a feedwater system model which supplies realistic boundary conditions to the RETRAN model of a Palo Verde Nuclear Generating Station reactor plant. The RETRAN thermal hydraulic code is used to analyze nuclear reactor system transients through a generalized thermal hydraulic volume/junction network. The feedwater system model is implemented using the control block modeling option available in the RETRAN code. The output of the control block model is coupled to the thermal hydraulic network by a fill junction. A forward Euler integration scheme is used by RETRAN for control block variables. The feedwater system model is formulated to allow implicit integration within the existing code framework. The potential need for small integration time steps is, therefore, alleviated. The model results are compared with test data

Extensive feedwater quality control and monitoring concept for preventing chemistry-related failures of boiler tubes in a subcritical thermal power plant

International Nuclear Information System (INIS)

Vidojkovic, Sonja; Onjia, Antonije; Matovic, Branko; Grahovac, Nebojsa; Maksimovic, Vesna; Nastasovic, Aleksandra

2013-01-01

Prevention and minimizing corrosion processes on steam generating equipment is highly important in the thermal power industry. The maintenance of feedwater quality at a level corresponding to the standards of technological designing, followed by timely respond to the fluctuation of measured parameters, has a decisive role in corrosion prevention. In this study, the comprehensive chemical control of feedwater quality in 210 MW Thermal Power Plant (TPP) was carried out in order to evaluate its potentiality to assure reliable function of the boiler and discover possible irregularity that might be responsible for frequent boiler tube failures. Sensitive on-line and off-line analytical instruments were used for measuring key and diagnostic parameters considered to be crucial for boiler safety and performances. Obtained results provided evidences for exceeded levels of oxygen, silica, sodium, chloride, sulfate, copper, and conductivity what distinctly demonstrated necessity of feedwater control improvement. Consequently, more effective feedwater quality monitoring concept was recommended. In this paper, the explanation of presumable root causes of corrosive contaminants was given including basic directions for their maintenance in proscribed limits. -- Highlights: • Feedwater quality monitoring practice in a thermal power plant has been evaluated. • The more efficient feedwater quality control have been applied. • Analysis of feedwater quality parameters has been performed. • Exceeded levels of corrosive contaminants were found. • Recommendations for their maintenance at proscribed values were given

BWR type reactors

International Nuclear Information System (INIS)

Kikuchi, Osamu.

1985-01-01

Purpose: To prevent cavitations in a recycling pump, as well as improve the safety and a reliability of a pressure vessel. Constitution: A feedwater pipeway is connected to the route between the pressure vessel and the recycling pipe and feedwater from the feedwater pipeway is directly introduced to the recycling pump. The temperature of water flowing into the recycling pump is lowered by the feedwater from the feedwater pipeway to prevent the cavitations. (Yoshino, Y.)

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)

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

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.

Effect of fuel volatility on performance of tail-pipe burner

Science.gov (United States)

Barson, Zelmar; Sargent, Arthur F , Jr

1951-01-01

Fuels having Reid vapor pressures of 6.3 and 1.0 pounds per square inch were investigated in a tail-pipe burner on an axial-flow-type turbojet engine at a simulated flight Mach number of 0.6 and altitudes from 20,000 to 45,000 feet. With the burner configuration used in this investigation, having a mixing length of only 8 inches between the fuel manifold and the flame holder, the low-vapor-pressure fuel gave lower combustion efficiency at a given tail-pipe fuel-air ratio. Because the exhaust-nozzle area was fixed, the lower efficiency resulted in lower thrust and higher specific fuel consumption. The maximum altitude at which the burner would operate was practically unaffected by the change in fuel volatility.

Analisis Termal High Pressure Feedwater Heater di PLTU PT. XYZ

Directory of Open Access Journals (Sweden)

Maria Ulfa Damayanti

2017-01-01

Full Text Available Abstrak- PT. XYZ mengoperasikan tiga unit Pembangkit Listrik Tenaga Uap (PLTU unit 3, 7 dan 8 berkapasitas 2.030 MegaWatt. Pada PLTU Paiton unit 7 dan 8 terdapat delapan buah feedwater heater yaitu empat buah Low Pressure Water Heater (LPWH, tiga buah High Pressure Water Heater (HPWH, dan sebuah dearator. Pada PLTU Paiton unit 7 dan 8 terdapat kerusakan pada HPWH 6 yang menyebabkan penurunan efisiensi dari siklus secara keseluruhan. Penurunan efisiensi dapat terjadi karena temperatur feedwater sebelum masuk ke boiler terlalu rendah, sehingga kalor yang dibutuhkan oleh boiler untuk memanaskan feedwater meningkat. Oleh karena itu konsumsi batubara akan meningkat dan menyebabkan terjadi kenaikan biaya operasional harian dalam sistem pembangkit. Dari data Divisi Produksi PT. XYZ Unit 7 dan 8 diperoleh spesifikasi HPWH 6, 7, dan 8 dan propertis fluida dalam HPWH 6, 7, dan 8. Data tersebut digunakan sebagai dasar analisis termal yang meliputi performa masing-masing HPH. Tahap selanjutnya dalam analisis termal adalah memvariasikan beban 25%, 50%, 75%, 100%, dan 105%. Tahap terakhir analisis adalah menghitung performa dengan variasi sumbatan (plug 5%, 10%, 15%, dan 20% sesuai dengan variasi beban. Hasil yang didapatkan dari penelitian tugas akhir ini adalah nilai effectiveness tertinggi tercapai pada pembebanan 100% serta menghasilkan pressure drop tertinggi pada pembebanan 105%, nilai effectiveness terbesar serta nilai pressure drop terkecil terjadi pada zona Condensing, serta sumbatan (plugging pada HPH akan menyebabkan penurunan nilai effectiveness dan kenaikan pressure drop sisi tube.

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

Process and device for the protection of steam-raising units, particularly of nuclear reactors

International Nuclear Information System (INIS)

Beyer, W.; Wieling, N.; Stellwag, B.

1986-01-01

To protect the housing against corrosion by chemical conditioning of the feedwater, the redox potential of the feedwater and the corrosion potential of at least one pipe of the pipe bundle is continuously determined during operation of the steam raising unit. With potentials indicating the danger of corrosion, the quality of the secondary water can be improved by suitable measures. (orig./HP) [de

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

No, Young Gyu; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-10-15

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.

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

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

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.

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

International Nuclear Information System (INIS)

Nurkkala, P.; Hoikkanen, J.

1997-01-01

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.)

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)

1998-12-31

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.

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

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

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...

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)

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

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

Firefighter Nozzle Reaction

DEFF Research Database (Denmark)

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

2017-01-01

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...... 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 seal

International Nuclear Information System (INIS)

Herman, R.F.

1977-01-01

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 sealing members operatively disposed between the outlet nozzle and the hoop. The sealing members are biased against the pressure vessel and the hoop and are connected by a leak restraining member establishing a leak-proof condition between the inlet and outlet coolants in the region about the outlet nozzle. Furthermore, the flexible responsiveness of the seal assures that the seal will not structurally couple the hoop to the pressure vessel

Nozzle seal

International Nuclear Information System (INIS)

Walling, G.A.

1977-01-01

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 sealing rings operatively disposed between the outlet nozzles and the hoop. The sealing rings connected by flexible members are biased against the pressure vessel and the hoop, establishing a leak-proof condition between the inlet and outlet coolants in the region about the outlet nozzle. Furthermore, the flexible responsiveness of the seal assures that the seal will not structurally couple the hoop to the pressure vessel. 4 claims, 2 figures

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

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

Identification of BWR feedwater control system using autoregressive integrated model

International Nuclear Information System (INIS)

Kanemoto, Shigeru; Andoh, Yasumasa; Yamamoto, Fumiaki; Idesawa, Masato; Itoh, Kazuo.

1983-01-01

With the view of contributing toward more reliable interpretation of noise behavior under normal operating conditions, which is essential for correct detection and/or diagnosis of incipient anomalies in nuclear power plants by noise analysis technique, studies has been undertaken of the noise behavior in a BWR feedwater control system, with use made of a multivariate autoregressive modeling technique. Noise propagation mechanisms as well as open- and closed-loop responses in the system are identified from noise data by a method in which an autoregressive integrated model is introduced. The closed-loop responses obtained with this method are compared with transient data from an actual test, and confirmed to be reliable in estimating semi-quantitative features. Other analyses performed with this model also yield results that appear most reasonable in their physical characteristics. These results have demonstrated the effectiveness of the noise analyses technique based on the autoregressive integrated model for evaluating and diagnosing the performance of feedwater control systems. (author)

Analysis of Total Loss of Feedwater for APR1400 using SPACE

Energy Technology Data Exchange (ETDEWEB)

Hong, Seong Min; Park, Seok Jeong; Park, Chan Eok; Choi, Jong Ho; Lee, Gyu Cheon [KEPCO Engineering and Construction, Deajeon (Korea, Republic of)

2016-10-15

The Total Loss of FeedWater (TLOFW) event is an accident that main feedwater and auxiliary feedwater of secondary side are not supplied to steam generators. APR1400 uses the Safety Depressurization and Vent System (SDVS) for the F and B operation and SDVS is designed to perform the rapid depressurization function of Reactor Coolant System (RCS) through the remote manual operation when TLOFW is occurred. If RCS pressure falls below a Safety Injection Pump (SIP) working pressure, it can be possible to start the F and B operation which injects SIP flow to RCS and releases the RCS vapor and two-phase flow through Pilot Operated Safety Relief Valves (POSRVs) by opening the POSRVs, and then it can be possible to remove the decay heat. The design requirement of SDVS is that the core water level should be maintained at higher than 2 feet from the top of active core during the F and B operation. The TLOFW analysis was carried out to evaluate the capability of decay heat removal for APR1400 using newly developed SPACE code. The analysis results show that the F and B operation with 2 POSRVs and 2 SIPs and the F and B operation with 4 POSRVs and 4 SIPs meet the SDVS design requirement for the fuel cladding temperature. The comparison with RELAP5 shows good agreement and it validates the applicability of SPACE code for this type of accident analysis.

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.

Advanced exhaust nozzle technology

Energy Technology Data Exchange (ETDEWEB)

Glidewell, R J; Warburton, R E

1981-01-01

Recent developments in turbine engine exhaust nozzle technology include nonaxisymmetric nozzles, thrust reversing, and thrust vectoring. Trade studies have been performed to determine the impact of these developments on the thrust-to-weight ratio and specific fuel consumption of an advanced high performance, augmented turbofan engine. Results are presented in a manner which provides an understanding of the sources and magnitudes of differences in the basic elements of nozzle internal performance and weight as they relate to conventional, axisymmetric nozzle technology. Conclusions are presented and recommendations are made with regard to future directions of advanced development and demonstration. 5 refs.

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)

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.

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)

1996-12-31

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.

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

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)

Numerical Simulation of Twin Nozzle Injectors

OpenAIRE

Milak, Dino

2015-01-01

Fuel injectors for marine applications have traditionally utilized nozzles with symmetric equispaced orifice configuration. But in light of the new marine emission legislations the twin nozzle concept has arisen. The twin nozzle differs from the conventional configuration by utilizing two closely spaced orifices to substitute each orifice in the conventional nozzle. Injector manufacturers regard twin nozzle injectors as a promising approach to facilitate stable spray patterns independent of t...

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

2014-01-01

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

Energy Technology Data Exchange (ETDEWEB)

Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-10-15

The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

Remote Visual Testing (RVT) for the diagnostic inspection of feedwater heaters

International Nuclear Information System (INIS)

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

1993-01-01

Feedwater heaters are an important component in the overall plant heat rate, reliability, availability, performance and maintenance considerations at power stations. The ability to diagnose heater problems in-situ properly can lead to: (1) Preventative plugging of damaged, but unfailed tubes; (2) In-place repair procedures; (3) Incorporation of corrective actions into replacement designs or heater/unit operations. The benefits and limitations of Non-Destructive Testing (NDT) on feedwater heaters are briefly reviewed. All Remote Visual Testing (RVT) including borescopes, fiberscopes, videoborescopes and Closed Circuit Television (CCTV) cameras are discussed along with currently accepted formats for documentation. The benefits of a comprehensive in-place inspection involving Remote Visual Testing are discussed in relationship to its diagnostic capabilities. The results of eight post-service heater inspections are discussed along with the root cause of failure of seven unique failure mechanisms. These inspections, including FWH access, RVT tool and data analysis, are detailed. 13 figs

Equivalent nozzle in thermomechanical problems

International Nuclear Information System (INIS)

Cesari, F.

1977-01-01

When analyzing nuclear vessels, it is most important to study the behavior of the nozzle cylinder-cylinder intersection. For the elastic field, this analysis in three dimensions is quite easy using the method of finite elements. The same analysis in the non-linear field becomes difficult for designs in 3-D. It is therefore necessary to resolve a nozzle in two dimensions equivalent to a 3-D nozzle. The purpose of the present work is to find an equivalent nozzle both with a mechanical and thermal load. This has been achieved by the analysis in three dimensions of a nozzle and a nozzle cylinder-sphere intersection, of a different radius. The equivalent nozzle will be a nozzle with a sphere radius in a given ratio to the radius of a cylinder; thus, the maximum equivalent stress is the same in both 2-D and 3-D. The nozzle examined derived from the intersection of a cylindrical vessel of radius R=191.4 mm and thickness T=6.7 mm with a cylindrical nozzle of radius r=24.675 mm and thickness t=1.350 mm, for which the experimental results for an internal pressure load are known. The structure was subdivided into 96 finite, three-dimensional and isoparametric elements with 60 degrees of freedom and 661 total nodes. Both the analysis with a mechanical load as well as the analysis with a thermal load were carried out on this structure according to the Bersafe system. The thermal load consisted of a transient typical of an accident occurring in a sodium-cooled fast reactor, with a peak of the temperature (540 0 C) for the sodium inside the vessel with an insulating argon temperature constant at 525 0 C. The maximum value of the equivalent tension was found in the internal area at the union towards the vessel side. The analysis of the nozzle in 2-D consists in schematizing the structure as a cylinder-sphere intersection, where the sphere has a given relation to the

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

Feedwater heater tube-to-tubesheet connections

International Nuclear Information System (INIS)

Yokell, S.

1993-01-01

This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

Arcjet nozzle area ratio effects

Science.gov (United States)

Curran, Francis M.; Sarmiento, Charles J.; Birkner, Bjorn W.; Kwasny, James

1990-01-01

An experimental investigation was conducted to determine the effect of nozzle area ratio on the operating characteristics and performance of a low power dc arcjet thruster. Conical thoriated tungsten nozzle inserts were tested in a modular laboratory arcjet thruster run on hydrogen/nitrogen mixtures simulating the decomposition products of hydrazine. The converging and diverging sides of the inserts had half angles of 30 and 20 degrees, respectively, similar to a flight type unit currently under development. The length of the diverging side was varied to change the area ratio. The nozzle inserts were run over a wide range of specific power. Current, voltage, mass flow rate, and thrust were monitored to provide accurate comparisons between tests. While small differences in performance were observed between the two nozzle inserts, it was determined that for each nozzle insert, arcjet performance improved with increasing nozzle area ratio to the highest area ratio tested and that the losses become very pronounced for area ratios below 50. These trends are somewhat different than those obtained in previous experimental and analytical studies of low Re number nozzles. It appears that arcjet performance can be enhanced via area ratio optimization.

Arcjet Nozzle Area Ratio Effects

Science.gov (United States)

Curran, Francis M.; Sarmiento, Charles J.; Birkner, Bjorn W.; Kwasny, James

1990-01-01

An experimental investigation was conducted to determine the effect of nozzle area ratio on the operating characteristics and performance of a low power dc arcjet thruster. Conical thoriated tungsten nozzle inserts were tested in a modular laboratory arcjet thruster run on hydrogen/nitrogen mixtures simulating the decomposition products of hydrazine. The converging and diverging sides of the inserts had half angles of 30 and 20 degrees, respectively, similar to a flight type unit currently under development. The length of the diverging side was varied to change the area ratio. The nozzle inserts were run over a wide range of specific power. Current, voltage, mass flow rate, and thrust were monitored to provide accurate comparisons between tests. While small differences in performance were observed between the two nozzle inserts, it was determined that for each nozzle insert, arcjet performance improved with increasing nozzle area ratio to the highest area ratio tested and that the losses become very pronounced for area ratios below 50. These trends are somewhat different than those obtained in previous experimental and analytical studies of low Re number nozzles. It appears that arcjet performance can be enhanced via area ratio optimization.

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

Characterisation of subsonic axisymmetric nozzles

Czech Academy of Sciences Publication Activity Database

Tesař, Václav

2008-01-01

Roč. 86, č. 11 (2008), s. 1253-1262 ISSN 0263-8762 R&D Projects: GA AV ČR IAA200760705 Institutional research plan: CEZ:AV0Z20760514 Keywords : nozzle * characterisation * nozzle properties * nozzle invariants Subject RIV: BK - Fluid Dynamics Impact factor: 0.989, year: 2008

Altitude-wind-tunnel investigation of tail-pipe burning with a Westinghouse X24C-4B axial-flow turbojet engine

Science.gov (United States)

Fleming, William A; Wallner, Lewis E

1948-01-01

Thrust augmentation of an axial-flow type turbojet engine by burning fuel in the tail pipe has been investigated in the NACA Cleveland altitude wind tunnel. The performance was determined over a range of simulated flight conditions and tail-pipe fuel flows. The engine tail pipe was modified for the investigation to reduce the gas velocity at the inlet of the tail-pipe combustion chamber and to provide an adequate seat for the flame; four such modifications were investigated. The highest net-thrust increase obtained in the investigation was 86 percent with a net thrust specific fuel consumption of 2.91 and a total fuel-air ratio of 0.0523. The highest combustion efficiencies obtained with the four configurations ranged from 0.71 to 0.96. With three of the tail-pipe burners, for which no external cooling was provided, the exhaust nozzle and the rear part of the burner section were bright red during operation at high tail-pipe fuel-air ratios. With the tail-pipe burner for which fuel and water cooling were provided, the outer shell of the tail-pipe burner showed no evidence of elevated temperatures at any operating condition.

Instrument failure detection of flow measurement in the feedwater system of the Paks Nuclear Power Plant, Hungary

International Nuclear Information System (INIS)

Racz, A.

1990-12-01

The applicability of two different methods for early detection of instrument failures of the flow measurement in feedwater systems are investigated. Both methods are based on Kalman filtering technique of stochastic processes. The reliability of the model for description of a feedwater system is checked by comparing calculated values with measured data. Possible instrument failures are simulated in order to show the capability of the proposed procedures. A practical measurement system arrangement is suggested. (author) 10 refs.; 16 figs.; 4 tabs

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

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.

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.

Laval nozzles for cluster-jet targets

Energy Technology Data Exchange (ETDEWEB)

Grieser, Silke; Bonaventura, Daniel; Hergemoeller, Ann-Katrin; Hetz, Benjamin; Koehler, Esperanza; Lessmann, Lukas; Khoukaz, Alfons [Institut fuer Kernphysik, Westfaelische Wilhelms-Universitaet Muenster, 48149 Muenster (Germany)

2016-07-01

Cluster-jet targets are highly suited for storage ring experiments due to the fact that they provide high and constant beam densities. Therefore, a cluster-jet target is planned to be the first internal target for the PANDA experiment at FAIR. A cluster source generates a continuous flow of cryogenic solid clusters by the expansion of pre-cooled gases within fine Laval nozzles. For the production of clusters the geometry of the nozzle is crucial. The production of such nozzles with their complex inner geometry represents a major technical challenge. The possibility to produce new fine Laval nozzles ensures the operation of cluster-jet targets, e.g. for the PANDA experiment, and opens the way for future investigations on the cluster production process to match the required targets performance. Optimizations on the recently developed production process and the fabrication of new glass nozzles were done. Initial measurements of these nozzles at the PANDA cluster-jet target prototype and the investigation of the cluster beam origin within the nozzle will be presented and discussed. For the future more Laval nozzles with different geometries will be produced and additional measurements with these new nozzles at the PANDA cluster-jet target prototype towards higher performance will be realized.

Results and analysis of a loss-of-feedwater induced ATWS experiment in the LOFT Facility

International Nuclear Information System (INIS)

Grush, W.H.; Koizumi, Y.; Woerth, S.C.

1983-01-01

An anticipated transient without scram (ATWS), initiated by a loss of feedwater, was experimentally simulated in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR). Primary system pressure was controlled using a two-position actuator relief valve to simulate a scaled power-operated relief valve (PORV) and safety relief valve (SRV) representative of those in a commercial PWR. Auxiliary feedwater injection was delayed during the experiment until the plant recovery phase where long-term shutdown was achieved by an operator-controlled plant recovery procedure without inserting the control rods. The system transient response predicted by the RELAP5/MOD1 computer code showed good agreement with the experimental data

Optimization algorithms intended for self-tuning feedwater heater model

International Nuclear Information System (INIS)

Czop, P; Barszcz, T; Bednarz, J

2013-01-01

This work presents a self-tuning feedwater heater model. This work continues the work on first-principle gray-box methodology applied to diagnostics and condition assessment of power plant components. The objective of this work is to review and benchmark the optimization algorithms regarding the time required to achieve the best model fit to operational power plant data. The paper recommends the most effective algorithm to be used in the model adjustment process.

Study on applicability of evaluation model of manpower needs for dismantling of equipments in FUGEN-1. Dismantling process in 3rd/4th feedwater heater room

International Nuclear Information System (INIS)

Shibahara, Yuji; Izumi, Masanori; Nanko, Takashi; Tachibana, Mitsuo; Ishigami, Tsutomu

2010-10-01

Manpower needs for the dismantling process on the dismantling of equipments in FUGEN 3rd/4th feedwater heater room was calculated with the management data evaluation system (PRODIA Code), and it was inspected whether the conventional evaluation model had applicability for FUGEN or not. It was confirmed that the conventional evaluation model for feedwater heater had no applicability. In comparison of the calculated value with the actual data, we found two difference: 1) the calculated value were significantly larger than the actual data, 2) the actual data for the dismantling of 3rd feedwater heater was twice larger than that of 4th feedwater heater, though these equipments were almost same weight. It was found that these were brought 1) by the difference in the work descriptions of dismantling between JPDR and FUGEN, and 2) by that in the cutting number between 3rd feedwater heater and 4th one. The manpower needs for the dismantling of both feedwater heaters were calculated with a new calculation equation reflecting the descriptions of dismantling, and it was found that these results showed the good agreement with the actual data. (author)

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

Automatic regulation of the feedwater turbo-pump capacity for the single-turbine 1000 MW NPP unit

International Nuclear Information System (INIS)

Pavlysh, O.N.; Garbuzov, I.P.; Reukov, Yu.N.

1985-01-01

A schematic of the flow regulators (FR) of feedwater turbo-pumps (FTP) for the single-turbine 1000 MW NPP unit is presented. The FR operate in response to feedwoter signals from FTP or in response to FTP rotor rotational speed and control automatic speed governars. The FR automatic regulation ensures limitation of FTP rotor maximum rotational speed at a feedwater flow rate excess equal to 3600 T/h. The transients in the automatic regulation system are considered. Production tests of FTP FR confirmed the FR operation reliability and the right choice of the regulator concept and structure

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

Development of an advanced PFM code for the integrity evaluation of nuclear piping system under combined aging mechanisms

International Nuclear Information System (INIS)

Datta, Debashis

2010-02-01

A nuclear piping system is composed of several straight pipes and elbows joined by welding. These weld sections are usually the most susceptible failure parts susceptible to various degradation mechanisms. Whereas a specific location of a reactor piping system might fail by a combination of different aging mechanisms, e.g. fatigue and/or stress corrosion cracking, the majority of the piping probabilistic fracture mechanics (PFM) codes can only consider a single aging mechanism at a time. So, a probabilistic fracture mechanics computer code capable of considering multiple aging mechanisms was developed for an accurate failure analysis of each specific component of a nuclear piping section. The newly proposed crack morphology based probabilistic leak flow rate module is introduced in this code to separately treat fatigue and SCC type cracks. Improved models e.g. stressors models, elbow failure model, SIFs model, local seismic occurrence probability model, performance based crack detection models, etc., are also included in this code. Recent probabilistic fatigue (S-N) and SCC crack initiation (S-T) and subsequent crack growth rate models are coded. An integrated probabilistic risk assessment and probabilistic fracture mechanics methodology is proposed. A complete flow chart regarding the combined aging mechanism model is presented. The combined aging mechanism based module can significantly reduce simulation efforts and time. Two NUREG benchmark problems, e.g. reactor pressure vessel outlet nozzle section and a surge line elbow located just below the pressurizer are reinvestigated by this code. The results showed that, contribution of pre-existing cracks in addition to initiating cracks, can significantly increase the overall failure probability. Inconel weld location of reactor pressure vessel outlet nozzle section showed the weakest point in terms of relative through-wall leak failure probability in the order of about 10 -2 at the 40-year plant life. Considering

Flow-throttling orifice nozzle

International Nuclear Information System (INIS)

Sletten, H.L.

1975-01-01

A series-parallel-flow type throttling apparatus to restrict coolant flow to certain fuel assemblies of a nuclear reactor is comprised of an axial extension nozzle of the fuel assembly. The nozzle has a series of concentric tubes with parallel-flow orifice holes in each tube. Flow passes from a high pressure plenum chamber outside the nozzle through the holes in each tube in series to the inside of the innermost tube where the coolant, having dissipated most of its pressure, flows axially to the fuel element. (U.S.)

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.).

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)

1996-12-31

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.).

Method and device for feeding purified water to a pressure vessel

International Nuclear Information System (INIS)

Hirato, Miharu.

1982-01-01

Purpose: To prevent thermal wear at the junction of feedwater pipes and purified water pipes, as well as maintain the function of the purified water feeding system by stopping the introduction of purified water to the heated water feeding system and introducing purified water to the recycling water system upon transient operation or start-up. Constitution: Since a feedwater heater does not function well during transient operation or upon start-up, the temperature of heated water flowing through the feedwater pipe is reduced to produce a temperature difference relative to the set temperature for the purified water feeding system. The temperature difference is detected by a temperature sensor and, when it arrives at a predetermined difference, an operation valve is switched to interrupt the feed of the purified water to the heated water feeding system and it is sent to a water recycling system. Then, the purified water is sent from the water recycling system by way of the discharge portion to the inside of a pressure vessel. Thus, since only the heated water flows to the junction between the cleaned water pipes and the heated water pipes, neither shocks are generated nor the performance of the purified water feeding system is impaired. (Moriyama, K.)

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.

Reliability analysis of the auxiliary feedwater system; Analiza zanesljivosti sistema pomozne napajalne vode

Energy Technology Data Exchange (ETDEWEB)

Susnik, J; Dusic, M [Institut Jozef Stefan, Ljubljana (Yugoslavia)

1984-07-01

The reliability of a NPP auxiliary feedwater system is evaluated using the fault tree analysis. The system is analyzed during the time interval 0 to 6 hours with the computer package program PREP/KITT which is described in more detail. (author)

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

Multielement suppressor nozzles for thrust augmentation systems.

Science.gov (United States)

Lawrence, R. L.; O'Keefe, J. V.; Tate, R. B.

1972-01-01

The noise reduction and nozzle performance characteristics of large-scale, high-aspect-ratio multielement nozzle arrays operated at low velocities were determined by test. The nozzles are selected for application to high-aspect-ratio augmentor suppressors to be used for augmentor wing airplanes. Significant improvements in noise characteristics for multielement nozzles over those of round or high-aspect-ratio slot nozzles are obtained. Elliptical noise patterns typical of slot nozzles are presented for high-aspect-ratio multielement nozzle arrays. Additional advantages are available in OASPL noise reduction from the element size and spacing. Augmentor-suppressor systems can be designed for maximum beam pattern directivity and frequency spectrum shaping advantages. Measurements of the nozzle wakes show a correlation with noise level data and frequency spectrum peaks. The noise and jet wake results are compared with existing prediction procedures based on empirical jet flow equations, Lighthill relationships, Strouhal number, and empirical shock-induced screech noise effects.

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

Energy Technology Data Exchange (ETDEWEB)

Gammage, D.; Idvorian, N. [Babcock & Wilcox Canada Ltd., Cambridge, Ontario (Canada)

2012-07-01

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

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

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.

Mathematical Investigation of the Cavitation Phenomenon in the Nozzle with Partially Surface Wetting

Directory of Open Access Journals (Sweden)

Jablonská Jana

2015-12-01

Full Text Available Partially surface wetting has a great influence on friction losses in the fluid flow in both the pipeline system and the complex shape of hydraulic elements. In many hydraulic elements (valves, pump impellers, cavitation is generated, which significantly changes the hydraulic flow parameters, so the last part of the article is devoted to the mathematical solution of this phenomena and evaluates the impact of wall wetting on the size and shape of the cavitation area which appears in the nozzle and in small gaps at special conditions. If the cavitation appears e. g. near the wall of pipes, the blades of turbine or a pump, then it destroys the material surface. On the basis of this physical experiment (nozzle, a two-dimensional (2D mathematical cavitation model of Schnerr-Sauer was made and calculated shape and size of the cavitation region was compared with the experiment. Later this verified model of cavitation was used for cavitation research flow with partial surface wetting. The pressure drop and the size of the cavitation area as it flows from partially surface wetting theory was tested depending on the adhesion coefficient.

Supersonic flaw detection device for nozzle

International Nuclear Information System (INIS)

Hata, Moriki.

1996-01-01

In a supersonic flaw detection device to be attached to a body surface of a reactor pressure vessel for automatically detecting flaws of a welded portion of a horizontally connected nozzle by using supersonic waves, a running vehicle automatically running along a circumferential direction of the nozzle comprises a supersonic flaw detection means for detecting flaws of the welded portion of the nozzle by using supersonic waves, and an inclination angle sensor for detecting the inclination angle of the running vehicle relative to the central axis of the nozzle. The running distance of the vehicle running along the circumference of the nozzle, namely, the position of the running vehicle from a reference point of the nozzle can be detected accurately by dividing the distance around the nozzle by the inclination angle detected by the inclination angle sensor. Accordingly, disadvantages in the prior art, for example, that the detected values obtained by using an encoder are changed by slipping or idle running of the magnet wheels are eliminated, and accurate flaw detection can be conducted. In addition, an operation of visually adjusting the reference point for the device can be eliminated. An operator's exposure dose can be reduced. (N.H.)

Gas cooled reactors

International Nuclear Information System (INIS)

Kojima, Masayuki.

1985-01-01

Purpose: To enable direct cooling of reactor cores thereby improving the cooling efficiency upon accidents. Constitution: A plurality sets of heat exchange pipe groups are disposed around the reactor core, which are connected by way of communication pipes with a feedwater recycling device comprising gas/liquid separation device, recycling pump, feedwater pump and emergency water tank. Upon occurrence of loss of primary coolants accidents, the heat exchange pipe groups directly absorb the heat from the reactor core through radiation and convection. Although the water in the heat exchange pipe groups are boiled to evaporate if the forcive circulation is interrupted by the loss of electric power source, water in the emergency tank is supplied due to the head to the heat exchange pipe groups to continue the cooling. Furthermore, since the heat exchange pipe groups surround the entire circumference of the reactor core, cooling is carried out uniformly without resulting deformation or stresses due to the thermal imbalance. (Sekiya, K.)

Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

International Nuclear Information System (INIS)

Ahn, Mu-Young; Cho, Seungyon; Lee, Eo Hwak; Park, Yi-Hyun; Lee, Youngmin

2016-01-01

Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

Pipe stress analysis on HCCR-TBS ancillary systems in conceptual design

Energy Technology Data Exchange (ETDEWEB)

Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Eo Hwak [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Park, Yi-Hyun; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Highlights: • Pipe stress is performed on Korean HCCR-TBS for the load combinations including seismic events. • The resultant stress meets the requirement of the design code & standard except one position where modification is needed. • The results gives useful information for the design evolution in the next desgin phase. - Abstract: Korean Helium Cooled Ceramic Reflector (HCCR) Test Blanket System (TBS) will be tested in ITER to demonstrate feasibility of the breeding blanket concept. The HCCR-TBS comprises Test Blanket Module (TBM) with associated shield, and ancillary systems located in various positions of ITER building. Currently, conceptual design for the HCCR-TBS is in progress. This paper presents pipe stress analysis results for the HCCR-TBS ancillary systems. The pipe stress analysis was performed in accordance with ASME B31.3 for major pipes of the Helium Cooling System (HCS) and the Coolant Purification System (CPS), which are operated in high pressure and temperature. The pipe stress for various load cases and load combinations were calculated. Operational pressure and temperature during plasma operation are applied as pressure load and thermal load, respectively. In addition seismic events were combined to investigate the code compliance for sustained load case and occasional load case. It was confirmed that the resultant stress meets the requirements of ASME B31.3 except one position in which it needs modification. These results give useful information for the next design phase, for example, nozzle loads for the component selection, the support design parameters, etc.

Application of EASY5 and MMS modules to BWR controller design

International Nuclear Information System (INIS)

Carmichael, L.A.; Rayes, L.; Yasutake, T.

1987-01-01

The application of EPRI's MMS Library and BCS' EASY5 simulation language to the design of a digital feedwater control system for the Monticello Boiling Water Nuclear Power Plant is discussed. In order to first design and then verify the digital feedwater controller algorithms, a digital simulation model of the Monticello plant was constructed using a combination of custom designed modules, existing MMS two-phase library modules, and standard modules available in the EASY5 library. Details of the process models, namely the BWR nuclear steam supply system, the steamline piping, and the feedwater piping are described in a companion paper. Details of the models for the existing BWR turbine pressure inlet pressure control and recirculation flow control system are described. These models are required to be operational during the transient analysis portion of the feedwater controller design verification, since they interact strongly with the reactor steam flow and water level. The design of the digital feedwater flow control loop is described. Its design is of particular interest because it requires consideration of control loop interaction and is, therefore, a simple example of multivariable non-interacting control design

Safety design of Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR)

International Nuclear Information System (INIS)

Takahashi, Minoru; Uchida, Shoji; Yamada, Yumi; Koyama, Kazuya

2008-01-01

In Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb-Bi coolant above the core, and Pb-Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb-Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits. (author)

Cross-talk effect in electrostatic based capillary array nozzles

International Nuclear Information System (INIS)

Choi, Kyung Hyun; Rahman, Khalid; Khan, Arshad; Kim, Dong Soo

2011-01-01

Electrohydrodynamic printing is a promising technique for printed electronics application. Most researchers working in this field are using a single nozzle configuration. However, for large area printing a multi-nozzle setup will be required for time and cost effective process. In this paper the influence of electric field and flow-rate on jetting angle on multi-nozzle array has been investigated experimentally. A three nozzle setup has been used in a linear array by using glass capillary as a nozzle with independent voltage applied on each nozzle and independent ink supply. The experiments are performed by changing the nozzle to nozzle gap and the effect on the jetting angle has been investigated. It has been observed that by increasing the applied voltage the jetting angle also increases at fixed flow-rate. In case of increasing the flow-rate, the jetting angle first increases with increase in flow-rate, but as the flow-rate increases at certain level the jetting angle decreases; moreover, at a high flow-rate the cone-jet length starts increasing. Numerical simulation has been performed to have a better understanding of the electric-field with respect to jetting angles. The influence of one nozzle on another nozzle is also investigated by operating the nozzle independently by using different operating cases. The cross-talk effect is also minimized by reducing the nozzle diameter. At 250 μm nozzle diameter the cross-talk effect was negligible for 5 mm nozzle-to-nozzle gap. This study will help in better understanding of the interaction between different nozzles in multi-nozzle cases and better design of the multi-nozzle system by minimizing the effects of adjacent nozzles for multi-nozzle electrohydrodynamic printing system

Modelling of hydrothermal characteristics of centrifugal nozzles

International Nuclear Information System (INIS)

Yarkho, A.A.; Omelchenko, M.P.; Borshchev, V.A.

1990-01-01

Presented for the first time is a method of recalculating the hydrothermal characteristics of centrifugal nozzles obtained in laboratory conditions for full-scale nozzles. From the experimental hydrothermal characteristics of nozzles observed in the laboratory it is allowed to calculate the hydrothermal characteristics of any other centrifugal nozzle whose diameter and dimensionless geometric characteristic are known

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.

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...

Nuclear thermal rocket nozzle testing and evaluation program

International Nuclear Information System (INIS)

Davidian, K.O.; Kacynski, K.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. In this report, the Nuclear Thermal Rocket nozzle testing and evaluation program being conducted at the NASA Lewis Research Center 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 plus or minus 1.17%

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)

Modal analysis of main steam line piping under high energy line break condition

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae-Jin; Kim, Seung Hyun; Je, Sang-Yun; Chang, Yoon-Suk [Kyung Hee University, Yongin (Korea, Republic of)

2016-10-15

If HELB (High Energy Line Break) occurs in NPPs (Nuclear Power Plants), not only environmental effect like release of radioactive material but also secondary structural defects should be considered. Jet impingement phenomenon caused by sudden pipe rupture may lead to severe damage on neighboring safe-related components and other structure. Lots of studies have been conducted to assess dynamic behaviors of the SG and MSL piping while pipe whip restraints and jet impingement shields are taken into account during design stage. Arroyo et al. performed modal analyses of a simple square component to examine the jet impingement phenomenon. Also, structural characteristics were predicted to assure structural integrity against the HELB. In this study, we examined dynamic characteristics of SG and MSL piping in a typical 1000MWe NPP. Simulation was performed by using two commercial computational softwares. In particular, modal analyses were conducted to determine mode shapes and natural frequencies of the structure and maximum displacements. The data obtain from each software were compared and observation was discussed in relation to the jet impingement phenomenon. In this research, modal analyses on the SG and MSL piping were carried out to get natural frequencies, vibration mode shapes and maximum displacements. Thereby, the following key finding was observed. (1) Maximum displacement was calculated at the top of SG outlet nozzle with y-directional bending at the third mode. (2) The differences between two models were respectively 7% in natural frequencies and less than 1% in maximum displacements.

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

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

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.

Containment behavior in MSLB with FIV malfunction

Energy Technology Data Exchange (ETDEWEB)

Choi, Hoon; Song, Dong Soo; Jun, Hwang Yong [Korea Hydro and Nuclear Power Co. Ltd, Daejeon (Korea, Republic of)

2012-10-15

In case of Main Steam Line Break(MSLB) accident, sustained high feedwater flow would cause additional cooldown of primary system. Therefore, in addition to the normal control action that closes the main feedwater valves, a safety injection signal rapidly closes all Feed water Control Valve(FCV)s and Feedwater Isolation Valve(FIV)s, trips the main feedwater pumps, and closes the feedwater pump discharge valves. With a single failure of FCVs, FIVs should act as back up protection measures. However, in a certain plant, the FIVs are not automated. If the FIVs could not be credited, the trip of main feedwater pumps can be act as back up protection measures for the single failure of FVCs. In that case, un isolated feedwater which is contained in the pipe between the main feedwater pump and the upstream of the FCV might be flash and be supplied to the broken steam generator. The containment integrity was studied for this case.

The impact of feedwater and condensate return excursions on boiler system component failures

Energy Technology Data Exchange (ETDEWEB)

Esmacher, Mel J. [GE Water and Process Technologies, The Woodlands, TX (United States); Rossi, Anthony [GE Water and Process Technologies, Trevose, PA (United States)

2010-02-15

During boiler operation, the transport of contaminants in boiler feedwater or condensate return via hardness excursions or transport of metal oxides due to corrosion can cause fouling and subsequent tube failure due to under-deposit corrosion or overheating. Case histories are reviewed and suitable corrective actions discussed. (orig.)

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

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

A study on the shell wall thinning causes identified through experiment, numerical analysis and ultrasonic test of high-pressure feedwater heater

International Nuclear Information System (INIS)

Hwang, Kyeong Mo; Woo, Lee; Jin, Tae Eun; Kim, Kyung Hoon

2008-01-01

Feedwater heaters of many nuclear power plants have recently experienced severe wall thinning damage, which accelerates as the operation progresses. Several nuclear power plants in Korea have undergone this damage around the impingement baffle - installed downstream of the high-pressure turbine extraction steam line - inside numbers 5A and 5B feedwater heaters. At that point, the extracted steam from the high-pressure turbine consists in the form of two-phase fluid at high temperature, high pressure and high velocity. Since it flows in reverse direction after impinging the impingement baffle, the shell wall of number 5 high-pressure feedwater heater may be affected by flow-accelerated corrosion. This paper describes the comparisons between the numerical analysis results using the FLUENT code and the downscaled experimental data in an effort to determine root causes of the shell wall thinning of the high-pressure feedwater heaters. The numerical analysis and experimental data were also confirmed by the actual wall thickness measured by ultrasonic tests. From the comparison of the results for the local velocity profiles and the wall thinning measurements, the local velocity component only in the y-direction flowing vertically to the shell wall, and not in the x- and z-directions, was analogous to the wall thinning data

Single nozzle spray drift measurements of drift reducing nozzles at two forward speeds

NARCIS (Netherlands)

Stallinga, H.; Zande, van de J.C.; Michielsen, J.G.P.; Velde, van P.

2016-01-01

In 2011‒2012 single nozzle field experiments were carried out to determine the effect of different flat fan spray nozzles of the spray drift reduction classes 50, 75, 90 and 95% on spray drift at two different forward speeds (7.2 km h-1 and 14.4 km h-1). Experiments were performed with a single

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.

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.

A study on the temperature distribution in the hot leg pipe

International Nuclear Information System (INIS)

Choe, Yoon-Jae; Baik, Se-Jin; Jang, Ho-Cheol; Lee, Byung-Jin; Im, In-Young; Ro, Tae-Sun

2003-01-01

In the hot leg pipes of reactor coolant system of the Korean Standard Nuclear Power Plant (KSNP), a non-uniform distribution in temperature has been observed across the cross-section, which is attributed to the non-uniformity of power distribution in the reactor core usually having a peak in the center region, and to the colder coolant bypass flow through the reactor vessel outlet nozzle clearances. As a result, the arithmetic mean temperature of four Resistance Temperature Detectors (RTDs) installed in each hot leg - two in the upper region and two in the lower region around the pipe wall may not correctly represent the actual coolant bulk temperature. It is also believed that there is a skewness in the velocity profile in the hot leg pipe due to the sudden changes in the flow direction and area from the core to the hot leg pipe, through the reactor vessel outlet plenum. These temperature non-uniformity and velocity skewness affect the measurement of the plant parameter such as the reactor coolant flow rate which is calculated by using the bulk temperature of hot leg pipes. A computational analysis has been performed to simulate the temperature and velocity distributions and to evaluate the uncertainty of temperature correction offset in the hot leg pipe. A commercial CFD code, FLUENT, is used for this analysis. The analysis results are compared with the operational data of KSNP and the scaled-down model test data for System 80. From the comparisons, an uncertainty of correction offset is obtained to measure the bulk temperature of hot leg more accurately, which can be also applied to the operating plants, leading to the reduction of temperature measurement uncertainty. Since the uncertainty of temperature in the hot leg pipe is one of major parameters to calculate the uncertainty of the reactor coolant flow rate, the analysis results can contribute to the improvement of the plant performance and safety by reducing the uncertainty of temperature measurement

Stress analyses of flat plates with attached nozzles. Vol. 2: Experimental stress analyses of a flat plate with one nozzle attached

International Nuclear Information System (INIS)

Battiste, R.L.; Peters, W.H.; Ranson, W.F.; Swinson, W.F.

1975-07-01

Vol. 1 of this report compares experimental results with theoretical stress distributions for a flat plate with one nozzle configuration and for a flat plate with two closely spaced nozzles attached. This volume contains the complete test results for a flat plate with one nozzle attached that was subjected to 1:1 and 1:2 biaxial planar loadings on the plate, to a thrust loading on the nozzle, and to a moment loading on the nozzle. The plate tested was 36 x 36 x 0.375 in., and the attached nozzle had an outer dia of 2.625 in. and a 0.250-in.-thick wall. The nozzle was located in the center of the plate and was considered to be free of weld distortions and irregularities in the junction area. (U.S.)

Design and analysis approach for linear aerospike nozzle

International Nuclear Information System (INIS)

Khan, S.U.; Khan, A.A.; Munir, A.

2014-01-01

The paper presents an aerodynamic design of a simplified linear aerospike nozzle and its detailed exhaust flow analysis with no spike truncation. Analytical method with isentropic planar flow was used to generate the nozzle contour through MATLAB . The developed code produces a number of outputs comprising nozzle wall profile, flow properties along the nozzle wall, thrust coefficient, thrust, as well as amount of nozzle truncation. Results acquired from design code and numerical analyses are compared for observing differences. The numerical analysis adopted an inviscid model carried out through commercially available and reliable computational fluid dynamics (CFD) software. Use of the developed code would assist the readers to perform quick analysis of different aerodynamic design parameters for the aerospike nozzle that has tremendous scope of application in future launch vehicles. Keyword: Rocket propulsion, Aerospike Nozzle, Control Design, Computational Fluid Dynamics. (author)

Effect of nozzle geometry for swirl type twin-fluid water mist nozzle on the spray characteristic

Energy Technology Data Exchange (ETDEWEB)

Yoon, Soon Hyun; Kim, Do Yeon; Kim, Dong Keon [Pusan National University, Busan (Korea, Republic of); Kim, Bong Hwan [Jinju National University, Jinju (Korea, Republic of)

2011-07-15

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.

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

Influence of the loop design of the feedwater- and steam quality in a power plant with pressurized water reactor

International Nuclear Information System (INIS)

Bennert, J.; Becher, L.

1977-01-01

At nuclear power plants with pressurized water reactors, condensate occurs on the high pressure part of the water-steam circuit, caused by the operation with low steam parameters. The behaviour of the electrolytes which entered into the circuit (solubility, distribution in water and/or steam) shows that these electrolytes (salts) are to be found mainly in the condensate. The insinuated electrolytes are reconcentrated during the common arrangements with 'Small Circuit' - consisting of steam generator, high pressure turbine, water separator, feedwater vessel, and have a negative influence on the feedwater - boiler water - and the steam quality. Remedy is possible by modified arrangements, during which these electrolyte-containing condensates will be treated and traced back into the main circuit. Nevertheless that the efficiency decrease is insignificant and additional efforts are necessary, a change over to these arrangements is recommendable, due to the fact that the feedwater quality, the boiler water quality, the steam quality in front of the turbine, and finally also the operational safety, as well as the availability will be improved. (orig.) [de

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)

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.

Pengaruh Jarak dan Posisi Nozzle terhadap Daya Turbin Pelton

OpenAIRE

Kurniawan, Yani; Pane, Erlanda Augupta; Ismail, Ismail

2017-01-01

Pelton Turbine is a turbine which use nozzle as officers the direction of a stream water in order to move around of blade turbine. The rotating of turbine blade efected by some parameters such as the distance of the nozzle, position of nozzle, diameter of nozzle, number of nozzle, and the geometry shape of the blade turbine. An experimental study to analyze the affect of distance and position nozzle to Pelton Turbine of performance. The research method used experiment parameter was position o...

Pengaruh Jarak dan Posisi Nozzle Terhadap Daya Turbin Pelton

OpenAIRE

Yani Kurniawan; Erlanda Augupta Pane; Ismail

2017-01-01

Pelton Turbine is a turbine which use nozzle as officers the direction of a stream water in order to move around of blade turbine. The rotating of turbine blade efected by some parameters such as the distance of the nozzle, position of nozzle, diameter of nozzle, number of nozzle, and the geometry shape of the blade turbine. An experimental study to analyze the affect of distance and position nozzle to Pelton Turbine of performance. The research method used experiment parameter was position o...

An evaluation of TRAC-PF1/MOD1 computer code performance during posttest simulations of Semiscale MOD-2C feedwater line break transients

International Nuclear Information System (INIS)

Hall, D.G.; Watkins, J.C.

1987-01-01

This report documents an evaluation of the TRAC-PF1/MOD1 reactor safety analysis computer code during computer simulations of feedwater line break transients. The experimental data base for the evaluation included the results of three bottom feedwater line break tests performed in the Semiscale Mod-2C test facility. The tests modeled 14.3% (S-FS-7), 50% (S-FS-11), and 100% (S-FS-6B) breaks. The test facility and the TRAC-PF1/MOD1 model used in the calculations are described. Evaluations of the accuracy of the calculations are presented in the form of comparisons of measured and calculated histories of selected parameters associated with the primary and secondary systems. In addition to evaluating the accuracy of the code calculations, the computational performance of the code during the simulations was assessed. A conclusion was reached that the code is capable of making feedwater line break transient calculations efficiently, but there is room for significant improvements in the simulations that were performed. Recommendations are made for follow-on investigations to determine how to improve future feedwater line break calculations and for code improvements to make the code easier to use

Ferromagnetic material inspection for feedwater heater and condenser tubes

International Nuclear Information System (INIS)

Anon.

1987-01-01

In recent years, special ferritic stainless steels, such as AL29-4C/sup TM/, Sea-Cure/sup TM/, E-Brite/sup TM/, 439, and similar alloys have been introduced as tube material in condensers, feedwater heaters, moisture separator/reheaters, and other heat exchangers. In addition, carbon steel tubes are widely used in feedwater heaters and heat exchangers in chemical plants. The main problem with the in-service inspection of these ferritic alloys and carbon steel tubes lies in their highly ferromagnetic properties. These properties severely limit the application of the standard eddy current techniques. The effort was undertaken under EPRI sponsorship to develop a reliable technique for in-service inspection of ferromagnetic tubes. The new method combines the measurement of magnetic flux leakage generated around the defects with measurement of total flux in the tube wall. The heart of the inspection system is a special ID probe that magnetizes the tube and generates signals for any tube defect. A permanent record of inspection is provided with a strip-chart or magnetic tape recorder. The laboratory and field evaluation of this new system demonstrated its very good sensitivity to small defects, its reliability, and its ruggedness. Defects as small as 10% external wall loss in heavy wall carbon steel tube were detected. Tubes in the power plant were inspected at a rate of 300-500 tubes per eight-hour shift. The other advantages of this newly developed technique are its simplicity, low cost of instrumentation, easy data interpretation, and full portability

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

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)

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.

Some current engineering topics in nuclear power plant components

International Nuclear Information System (INIS)

Amana, M.

1977-01-01

An analysis based on the principle of fracture mechanics, is presented for several engineering problems occuring in nuclear power plant components. The specific problems covered are: underclad cracking; stress corrosion cracking; cracks in HAZ of nozzle weld; feedwater nozzle corner crack; shift of transition temperature due to neutron irradiation; LWR local-ECC thermal shock experiment; and design and material selection of RPV in terms of fracture mechanics. (B.R.H.)

Pengaruh Jarak dan Posisi Nozzle Terhadap Daya Turbin Pelton

Directory of Open Access Journals (Sweden)

Yani Kurniawan

2017-12-01

Full Text Available Pelton Turbine is a turbine which use nozzle as officers the direction of a stream water in order to move around of blade turbine. The rotating of turbine blade efected by some parameters such as the distance of the nozzle, position of nozzle, diameter of nozzle, number of nozzle, and the geometry shape of the blade turbine. An experimental study to analyze the affect of distance and position nozzle to Pelton Turbine of performance. The research method used experiment parameter was position of nozzle with three variations, first position is the right side horizontal of bottom shaft turbine, second position is vertical to down direction, and third position is the left side horizontal of upper shaft turbine. The parameter of nozzle distance used five variations was 24 cm, 23 cm, 22 cm, 21 cm, dan 20 cm, which measured from the end of position nozzle to blade turbine. The result shows that the right side horizontal of bottom shaft turbine with distance of nozzle 23 cm had the maximum performance to produce a power 125 Watt with the rotation of shaft turbine 263 rpm.

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.

Device for the analysis of feedwater and condensation samples from power plants

International Nuclear Information System (INIS)

Mostofin, A.A.; Sorokina, N.S.

1978-01-01

An improved version of a device for automatic measurement of the salt and NH 3 contents of feedwater and condensate samples from nuclear power plants is described. Only one sample is required for determining both values. The invention proposes on the one hand to change the dimensions of a throttle opening and on the other to install a second measuring instrument (conductivity measuring instrument). (UWI) [de

'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)

Lower nozzle of PWR fuel assembly

International Nuclear Information System (INIS)

Furutani, Nobuo.

1994-01-01

A lower nozzle comprises a regular square plate and legs. The plate has a plurality of holes for securing thimble tubes and a great number of water flowing ports. Ridges each having a lower end surface inclined toward inner side of the plate are disposed at the outer circumference of the plate. The legs suspend downwardly from the corners of the plate and support the plate at a predetermined gap between a lower reactor core plate and the plate. The inclined surfaces of the ridges disposed at the outer circumference of the plate retain coolants, that were caused to flow to the outside passing between the legs of the nozzle, while dividing them to the inside of the nozzle and circulate the coolants upwardly passing through the water flowing ports of the plate. Further, since obstacles abut against the inclined surfaces of the ridges and flow to the inner side of the lower nozzle, obstacles in the coolants can be captured substantially entirely by the lower nozzle. (I.N.)

Airfoil nozzle and shroud assembly

Science.gov (United States)

Shaffer, J.E.; Norton, P.F.

1997-06-03

An airfoil and nozzle assembly are disclosed including an outer shroud having a plurality of vane members attached to an inner surface and having a cantilevered end. The assembly further includes a inner shroud being formed by a plurality of segments. Each of the segments having a first end and a second end and having a recess positioned in each of the ends. The cantilevered end of the vane member being positioned in the recess. The airfoil and nozzle assembly being made from a material having a lower rate of thermal expansion than that of the components to which the airfoil and nozzle assembly is attached. 5 figs.

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

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)

FEM Analysis and Measurement of Residual Stress by Neutron Diffraction on the Dissimilar Overlay Weld Pipe

International Nuclear Information System (INIS)

Kim, Kang Soo; Lee, Ho Jin; Woo, Wan Chuck; Seong, Baek Seok; Byeon, Jin Gwi; Park, Kwang Soo; Jung, In Chul

2010-01-01

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

Operation method for after-heat removal

International Nuclear Information System (INIS)

Tamano, Toyomi; Sakuragi, Masanori; Kogiso, Zen-ichi; Wake, Minoru.

1987-01-01

Purpose: To moderate thermal shocks applied to a feedwater pipe plate portion at the inlet of a steam generator thereby maintaining the integrity and safety of an LMFBR type plant. Method: Water with feed from the condenser to a steam generator. Steams generated in the steam generator are introduced to an air/water separator in a recycling system and a control device is actuated. Water separated by the air/water separator is recycled to the steam generator, while monitoring the temperature variation coefficient for the feedwater temperature at the inlet of the steam generator. If the temperature variation coefficient exceeds a predetermined setting value, the recycling flow rate is decreased in accordance with the deviation. This can greatly moderate the thermal shocks applied to the feedwater pipe plate portion at the inlet of the steam generator upon starting of the recycling system. (Takahashi, M.)

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.

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

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.

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

KAUST Repository

Dehwah, Abdullah; Li, Sheng; Almashharawi, Samir; Winters, Harvey; Missimer, Thomas M.

2015-01-01

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.

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.

Cleanings of the silica scale settled in the transportation-pipes of the geothermal hot water of the Onuma Geothermal Power Station

Energy Technology Data Exchange (ETDEWEB)

Ito, J

1978-09-01

At the Onuma Geothermal Power Station, silica scale deposits in the hot water transportation pipes between production wells and injection wells, increased the thickness. The operations for cleaning the scale were effectively carried out by the following three methods. (1) Poli-Pig method: The shell-shaped plastic foam sponge mass named Poli-Pig was pressed in the pipes. Various shaped Poli-Pig such as armed by the steel spikes made scratches on the surface of the scale, and then stripped off. This method is effective when thickness of the scale is thinner than 20 mm. (2) Impact-Cutter method. Various shaped steel cutter blocks were attached at the end of a flexible shaft, and gave continuous impact by rotation on the scale and then smashing it away. This method is effective for various thickness, but pipes had to be cut off matched to the length of the flexible shaft. (3) Water-jet method. High pressured water jet through the special nozzle smashed away the scale. For this method the pipe had to be cut off at every joint.

Flame Interactions and Thermoacoustics in Multiple-Nozzle Combustors

Science.gov (United States)

Dolan, Brian

The first major chapter of original research (Chapter 3) examines thermoacoustic oscillations in a low-emission staged multiple-nozzle lean direct injection (MLDI) combustor. This experimental program investigated a relatively practical combustor sector that was designed and built as part of a commercial development program. The research questions are both practical, such as under what conditions the combustor can be safely operated, and fundamental, including what is most significant to driving the combustion oscillations in this system. A comprehensive survey of operating conditions finds that the low-emission (and low-stability) intermediate and outer stages are necessary to drive significant thermoacoustics. Phase-averaged and time-resolved OH* imaging show that dramatic periodic strengthening and weakening of the reaction zone downstream of the low-emission combustion stages. An acoustic modal analysis shows the pressure wave shapes and identifies the dominant thermoacoustic behavior as the first longitudinal mode for this combustor geometry. Finally, a discussion of the likely significant coupling mechanisms is given. Periodic reaction zone behavior in the low-emission fuel stages is the primary contributor to unsteady heat release. Differences between the fuel stages in the air swirler design, the fuel number of the injectors, the lean blowout point, and the nominal operating conditions all likely contribute to the limit cycle behavior of the low-emission stages. Chapter 4 investigates the effects of interaction between two adjacent swirl-stabilized nozzles using experimental and numerical tools. These studies are more fundamental; while the nozzle hardware is the same as the lean direct injection nozzles used in the MLDI combustion concept, the findings are generally applicable to other swirl-stabilized combustion systems as well. Much of the work utilizes a new experiment where the distance between nozzles was varied to change the level of interaction

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)

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.

In-pipe aerodynamic characteristics of a projectile in comparison with free flight for transonic Mach numbers

Science.gov (United States)

Hruschka, R.; Klatt, D.

2018-03-01

The transient shock dynamics and drag characteristics of a projectile flying through a pipe 3.55 times larger than its diameter at transonic speed are analyzed by means of time-of-flight and pipe wall pressure measurements as well as computational fluid dynamics (CFD). In addition, free-flight drag of the 4.5-mm-pellet-type projectile was also measured in a Mach number range between 0.5 and 1.5, providing a means for comparison against in-pipe data and CFD. The flow is categorized into five typical regimes the in-pipe projectile experiences. When projectile speed and hence compressibility effects are low, the presence of the pipe has little influence on the drag. Between Mach 0.5 and 0.8, there is a strong drag increase due to the presence of the pipe, however, up to a value of about two times the free-flight drag. This is exactly where the nose-to-base pressure ratio of the projectile becomes critical for locally sonic speed, allowing the drag to be estimated by equations describing choked flow through a converging-diverging nozzle. For even higher projectile Mach numbers, the drag coefficient decreases again, to a value slightly below the free-flight drag at Mach 1.5. This behavior is explained by a velocity-independent base pressure coefficient in the pipe, as opposed to base pressure decreasing with velocity in free flight. The drag calculated by CFD simulations agreed largely with the measurements within their experimental uncertainty, with some discrepancies remaining for free-flying projectiles at supersonic speed. Wall pressure measurements as well as measured speeds of both leading and trailing shocks caused by the projectile in the pipe also agreed well with CFD.

Focusing liquid microjets with nozzles

International Nuclear Information System (INIS)

Acero, A J; Ferrera, C; Montanero, J M; Gañán-Calvo, A M

2012-01-01

The stability of flow focusing taking place in a converging–diverging nozzle, as well as the size of the resulting microjets, is examined experimentally in this paper. The results obtained in most aspects of the problem are similar to those of the classical plate-orifice configuration. There is, however, a notable difference between flow focusing in nozzles and in the plate-orifice configuration. In the former case, the liquid meniscus oscillates laterally (global whipping) for a significant area of the control parameter plane, a phenomenon never observed when focusing with the plate-orifice configuration. Global whipping may constitute an important drawback of flow focusing with nozzles because it reduces the robustness of the technique. (paper)

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.

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.

Water level detection pipeline

International Nuclear Information System (INIS)

Koshikawa, Yukinobu; Imanishi, Masatoshi; Niizato, Masaru; Takagi, Masahiro

1998-01-01

In the present invention, water levels of a feedwater heater and a drain tank in a nuclear power plant are detected at high accuracy. Detection pipeline headers connected to the upper and lower portions of a feedwater heater or a drain tank are connected with each other. The connection line is branched at appropriate two positions and an upper detection pipeline and a lower detection pipeline are connected thereto, and a gauge entrance valve is disposed to each of the detection pipelines. A diaphragm of a pressure difference generator is connected to a flange formed to the end portion. When detecting the change of water level in the feedwater heater or the drain tank as a change of pressure difference, gauge entrance valves on the exit side of the upper and lower detection pipelines are connected by a connection pipe. The gauge entrance valve is closed, a tube is connected to the lower detection pipe to inject water to the diaphragm of the pressure difference generator passing through the connection pipe thereby enabling to calibrate the pressure difference generator. The accuracy of the calibration of instruments is improved and workability thereof upon flange maintenance is also improved. (I.S.)

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

Proceedings of the specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification

International Nuclear Information System (INIS)

1998-01-01

This specialists meeting on experience with thermal fatigue in LWR piping caused by mixing and stratification, was held in June 1998 in Paris. It included five sessions. Session 1: operating experience (7 papers): Historical perspective; EDF experience with local thermohydraulic phenomena in PWRs: impacts and strategies; Thermal fatigue in safety injection lines of French PWRs: technical problems, regulatory requirements, concerns about other areas; US NRC Regulatory perspective on unanticipated thermal fatigue in LWR piping; Failure to the Residual Heat Removal system suction line pipe in Genkai unit 1 caused by thermal stratification cycling; Emergency Core Cooling System pipe crack incident at Tihange unit 1; Two leakages induced by thermal stratification at the Loviisa power plant). Session 2: thermal hydraulic phenomena (5 papers): Thermal stratification in small pipes with respect to fatigue effects and so called 'Banana effect'; Thermal stratification in the surge line of the Korean next generation reactor; Thermal stratification in horizontal pipes investigated in UPTF-TRAM and HDR facilities; Research on thermal stratification in un-isolable piping of reactor pressure boundary; Thermal mixing phenomena in piping systems: 3D numerical simulation and design considerations. Session 3: response of material and structure (5 papers): Fatigue induced by thermal stratification, Results of tests and calculations of the COUFAST model; Laboratory simulation of thermal fatigue cracking as a basis for verifying life models; Thermo-mechanical analysis methods for the conception and the follow up of components submitted to thermal stratification transients; Piping analysis methods of a PWR surge line for stratified flow; The thermal stratification effect on surge lines, The VVER estimation. Session 4: monitoring aspects (4 papers): Determination of the thermal loadings affecting the auxiliary lines of the reactor coolant system in French PWR plants; Expected and

Experience feedback of an operation event during the experiment of feed-water pump switch

International Nuclear Information System (INIS)

Sun Shuhai; Li Huasheng; Zhang Hao

2012-01-01

In this paper an event is summarized and analyzed, which caused the quit of the high-pressure heaters and the nuclear power rising, during the experiment of the driven feed-water pump switch. The good experience feedback on this event is brought out through gathering related information of domestic nuclear plants. (authors)

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

Effect of nozzle arrangement on Venturi scrubber performance

Energy Technology Data Exchange (ETDEWEB)

Ananthanarayanan, N.V.; Viswanathan, S.

1999-12-01

The effect of nozzle arrangement on flux distribution is studied in a rectangular, pilot-scale, Pease-Anthony-type Venturi scrubber. The annular, two-phase, heterogeneous, three-dimensional gas-liquid flow inside the scrubber is modeled using a commercial computational fluid dynamic (CFD) package, FLUENT. The comparison of predicted liquid drop concentration shows good agreement with experimental data. The model predicts the fraction of liquid flowing as film on the walls reasonably well. Visualization of flux patterns studied using four typical nozzle configurations indicate that the nonuniformity in flux distribution increases when the nozzle-to-nozzle distance is greater than 10% of the width of the side on which the nozzles are placed. An analysis of the effect of multiple jet penetration lengths on liquid flux distribution yielded a comparable distribution at 10--45% less liquid than uniform penetration for a particular nozzle configuration. This would lead to significant improvements in scrubber performance by achieving comparable collection efficiency at a lower pressure drop.

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,

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

Development of top nozzle for Korean standard LWR fuel

Energy Technology Data Exchange (ETDEWEB)

Lee, S. K.; Kim, I. K.; Choi, K. S.; Kim, Y. H.; Lee, J. N.; Kim, H. K. [KNFC, Taejon (Korea, Republic of)

2001-10-01

Performance evaluation was executed for each component and its assembly for the deduced Top Nozzles to develop the new Top Nozzle for LWR. This new Top Nozzle is composed of the optimum components among the derived Top Nozzles that have been evaluated in the viewpoint of structural integrity, simpleness of dismantle and assembly, manufacturability etc. In this study, the developed Top Nozzle satisfied all the related design criteria. In special, it makes fuel repair time reduced by assembling and disassembling itself as one body, and improves Fuel Assembly holddown ability by revising the design parameters of its spring and the structural integrity through the betterment of its geometrical shpae of Flange and Holddown Plate as compared with the existing LWR Top Nozzles.

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...

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.)

Investigation of a two-phase nozzle flow and validation of several computer codes by the experimental data

International Nuclear Information System (INIS)

Kedziur, F.

1980-03-01

Stationary experiments with a convergent nozzle are performed in order to validate advanced two-phase computer codes, which find application in the blowdown-phase of a loss-of-coolant accident (LOCA). The steam/water flow presents a broad variety of initial conditions: The pressure varies between 2 and 13 MPa, the void fraction between 0 (subcooled) and about 80%, a great number of subcritical as well as critical experiments with different flow pattern is investigated. Additional air/water experiments serve for the separation of phase transition effects. The transient acceleration of the fluid in the LOCA-case is simulated by a local acceleration in the experiments. The layout of the nozzle and the applied measurement technique allow for a separate testing of physical models and the determination of empirical model parameters, respectively: In the four codes DUESE, DRIX-20, RELAP4/MOD6 and STRUYA the models - if they exist - for slip between the phases, thermodynamic non-equilibrium, pipe friction and critical mass flow rate are validated and criticised in comparison with the experimental data, and the corresponding model parameters are determined. The parameters essentially are a function of the void fraction. (orig.) [de

The effects of parameter variation on MSET models of the Crystal River-3 feedwater flow system

International Nuclear Information System (INIS)

Miron, A.

1998-01-01

In this paper we develop further the results reported in Reference 1 to include a systematic study of the effects of varying MSET models and model parameters for the Crystal River-3 (CR) feedwater flow system The study used archived CR process computer files from November 1-December 15, 1993 that were provided by Florida Power Corporation engineers Fairman Bockhorst and Brook Julias. The results support the conclusion that an optimal MSET model, properly trained and deriving its inputs in real-time from no more than 25 of the sensor signals normally provided to a PWR plant process computer, should be able to reliably detect anomalous variations in the feedwater flow venturis of less than 0.1% and in the absence of a venturi sensor signal should be able to generate a virtual signal that will be within 0.1% of the correct value of the missing signal

Dual-nozzle microfluidic droplet generator

Science.gov (United States)

Choi, Ji Wook; Lee, Jong Min; Kim, Tae Hyun; Ha, Jang Ho; Ahrberg, Christian D.; Chung, Bong Geun

2018-05-01

The droplet-generating microfluidics has become an important technique for a variety of applications ranging from single cell analysis to nanoparticle synthesis. Although there are a large number of methods for generating and experimenting with droplets on microfluidic devices, the dispensing of droplets from these microfluidic devices is a challenge due to aggregation and merging of droplets at the interface of microfluidic devices. Here, we present a microfluidic dual-nozzle device for the generation and dispensing of uniform-sized droplets. The first nozzle of the microfluidic device is used for the generation of the droplets, while the second nozzle can accelerate the droplets and increase the spacing between them, allowing for facile dispensing of droplets. Computational fluid dynamic simulations were conducted to optimize the design parameters of the microfluidic device.

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

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.

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

International Nuclear Information System (INIS)

Baum, M.R.

1987-01-01

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

Effect of Injector Nozzle Holes on Diesel Engine Performance

OpenAIRE

Semin,; Yusof, Mohd Yuzri Mohd; Arof, Aminuddin Md; Shaharudin, Daneil Tomo; Ismail, Abdul Rahim

2010-01-01

All of the injector nozzle holes have examined and the results are shown that the seven holes nozzle have provided the best burning result for the fuel in-cylinder burned in any different engine speeds and the best burning is in low speed engine. In engine performance effect, all of the nozzles have examined and the five holes nozzle provided the best result in indicted power, indicated torque and ISFC in any different engine speeds.

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)

Pipe support

International Nuclear Information System (INIS)

Pollono, L.P.

1979-01-01

A pipe support for high temperature, thin-walled piping runs such as those used in nuclear systems is described. A section of the pipe to be suppported is encircled by a tubular inner member comprised of two walls with an annular space therebetween. Compacted load-bearing thermal insulation is encapsulated within the annular space, and the inner member is clamped to the pipe by a constant clamping force split-ring clamp. The clamp may be connected to pipe hangers which provide desired support for the pipe

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

Design and testing of low-divergence elliptical-jet nozzles

Energy Technology Data Exchange (ETDEWEB)

Rouly, Etienne; Warkentin, Andrew; Bauer, Robert [Dalhousie University, Halifax (China)

2015-05-15

A novel approach was developed to design and fabricate nozzles to produce high-pressure low-divergence fluid jets. Rapid-prototype fabrication allowed for myriad experiments investigating effects of different geometric characteristics of nozzle internal geometry on jet divergence angle and fluid distribution. Nozzle apertures were elliptical in shape with aspect ratios between 1.00 and 2.45. The resulting nozzle designs were tested and the lowest elliptical jet divergence angle was 0.4 degrees. Nozzle pressures and flowrates ranged from 0.32 to 4.45 MPa and 13.6 to 37.9 LPM, respectively. CimCool CimTech 310 machining fluid was used in all experiments at a Brix concentration of 6.6 percent.

Computer Graphic Design Using Auto-CAD and Plug Nozzle Research

Science.gov (United States)

Rogers, Rayna C.

2004-01-01

The purpose of creating computer generated images varies widely. They can be use for computational fluid dynamics (CFD), or as a blueprint for designing parts. The schematic that I will be working on the summer will be used to create nozzles that are a part of a larger system. At this phase in the project, the nozzles needed for the systems have been fabricated. One part of my mission is to create both three dimensional and two dimensional models on Auto-CAD 2002 of the nozzles. The research on plug nozzles will allow me to have a better understanding of how they assist in the thrust need for a missile to take off. NASA and the United States military are working together to develop a new design concept. On most missiles a convergent-divergent nozzle is used to create thrust. However, the two are looking into different concepts for the nozzle. The standard convergent-divergent nozzle forces a mixture of combustible fluids and air through a smaller area in comparison to where the combination was mixed. Once it passes through the smaller area known as A8 it comes out the end of the nozzle which is larger the first or area A9. This creates enough thrust for the mechanism whether it is an F-18 fighter jet or a missile. The A9 section of the convergent-divergent nozzle has a mechanism that controls how large A9 can be. This is needed because the pressure of the air coming out nozzle must be equal to that of the ambient pressure other wise there will be a loss of performance in the machine. The plug nozzle however does not need to have an A9 that can vary. When the air flow comes out it can automatically sense what the ambient pressure is and will adjust accordingly. The objective of this design is to create a plug nozzle that is not as complicated mechanically as it counterpart the convergent-divergent nozzle.

ATWS analysis for total loss of feedwater sequence in UCN 3 and 4

International Nuclear Information System (INIS)

Park, S. H.; Song, Y. M.; Kim, D. H.; Kim, S. D.; Park, S. Y.

1999-01-01

ATWS is a trip-failed severe accident initiated from the transients like a turbine trip, a control bank withdrawal, and a loss of feedwater which are expected to occur comparatively often (one or two occurrences / year). In this study, an ATWS sequence in Ulchin 3 and 4 is analyzed and the effects of the important systems are studied for accident management purpose using a MIDAS/PK computer code. The MIDAS/PK code has been developed via coupling a point kinetics module with the MELCOR code. The code calculates a primary peak pressure of about 24MPa at 240 seconds for the ATWS initiated by a TLOF (Total Loss of Feedwater) transient. Along with the basic ATWS analysis, several sensitivity runs are performed. From these, the turbines and the safety depressurization system (SDS) are judged to be important. The turbine trip resulting in a loss of offsite power and a RCP trip, degrades primary heat transfer to the secondary sides, and in turn, increases primary coolant temperature which reduces the reactor power due to the negative moderator temperature coefficient. Manual operation of SDS has an effect to lower the primary peak pressure considerably via supplementary depressurization in addition to the PORVs

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

Analysis of Nozzle Jet Plume Effects on Sonic Boom Signature

Science.gov (United States)

Bui, Trong

2010-01-01

An axisymmetric full Navier-Stokes computational fluid dynamics (CFD) study was conducted to examine nozzle exhaust jet plume effects on the sonic boom signature of a supersonic aircraft. A simplified axisymmetric nozzle geometry, representative of the nozzle on the NASA Dryden NF-15B Lift and Nozzle Change Effects on Tail Shock (LaNCETS) research airplane, was considered. The highly underexpanded nozzle flow is found to provide significantly more reduction in the tail shock strength in the sonic boom N-wave pressure signature than perfectly expanded and overexpanded nozzle flows. A tail shock train in the sonic boom signature, similar to what was observed in the LaNCETS flight data, is observed for the highly underexpanded nozzle flow. The CFD results provide a detailed description of the nozzle flow physics involved in the LaNCETS nozzle at different nozzle expansion conditions and help in interpreting LaNCETS flight data as well as in the eventual CFD analysis of a full LaNCETS aircraft. The current study also provided important information on proper modeling of the LaNCETS aircraft nozzle. The primary objective of the current CFD research effort was to support the LaNCETS flight research data analysis effort by studying the detailed nozzle exhaust jet plume s imperfect expansion effects on the sonic boom signature of a supersonic aircraft. Figure 1 illustrates the primary flow physics present in the interaction between the exhaust jet plume shock and the sonic boom coming off of an axisymmetric body in supersonic flight. The steeper tail shock from highly expanded jet plume reduces the dip of the sonic boom N-wave signature. A structured finite-volume compressible full Navier-Stokes CFD code was used in the current study. This approach is not limited by the simplifying assumptions inherent in previous sonic boom analysis efforts. Also, this study was the first known jet plume sonic boom CFD study in which the full viscous nozzle flow field was modeled, without

Stresses in reactor pressure vessel nozzles -- Calculations and experiments

International Nuclear Information System (INIS)

Brumovsky, M.; Polachova, H.

1995-01-01

Reactor pressure vessel nozzles are characterized by a high stress concentration which is critical in their low-cycle fatigue assessment. Program of experimental verification of stress/strain field distribution during elastic-plastic loading of a reactor pressure vessel WWER-1000 primary nozzle model in scale 1:3 is presented. While primary nozzle has an ID equal to 850 mm, the model nozzle has ID equal to 280 mm, and was made from 15Kh2NMFA type of steel. Calculation using analytical methods was performed. Comparison of results using different analytical methods -- Neuber's, Hardrath-Ohman's as well as equivalent energy ones, used in different reactor Codes -- is shown. Experimental verification was carried out on model nozzles loaded statically as well as by repeated loading, both in elastic-plastic region. Strain fields were measured using high-strain gauges, which were located in different distances from center of nozzle radius, thus different stress concentration values were reached. Comparison of calculated and experimental data are shown and compared

Analysis of KNU1 loss of normal feedwater

International Nuclear Information System (INIS)

Kim, Hho-Jung; Chung, Bub-Dong; Lee, Young-Jin; Kim, Jin-Soo

1986-01-01

Simulation of the system thermal-hydraulic parameters was carried out following the KNU1 (Korea Nuclear Unit-1) loss of normal feedwater transient sequence occurred on November 14, 1984. Results were compared with the plant transient data, and good agreements were obtained. Some deviations were found in the parameters such as the steam flowrate and the RCS (Reactor Coolant system) average temperature, around the time of reactor trip. It can be expected since the thermal-hydraulic parameters encounter rapid transitions due to the large reduction of the reactor thermal power in a short period of time and, thereby, the plant data involve transient uncertainties. The analysis was performed using the RELAP5/MOD1/NSC developed through some modifications of the interphase drag and the wall heat transfer modeling routines of the RELAP5/MOD1/CY018. (author)

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.

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.

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

Modified computation of the nozzle damping coefficient in solid rocket motors

Science.gov (United States)

Liu, Peijin; Wang, Muxin; Yang, Wenjing; Gupta, Vikrant; Guan, Yu; Li, Larry K. B.

2018-02-01

In solid rocket motors, the bulk advection of acoustic energy out of the nozzle constitutes a significant source of damping and can thus influence the thermoacoustic stability of the system. In this paper, we propose and test a modified version of a historically accepted method of calculating the nozzle damping coefficient. Building on previous work, we separate the nozzle from the combustor, but compute the acoustic admittance at the nozzle entry using the linearized Euler equations (LEEs) rather than with short nozzle theory. We compute the combustor's acoustic modes also with the LEEs, taking the nozzle admittance as the boundary condition at the combustor exit while accounting for the mean flow field in the combustor using an analytical solution to Taylor-Culick flow. We then compute the nozzle damping coefficient via a balance of the unsteady energy flux through the nozzle. Compared with established methods, the proposed method offers competitive accuracy at reduced computational costs, helping to improve predictions of thermoacoustic instability in solid rocket motors.

Nature of convection-stabilized dc arcs in dual-flow nozzle geometry

International Nuclear Information System (INIS)

Serbetci, I.; Nagamatsu, H.T.

1990-01-01

In this paper, an experimental investigation of the steady-state low-current air arcs in a dual-flow nozzle system is presented. First, the cold flow with no arc as determined for various nozzle geometries, i.e., two- and three-dimensional and orifice nozzles, and nozzle pressure ratios. Supersonic flow separation and oblique and detached shock waves were observed in the flow field. Using a finite-element computer program, the Mach number contours were determined in the flow field for various nozzle-gap spacings and pressure ratios. In addition, the dc arc voltage and current measurements were made for an electrode gap spacing of ∼ 5.5 cm and current levels of I ∼ 25, 50, and 100 A for the three nozzle geometries. The arc voltage and arc power increased rapidly as the flow speed increased from zero to sonic velocity at the nozzle throat. The shock waves in the converging-diverging nozzles resulted in a decrease in the overall resistance by about 15 percent

Estimation of stress intensity factors for circumferential cracked pipes under welding residual stress filed

International Nuclear Information System (INIS)

Oh, Chang Young; Kim, Yun Jae; Oh, Young Jin; Song, Tae Kwang; Kim, Yong Beum; Oh, Young Jin; Song, Tae Kwang; Kim, Yong Beum

2012-01-01

Recently, stress corrosion cracking(SCC) have been found in dissimilar metal welds of nozzles in some pressurized water reactors and on low carbon stainless steel piping systems of boiling water reactors. The important factor of SCC is the residual stress field caused by weld. For the evaluation of crack growth analysis due to SCC, stress intensity factor under a residual stress field should be estimated. Several solutions for stress intensity factor under residual stress field were recommended in flaw assessment codes such as the American Society of Mechanical Engineers (ASME) Section XI, R6, American Petroleum Institute (API579). Some relevant works have been studied. Dong et al. evaluated stress intensity factors in welded structures. Miyazaki et al. estimated stress intensity factors of surface crack in simple stress fields. This paper presents a simple method to estimate stress intensity factors in welding residual stress field. For general application, results of structure integrity assessment codes KI solutions were compared Finite element analyses of welding simulation and cracked pipes are described. Comparison results of KI solutions and proposed simplified solution are presented in the works

A computational investigation on the influence of the use of elliptical orifices on the inner nozzle flow and cavitation development in diesel injector nozzles

International Nuclear Information System (INIS)

Molina, S.; Salvador, F.J.; Carreres, M.; Jaramillo, D.

2014-01-01

Highlights: • The influence of elliptical orifices on the inner nozzle flow is compared. • Five nozzles with different elliptical and circular orifices are simulated. • Differences in the flow coefficients and cavitation morphology are observed. • Horizontal axis orifices are ease to cavitate, with a higher discharge coefficient. • A better mixing process quality is expected for the horizontal major axis nozzles. - Abstract: In this paper a computational study was carried out in order to investigate the influence of the use of elliptical orifices on the inner nozzle flow and cavitation development. With this aim, a large number of injection conditions have been simulated and analysed for 5 different nozzles: four nozzles with different elliptical orifices and one standard nozzle with circular orifices. The four elliptical nozzles differ from each other in the orientation of the major axis (vertical or horizontal) and in the eccentricity value, but keeping the same outlet section in all cases. The comparison has been made in terms of mass flow, momentum flux and other important non-dimensional parameters which help to describe the behaviour of the inner nozzle flow: discharge coefficient (C d ), area coefficient (C a ) and velocity coefficient (C v ). The simulations have been done with a code able to simulate the flow under either cavitating or non-cavitating conditions. This code has been previously validated using experimental measurements over the standard nozzle with circular orifices. The main results of the investigation have shown how the different geometries modify the critical cavitation conditions as well as the discharge coefficient and the effective velocity. In particular, elliptical geometries with vertically oriented major axis are less prone to cavitate and have a lower discharge coefficient, whereas elliptical geometries with horizontally oriented major axis are more prone to cavitate and show a higher discharge coefficient

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

Erosion wear of boron carbide ceramic nozzles by abrasive air-jets

International Nuclear Information System (INIS)

Deng Jianxin

2005-01-01

Boron carbide nozzles were produced by hot pressing. The erosion wear of this nozzle caused by abrasive particle impact was investigated by abrasive air-jets. Silica, silicon carbide and alumina powders with different hardness were used as the erodent abrasive particles. Results showed that the hardness of the erodent particles played an important role with respect to the erosion wear of the boron carbide nozzles. As the hardness of the erodent particles increases, there is a dramatic increase in erosion rate of the nozzles. The nozzle entrance area suffered from severe abrasive impact under large impact angles, and generated maximum tensile stresses. The wear mechanisms of boron carbide nozzle at this area appeared to be entirely brittle in nature with the evidence of large scale-chipping, and exhibited a brittle fracture induced removal process. While at the nozzle center wall section, most of the particles traveled parallel to the nozzle wall, and showed minimum tensile stresses. The wear mode in this area of the nozzle changed from impact to sliding erosion, and the wear mechanisms appeared to be the lateral cracking owing to a surface fatigue fracture mechanism

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

Annular Internal-External-Expansion Rocket Nozzles for Large Booster Applications

Science.gov (United States)

Connors, James F.; Cubbison, Robert W.; Mitchell, Glenn A.

1961-01-01

For large-thrust booster applications, annular rocket nozzles employing both internal and external expansion are investigated. In these nozzles, free-stream air flows through the center as well as around the outside of the exiting jet. Flaps for deflecting the rocket exhaust are incorporated on the external-expansion surface for thrust-vector control. In order to define nozzle off-design performance, thrust vectoring effectiveness, and external stream effects, an experimental investigation was conducted on two annular nozzles with area ratios of 15 and 25 at Mach 0, 2, and 3 in the Lewis 10- by 10-foot wind tunnel. Air, pressurized to 600 pounds per square inch absolute, was used to simulate the exhaust flow. For a nozzle-pressure-ratio range of 40 to 1000, the ratio of actual to ideal thrust was essentially constant at 0.98 for both nozzles. Compared with conventional convergent-divergent configurations on hypothetical boost missions, the performance gains of the annular nozzle could yield significant orbital payload increases (possibly 8 to 17 percent). A single flap on the external-expansion surface of the area-ratio-25 annular nozzle produced a side force equal to 4 percent of the axial force with no measurable loss in axial thrust.

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

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.

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

The role of nozzle convergence in diesel combustion

Energy Technology Data Exchange (ETDEWEB)

J. Benajes; S. Molina; C. Gonzaalez; R. Donde [CMT-Motores Termicos, Universidad Politecnica de Valencia, Valencia (Spain)

2008-08-15

An experimental study has been performed for identifying the role of injector nozzle hole convergence and cavitation in diesel engine combustion and pollutant emissions. For doing so, five nozzles were tested under different operating and experimental conditions. The critical cavitation number of each nozzle was analyzed. With this value, an estimation of the mixing process at different conditions obtained. This data is used to explain the combustion results which are analyzed in terms of the apparent combustion time, rate of heat release, in-cylinder pressures, adiabatic temperatures and soot and NOx emissions. Special emphasis is put in developing an expression to explicitly link the mixing process and the injection rate with the rate of heat release. The results show that the fuel-air mixing process can be improved by the use of both convergent and cavitating nozzles, thus lowering the soot emissions. The NOx production, being dependent of the injection rate and the mixing process, does not necessarily increase with the use of more convergent nozzles. 40 refs., 8 fig., tabs.

Design and Checkout of a High Speed Research Nozzle Evaluation Rig

Science.gov (United States)

Castner, Raymond S.; Wolter, John D.

1997-01-01

The High Flow Jet Exit Rig (HFJER) was designed to provide simulated mixed flow turbojet engine exhaust for one- seventh scale models of advanced High Speed Research test nozzles. The new rig was designed to be used at NASA Lewis Research Center in the Nozzle Acoustic Test Rig and the 8x6 Supersonic Wind Tunnel. Capabilities were also designed to collect nozzle thrust measurement, aerodynamic measurements, and acoustic measurements when installed at the Nozzle Acoustic Test Rig. Simulated engine exhaust can be supplied from a high pressure air source at 33 pounds of air per second at 530 degrees Rankine and nozzle pressure ratios of 4.0. In addition, a combustion unit was designed from a J-58 aircraft engine burner to provide 20 pounds of air per second at 2000 degrees Rankine, also at nozzle pressure ratios of 4.0. These airflow capacities were designed to test High Speed Research nozzles with exhaust areas from eighteen square inches to twenty-two square inches. Nozzle inlet flow measurement is available through pressure and temperature sensors installed in the rig. Research instrumentation on High Speed Research nozzles is available with a maximum of 200 individual pressure and 100 individual temperature measurements. Checkout testing was performed in May 1997 with a 22 square inch ASME long radius flow nozzle. Checkout test results will be summarized and compared to the stated design goals.

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

Evaluation of heatup and recovery in a loss of feedwater accident with multiple failure

International Nuclear Information System (INIS)

Bang, Young Seok; Seul, Kwang Won; Kim, Hho Jung

1991-01-01

A loss of feedwater accident with multiple failure has been studied in order to identify the potential severity of the accident when compared with the design basis accident in PWR. The PCS heatup and recovery mode in a LOFA with multiple failure was evaluated using the LOFT L9-1/L3-3 experiment. From experimental result, 4 separable subphase were identified and the associated phenomena were also addressed

Study of a flapper-nozzle system for a water hydraulic servovalve; Suiatsu survo ben ni mochiiru nozzle flapper kei no kenkyu

Energy Technology Data Exchange (ETDEWEB)

Urata, E.; Nakao, Y. [Kanagawa University, Kanagawa (Japan). Faculty of Engineering

1997-06-25

The paper discusses the characteristics of a flapper-nozzle system for water hydraulic servovalves. High pressure water at the supply port is first used as the working fluid for the hydrostatic bearings supporting the spool. Spool valve stiction induced by poor lubrication with water is thus avoided. The fluid is then led to the ends of the spool and is used as the working fllid of the flapper-nozzle system. In the new flapper-nozzle system the circumferential clearance of the spool becomes a laminar restriction that substitutes fixed orifice used in conventional servovalves. The linearity in the pressure-displacement relationship of the new flapper-nozzle system is better than that of conventional fixed orifice systems. 7 refs., 11 figs., 2 tabs.

System and method having multi-tube fuel nozzle with differential flow

Science.gov (United States)

Hughes, Michael John; Johnson, Thomas Edward; Berry, Jonathan Dwight; York, William David

2017-01-03

A system includes a multi-tube fuel nozzle with a fuel nozzle body and a plurality of tubes. The fuel nozzle body includes a nozzle wall surrounding a chamber. The plurality of tubes extend through the chamber, wherein each tube of the plurality of tubes includes an air intake portion, a fuel intake portion, and an air-fuel mixture outlet portion. The multi-tube fuel nozzle also includes a differential configuration of the air intake portions among the plurality of tubes.

A CFD-based aerodynamic design procedure for hypersonic wind-tunnel nozzles

Science.gov (United States)

Korte, John J.

1993-01-01

A new procedure which unifies the best of current classical design practices, computational fluid dynamics (CFD), and optimization procedures is demonstrated for designing the aerodynamic lines of hypersonic wind-tunnel nozzles. The new procedure can be used to design hypersonic wind tunnel nozzles with thick boundary layers where the classical design procedure has been shown to break down. An efficient CFD code, which solves the parabolized Navier-Stokes (PNS) equations using an explicit upwind algorithm, is coupled to a least-squares (LS) optimization procedure. A LS problem is formulated to minimize the difference between the computed flow field and the objective function, consisting of the centerline Mach number distribution and the exit Mach number and flow angle profiles. The aerodynamic lines of the nozzle are defined using a cubic spline, the slopes of which are optimized with the design procedure. The advantages of the new procedure are that it allows full use of powerful CFD codes in the design process, solves an optimization problem to determine the new contour, can be used to design new nozzles or improve sections of existing nozzles, and automatically compensates the nozzle contour for viscous effects as part of the unified design procedure. The new procedure is demonstrated by designing two Mach 15, a Mach 12, and a Mach 18 helium nozzles. The flexibility of the procedure is demonstrated by designing the two Mach 15 nozzles using different constraints, the first nozzle for a fixed length and exit diameter and the second nozzle for a fixed length and throat diameter. The computed flow field for the Mach 15 least squares parabolized Navier-Stokes (LS/PNS) designed nozzle is compared with the classically designed nozzle and demonstrates a significant improvement in the flow expansion process and uniform core region.

Mounting apparatus for a nozzle guide vane assembly

Science.gov (United States)

Boyd, Gary L.; Shaffer, James E.

1995-01-01

The present invention provides a ceramic nozzle guide assembly with an apparatus for mounting it to a metal nozzle case that includes an intermediate ceramic mounting ring. The mounting ring includes a plurality of projections that are received within a plurality of receptacles formed in the nozzle case. The projections of the mounting ring are secured within the receptacles by a ceramic retainer that allows contact between the two components only along arcuate surfaces thus eliminating sliding contact between the components.

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.

Nozzle Mounting Method Optimization Based on Robot Kinematic Analysis

Science.gov (United States)

Chen, Chaoyue; Liao, Hanlin; Montavon, Ghislain; Deng, Sihao

2016-08-01

Nowadays, the application of industrial robots in thermal spray is gaining more and more importance. A desired coating quality depends on factors such as a balanced robot performance, a uniform scanning trajectory and stable parameters (e.g. nozzle speed, scanning step, spray angle, standoff distance). These factors also affect the mass and heat transfer as well as the coating formation. Thus, the kinematic optimization of all these aspects plays a key role in order to obtain an optimal coating quality. In this study, the robot performance was optimized from the aspect of nozzle mounting on the robot. An optimized nozzle mounting for a type F4 nozzle was designed, based on the conventional mounting method from the point of view of robot kinematics validated on a virtual robot. Robot kinematic parameters were obtained from the simulation by offline programming software and analyzed by statistical methods. The energy consumptions of different nozzle mounting methods were also compared. The results showed that it was possible to reasonably assign the amount of robot motion to each axis during the process, so achieving a constant nozzle speed. Thus, it is possible optimize robot performance and to economize robot energy.

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.

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)

Heat structure coupling of CUPID and MARS for the multi-scale simulation of the passive auxiliary feedwater system

International Nuclear Information System (INIS)

Kyu Cho, Hyoung; Cho, Yun Je; Yoon, Han Young

2014-01-01

Graphical abstract: - Highlights: • PAFS is designed to replace a conventional active auxiliary feedwater system. • Multi-D T/H analysis code, CUPID was coupled with the 1-D system analysis code MARS. • The coupled CUPID and MARS was applied for the multi-scale analysis of the PAFS test facility. • The simulation result showed that the coupled code can reproduce important phenomena in PAFS. - Abstract: For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been developed. In the present study, the CUPID code was coupled with a system analysis code MARS in order to apply it for the multi-scale thermal-hydraulic analysis of the passive auxiliary feedwater system (PAFS). The PAFS is one of the advanced safety features adopted in the Advanced Power Reactor Plus (APR+), which is intended to completely replace the conventional active auxiliary feedwater system. For verification of the coupling and validation of the coupled code, the PASCAL test facility was simulated, which was constructed with an aim of validating the cooling and operational performance of the PAFS. The two-phase flow phenomena of the steam supply system including the condensation inside the heat exchanger tube were calculated by MARS while the natural circulation and the boil-off in the large water pool that contains the heat exchanger tube were simulated by CUPID. This paper presents the description of the PASCAL facility, the coupling method and the simulation results using the coupled code

The modelling of an SF6 arc in a supersonic nozzle: II. Current zero behaviour of the nozzle arc

International Nuclear Information System (INIS)

Zhang, Q; Liu, J; Yan, J D; Fang, M T C

2016-01-01

The present work (part II) forms the second part of an investigation into the behaviour of SF 6 nozzle arc. It is concerned with the aerodynamic and electrical behaviour of a transient nozzle arc under a current ramp specified by a rate of current decay (d i /d t ) before current zero and a voltage ramp (d V /d t ) after current zero. The five flow models used in part I [1] for cold gas flow and DC nozzle arcs have been applied to study the transient arc at three stagnation pressures ( P 0 ) and two values of d i /d t for the current ramp, representing a wide range of arcing conditions. An analysis of the physical mechanisms encompassed in each flow model is given with an emphasis on the adequacy of a particular model in describing the rapidly varying arc around current zero. The critical rate of rise of recovery voltage (RRRV) is found computationally and compared with test results of Benenson et al [2]. For transient nozzle arcs, the RRRV is proportional to the square of P 0 , rather than to the square root of P 0 for DC nozzle arcs. The physical mechanisms responsible for the strong dependence of RRRV on P 0 have been investigated. The relative merits of the flow models employed are discussed. (paper)

Design methods for structures under thermal ratchet

International Nuclear Information System (INIS)

Branca, T.R.; McLean, J.L.

1975-01-01

Previous work on the thermal ratchet analysis of a simple pipe is extended to the case of an intersection of a pipe with a spherical shell. The chosen nozzle configuration is subjected to an internal pressure which remains constant, and a cyclic thermal transient which is representative of the type of transient that might be expected for components of a LMFBR. A number of cross-sections through the nozzle were examined, each yielding a different combination of elastic primary and secondary stress. These stresses, together with their associated cyclic strain growth, as determined from an elastic-plastic-creep analysis of the nozzle, were then plotted on a Miller or Bree-type diagram. Thus, a number of points, one for each cross-section considered, were available for comparison with the data obtained from the ratchet analysis of simple pipe sections. Both the elastic and inelastic analyses on the nozzle were performed using the finite element method of structural analysis of the ANSYS computer code. The pipe ratchetting cases were computed using the Oak Ridge National Laboratory PLACRE code. For a simple pipe ratchet case, a brief comparison is given between the version of ANSYS used in this study, the ANSYS version used in previous work and PLACRE code. The three programs did not yield identical results. Further study is needed to resolve the discrepancies that were observed. The results of the comparison between the nozzle ratchet and pipe ratchet solutions indicate that reasonable predictions can be made for the nozzle ratchet strains based on elastic parameters and design curves developed from pipe ratchetting solutions. (author)

Structure Optimization and Numerical Simulation of Nozzle for High Pressure Water Jetting

Directory of Open Access Journals (Sweden)

Shuce Zhang

2015-01-01

Full Text Available Three kinds of nozzles normally used in industrial production are numerically simulated, and the structure of nozzle with the best jetting performance out of the three nozzles is optimized. The R90 nozzle displays the most optimal jetting properties, including the smooth transition of the nozzle’s inner surface. Simulation results of all sample nozzles in this study show that the helix nozzle ultimately displays the best jetting performance. Jetting velocity magnitude along Y and Z coordinates is not symmetrical for the helix nozzle. Compared to simply changing the jetting angle, revolving the jet issued from the helix nozzle creates a grinding wheel on the cleaning surface, which makes not only an impact effect but also a shearing action on the cleaning object. This particular shearing action improves the cleaning process overall and forms a wider, effective cleaning range, thus obtaining a broader jet width.

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.

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.

HPFRCC - Extruded Pipes

DEFF Research Database (Denmark)

Stang, Henrik; Pedersen, Carsten

1996-01-01

The present paper gives an overview of the research onHigh Performance Fiber Reinforced Cementitious Composite -- HPFRCC --pipes recently carried out at Department of Structural Engineering, Technical University of Denmark. The project combines material development, processing technique development......-w$ relationship is presented. Structural development involved definition of a new type of semi-flexiblecement based pipe, i.e. a cement based pipe characterized by the fact that the soil-pipe interaction related to pipe deformation is an importantcontribution to the in-situ load carrying capacity of the pipe...

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.

TMI-2 instrument nozzle examinations at Argonne National Laboratory

International Nuclear Information System (INIS)

Neimark, L.A.; Shearer, T.L.; Purohit, A.; Hins, A.G.

1993-09-01

Six of the 14 instrument-penetration-tube nozzles removed from the lower head of TMI-2 were examined to identify damage mechanisms, provide insight to the fuel relocation scenario, and provide input data to the margin-to-failure analysis. Visual inspection, gamma scanning, metallography, microhardness measurements, and scanning electron microscopy were used to obtain the desired information. The results showed varying degrees of damage to the lower head nozzles, from ∼50% melt-off to no damage at all to near-neighbor nozzles. The elevations of nozzle damage suggested that the lower elevations (near the lower head) were protected from molten fuel, apparently by an insulating layer of fuel debris. The pattern of nozzle damage was consistent with fuel movement toward the hot-spot location identified in the vessel wall. Evidence was found for the existence of a significant quantity of control assembly debris on the lower head before the massive relocation of fuel occurred

A Parametric Investigation of Nozzle Planform and Internal/External Geometry at Transonic Speeds

Science.gov (United States)

Cler, Daniel L.

1995-01-01

An experimental investigation of multidisciplinary (scarfed trailing edge) nozzle divergent flap geometry was conducted at transonic speeds in the NASA Langley 16-Foot Transonic Tunnel. The geometric parameters investigated include nozzle planform, nozzle contouring location (internal and/or external), and nozzle area ratio (area ratio 1.2 and 2.0). Data were acquired over a range of Mach Numbers from 0.6 to 1.2, angle-of-attack from 0.0 degrees to 9.6 degrees and nozzle pressure ratios from 1.0 to 20.0. Results showed that increasing the rate of change internal divergence angle across the width of the nozzle or increasing internal contouring will decrease static, aeropropulsive and thrust removed drag performance regardless of the speed regime. Also, increasing the rate of change in boattail angle across the width of the nozzle or increasing external contouring will provide the lowest thrust removed drag. Scarfing of the nozzle trailing edges reduces the aeropropulsive performance for the most part and adversely affects the nozzle plume shape at higher nozzle pressure ratios thus increasing the thrust removed drag. The effects of contouring were primary in nature and the effects of planform were secondary in nature. Larger losses occur supersonically than subsonically when scarfing of nozzle trailing edges occurs. The single sawtooth nozzle almost always provided lower thrust removed drag than the double sawtooth nozzles regardless the speed regime. If internal contouring is required, the double sawtooth nozzle planform provides better static and aeropropulsive performance than the single sawtooth nozzle and if no internal contouring is required the single sawtooth provides the highest static and aeropropulsive performance.

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.

Pipe damping

International Nuclear Information System (INIS)

Ware, A.G.

1985-01-01

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

Support structure for reactor core constituent element

International Nuclear Information System (INIS)

Aida, Yasuhiko.

1993-01-01

A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

Pipe inspection using the pipe crawler. Innovative technology summary report

International Nuclear Information System (INIS)

1999-05-01

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

Pipe inspection using the pipe crawler. Innovative technology summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1999-05-01

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

Design and Analyses of High Aspect Ratio Nozzles for Distributed Propulsion Acoustic Measurements

Science.gov (United States)

Dippold, Vance F., III

2016-01-01

A series of three convergent round-to-rectangular high-aspect ratio nozzles were designed for acoustics measurements. The nozzles have exit area aspect ratios of 8:1, 12:1, and 16:1. With septa inserts, these nozzles will mimic an array of distributed propulsion system nozzles, as found on hybrid wing-body aircraft concepts. Analyses were performed for the three nozzle designs and showed that the flow through the nozzles was free of separated flow and shocks. The exit flow was mostly uniform with the exception of a pair of vortices at each span-wise end of the nozzle.

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)

3D modeling and visualization software for complex geometries

International Nuclear Information System (INIS)

Guse, Guenter; Klotzbuecher, Michael; Mohr, Friedrich

2011-01-01

The reactor safety depends on reliable nondestructive testing of reactor components. For 100% detection probability of flaws and the determination of their size using ultrasonic methods the ultrasonic waves have to hit the flaws within a specific incidence and squint angle. For complex test geometries like testing of nozzle welds from the outside of the component these angular ranges can only be determined using elaborate mathematical calculations. The authors developed a 3D modeling and visualization software tool that allows to integrate and present ultrasonic measuring data into the 3D geometry. The software package was verified using 1:1 test samples (example: testing of the nozzle edge of the feedwater nozzle of a steam generator from the outside; testing of the reactor pressure vessel nozzle edge from the inside).

Development of built-in debris-filter bottom nozzle for PWR fuel assemblies

International Nuclear Information System (INIS)

Juntaro Shimizu; Kazuki Monaka; Masaji Mori; Kazuo Ikeda

2005-01-01

Mitsubishi Heavy Industries, Ltd. (MHI) has worked to improve the capability of anti debris bottom nozzle for a PWR fuel assembly. The Current debris filter bottom nozzle (DFBN) having 4mm diameter flow holes can capture the larger size of debris than the flow hole inner diameter. MHI has completed the development of the built-in debris filter bottom nozzle, which is the new idea of the debris-filter for high burnup (55GWd/t assembly average burnup). Built-in debris filter bottom nozzle consists of the blades and nozzle body. The blades made from inconel strip are embedded and welded on the grooved top surface of the bottom nozzle adapter plate. A flow hole is divided by the blade and the trap size of the debris is reduced. Because the blades block the coolant flow, it was anticipated to increase the pressure loss of the nozzle, however, adjusting the relation between blade and taper shape of the flow hole, the pressure loss has been successfully maintained the satisfactory level. Grooves are cut on the nozzle plate; nevertheless, the additional skirts on the four sides of the nozzle compensate the structural strength. (authors)

Elliptic nozzle aspect ratio effect on controlled jet propagation

Energy Technology Data Exchange (ETDEWEB)

Kumar, S M Aravindh; Rathakrishnan, Ethirajan, E-mail: aravinds@iitk.ac.in, E-mail: erath@iitk.ac.in [Department of Aerospace Engineering, Indian Institute of Technology, Kanpur (India)

2017-04-15

The present study deals with the control of a Mach 2 elliptic jet from a convergent–divergent elliptic nozzle of aspect ratio 4 using tabs at the nozzle exit. The experiments were carried out for rectangular and triangular tabs of the same blockage, placed along the major and minor axes of the nozzle exit, at different levels of nozzle expansion. The triangular tabs along the minor axis promoted superior mixing compared to the other controlled jets and caused substantial core length reduction at all the nozzle pressure ratios studied. The rectangular tabs along the minor axis caused core length reduction at all pressure ratios, but the values were minimal compared to that of triangular tabs along the minor axis. For all the test conditions, the mixing promotion caused by tabs along the major axis was inferior to that of tabs along the minor axis. The waves present in the core of controlled jets were visualized using a shadowgraph. Comparison of the present results with the results of a controlled Mach 2 elliptic jet of aspect ratio 2 (Aravindh Kumar and Sathakrishnan 2016 J. Propulsion Power 32 121–33, Aravindh Kumar and Rathakrishnan 2016 J. Aerospace Eng. at press (doi:10.1177/0954410016652921)) show that for all levels of expansion, the mixing effectiveness of triangular tabs along the minor axis of an aspect ratio 4 nozzle is better than rectangular or triangular tabs along the minor axis of an aspect ratio 2 nozzle. (paper)

Elliptic nozzle aspect ratio effect on controlled jet propagation

International Nuclear Information System (INIS)

Kumar, S M Aravindh; Rathakrishnan, Ethirajan

2017-01-01

The present study deals with the control of a Mach 2 elliptic jet from a convergent–divergent elliptic nozzle of aspect ratio 4 using tabs at the nozzle exit. The experiments were carried out for rectangular and triangular tabs of the same blockage, placed along the major and minor axes of the nozzle exit, at different levels of nozzle expansion. The triangular tabs along the minor axis promoted superior mixing compared to the other controlled jets and caused substantial core length reduction at all the nozzle pressure ratios studied. The rectangular tabs along the minor axis caused core length reduction at all pressure ratios, but the values were minimal compared to that of triangular tabs along the minor axis. For all the test conditions, the mixing promotion caused by tabs along the major axis was inferior to that of tabs along the minor axis. The waves present in the core of controlled jets were visualized using a shadowgraph. Comparison of the present results with the results of a controlled Mach 2 elliptic jet of aspect ratio 2 (Aravindh Kumar and Sathakrishnan 2016 J. Propulsion Power 32 121–33, Aravindh Kumar and Rathakrishnan 2016 J. Aerospace Eng. at press (doi:10.1177/0954410016652921)) show that for all levels of expansion, the mixing effectiveness of triangular tabs along the minor axis of an aspect ratio 4 nozzle is better than rectangular or triangular tabs along the minor axis of an aspect ratio 2 nozzle. (paper)

Use of direct washing of chemical dispense nozzle for defect control

Science.gov (United States)

Linnane, Michael; Mack, George; Longstaff, Christopher; Winter, Thomas

2006-03-01

Demands for continued defect reduction in 300mm IC manufacturing are driving process engineers to examine all aspects of the chemical apply process for improvement. Historically, the defect contribution from photoresist apply nozzles has been minimized through a carefully controlled process of "dummy dispenses" to keep the photoresist in the tip "fresh" and remove any solidified material, a preventive maintenance regime involving periodic cleaning or replacing of the nozzles, and reliance on a pool of solvent within the nozzle storage block to keep the photoresist from solidifying at the nozzle tip. The industry standard has worked well for the most part but has limitations in terms of cost effectiveness and absolute defect elimination. In this study, we investigate the direct washing of the chemical apply nozzle to reduce defects seen on the coated wafer. Data is presented on how the direct washing of the chemical dispense nozzle can be used to reduce coating related defects, reduce material costs from the reduction of "dummy dispense", and can reduce equipment downtime related to nozzle cleaning or replacement.

Pipe connector

International Nuclear Information System (INIS)

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

1978-01-01

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

Turbine combustor with fuel nozzles having inner and outer fuel circuits

Science.gov (United States)

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

2013-12-24

A combustor cap assembly for a turbine engine includes a combustor cap and a plurality of fuel nozzles mounted on the combustor cap. One or more of the fuel nozzles would include two separate fuel circuits which are individually controllable. The combustor cap assembly would be controlled so that individual fuel circuits of the fuel nozzles are operated or deliberately shut off to provide for physical separation between the flow of fuel delivered by adjacent fuel nozzles and/or so that adjacent fuel nozzles operate at different pressure differentials. Operating a combustor cap assembly in this fashion helps to reduce or eliminate the generation of undesirable and potentially harmful noise.

Evaluation of the effects of break nozzle configuration in the Semiscale Mod-1 system

International Nuclear Information System (INIS)

Hanson, R.G.

1977-08-01

The Semiscale Mod-1 Program has utilized two different break nozzle configurations in the test system. An evaluation has been made to determine the effect these break nozzle configurations have on system thermal-hydraulic response during a 200 percent double-ended cold leg break loss-of-coolant accident simulation. The first nozzle was a convergent-divergent nozzle (Henry nozzle) and the second, an elongated constant area throat nozzle. Analysis is confined primarily to system response phenomena observed to be affected by the nozzle configuration and concentrates on the fluid response at the break and the resulting core behavior during subcooled and saturated blowdown. The evaluation shows that considerable difference in system response occurs as a result of the difference in break nozzle configuration. The elongated throat nozzle was scaled from the Loss-of-Fluid Test (LOFT) nozzle geometry and since the LOFT counterpart tests were designed to provide results for the LOFT Program, the elongated throat nozzle was used in the subsequent LOFT counterpart tests

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

Shock unsteadiness in a thrust optimized parabolic nozzle

Science.gov (United States)

Verma, S. B.

2009-07-01

This paper discusses the nature of shock unsteadiness, in an overexpanded thrust optimized parabolic nozzle, prevalent in various flow separation modes experienced during start up {(δ P0 /δ t > 0)} and shut down {(δ P0/δ t The results are based on simultaneously acquired data from real-time wall pressure measurements using Kulite pressure transducers, high-speed schlieren (2 kHz) of the exhaust flow-field and from strain-gauges installed on the nozzle bending tube. Shock unsteadiness in the separation region is seen to increase significantly just before the onset of each flow transition, even during steady nozzle operation. The intensity of this measure ( rms level) is seen to be strongly influenced by relative locations of normal and overexpansion shock, the decrease in radial size of re-circulation zone in the back-flow region, and finally, the local nozzle wall contour. During restricted shock separation, the pressure fluctuations in separation region exhibit periodic characteristics rather than the usually observed characteristics of intermittent separation. The possible physical mechanisms responsible for the generation of flow unsteadiness in various separation modes are discussed. The results are from an experimental study conducted in P6.2 cold-gas subscale test facility using a thrust optimized parabolic nozzle of area-ratio 30.

Design methods of Coanda effect nozzle with two streams

Directory of Open Access Journals (Sweden)

Michele TRANCOSSI

2014-03-01

Full Text Available This paper continues recent research of the authors about the ACHEON Coanda effect two streams nozzle. This nozzle aims to produce an effective deflection of a propulsive jet with a correspondent deviation of the thrust vector in a 2D plane. On the basis of a previously published mathematical model, based on integral equations, it tries to produce an effective design guideline, which can be adopted for design activities of the nozzle for aeronautic propulsion. The presented model allows defining a governing method for this innovative two stream synthetic jet nozzle. The uncertainness level of the model are discussed and novel aircraft architectures based on it are presented. A CFD validation campaign is produced focusing on validating the model and the designs produced.

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.

Extending cavitation models to subcooled and superheated nozzle flow

International Nuclear Information System (INIS)

Schmidt, D.P.; Corradini, M.L.

1997-01-01

Existing models for cavitating flow are extended to apply to discharge of hot liquid through nozzles. Two types of models are considered: an analytical model and a two-dimensional numerical model. The analytical model of cavitating nozzle flow is reviewed and shown to apply to critical nozzle flow where the liquid is subcooled with respect to the downstream conditions. In this model the liquid and vapor are assumed to be in thermodynamic equilibrium. The success of this analytical model suggests that hydrodynamic effects dominate the subcooled nozzle flow. For more detailed predictions an existing multi-dimensional cavitation model based on hydrodynamic non-equilibrium is modified to apply to discharge of hot liquid. Non-equilibrium rate data from experimental measurements are used to close the equations. The governing equations are solved numerically in time and in two spatial dimensions on a boundary fitted grid. Results are shown for flow through sharp nozzles, and the coefficient of discharge is found to agree with experimental measurements for both subcooled and flashing fluid. (author)

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.

Design and performance of atomizing nozzles for spray calcination of high-level wastes

International Nuclear Information System (INIS)

Miller, F.A.; Stout, L.A.

1981-05-01

A key aspect of high-level liquid-waste spray calcination is waste-feed atomization by using air atomizing nozzles. Atomization substantially increases the heat transfer area of the waste solution, which enhances rapid drying. Experience from the spray-calciner operations has demonstrated that nozzle flow conditions that produce 70-μ median-volume-diameter or smaller spray droplets are required for small-scale spray calciners (drying capacity less than 80 L/h). For large-scale calciners (drying capacity greater than 300 L/h), nozzle flow conditions that produce 100-μ median-volume-diameter or smaller spray droplets are required. Mass flow ratios of 0.2 to 0.4, depending on nozzle size, are required for proper operation of internal-mix atomizing nozzles. Both internal-mix and external-mix nozzles have been tested at PNL. Due to the lower airflow requirements and fewer large droplets produced, the internal-mix nozzle has been chosen for primary development in the spray calciner program at PNL. Several nozzle air-cap materials for internal-mix nozzles have been tested for wear resistance. Results show that nozzle air caps of stainless steel and Cer-vit (a machineable glass ceramic) are suceptible to rapid wear by abrasive slurries, whereas air caps of alumina and reaction-bonded silicon nitride show only slow wear. Longer-term testing is necessary to determine more accurately the actual frequency of nozzle replacement. Atomizing nozzle air caps of alumina are subject to fracture from thermal shock, whereas air caps of silicon nitride and Cer-vit are not. Fractured nozzles are held in place by the air-cap retaining ring and continue to atomize satisfactorily. Therefore, fractures caused by thermal shocking do not necessarily result in nozzle failure

Fluid flow nozzle energy harvesters

Science.gov (United States)

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

2015-04-01

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 [1] 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.

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.

Operational experience on reduction of feedwater iron and liquid radwaste input for Kuosheng Nuclear Power Plant

International Nuclear Information System (INIS)

Wen, T.J.; Huang, Theresa Chen; Liu, Wen Tsung; Liu, T.C.; Shyur, Tzu Sheng; Shen, S.C.

1998-01-01

Other than cobalt alloys, or low cobalt materials, feedwater iron content plays an important role in crud activation and transport causing the growth of out-of-core radiation fields and associated with radwaste generation. Before installing prefilter in the upstream of condensate deep-bed demineralizer, increasing demand for suspended solid removal required new backwash and regeneration technique in Kuosheng Nuclear Power Plant. At steady state full power operation, the average iron concentration in condensate demineralizer influent was 8-15 ppb. Considering both the necessity of backwash and reduction of liquid radwaste input, several actions had been taken to promote the crud removal capabilities without using ultrasonic resin cleaner and controlled feedwater iron content between 0.5 and 2.0 ppb. This modified resin backwash technique would also generate minimum liquid radwaste. Meanwhile, significant efforts have been made to promote the quality of waste water by carefully control input streams as well as backwash modification to reduce liquid radwaste generation. The daily quantity of liquid radwaste has decreased dramatically in the past two years and is effectively controlled under the expected average daily input of design basis. (author)

Numerical simulation of residual stress in piping components at Framatome-ANP

International Nuclear Information System (INIS)

Gilies, P.; Franco, C.; Cipiere, M.-F.; Ould, P.

2005-01-01

Numerous manufacturing processes induce residual stresses and distortions in piping components and associated welds: quenching of cast pipings, machining and welding. In Pressurized Water Reactors, most of the components have a large thickness for sustaining pressure and distortions are a minor source of concern. This is not the case for residual stresses which may have a strong influence on several type of damage such as fatigue, corrosion, brittle fracture. In low toughness components, residual stress fields may contribute to ductile tearing initiation. These potential damages are mitigated after welding by stress relief heat treatment, which is applied in a systematic manner to ferritic components of the primary system in nuclear reactors. This treatment is not applied on austenitic piping for which the heat treatment temperature is limited due to the risk of sensitization and residual stresses are difficult to eliminate completely. Since on site measurements are costly and difficult to perform, numerical simulation appears to be an attractive tool for estimating residual stress distributions. Framatome-ANP is working on modelling manufacturing processes with that purpose in mind. This paper presents three kinds of applications illustrating efforts on welding, quenching and machining simulation. First a comparison is shown between computations and measurements of residual stress induced by welding of a dissimilar weld metal junction. Then numerical simulations of quenching of a cast stainless steel nozzle are presented. Finally quenching followed by machining and grinding of this cast component are considered in a full simulation of the manufacturing process. Computed distortions and residual stresses are compared with experimental measurements at different stages of the manufacturing process. (authors)

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

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

MARVIKEN-CFT, Marviken Full Scale Critical Flow Tests

International Nuclear Information System (INIS)

1998-01-01

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

Transient Side Load Analysis of Out-of-Round Film-Cooled Nozzle Extensions

Science.gov (United States)

Wang, Ten-See; Lin, Jeff; Ruf, Joe; Guidos, Mike

2012-01-01

There was interest in understanding the impact of out-of-round nozzle extension on the nozzle side load during transient startup operations. The out-of-round nozzle extension could be the result of asymmetric internal stresses, deformation induced by previous tests, and asymmetric loads induced by hardware attached to the nozzle. The objective of this study was therefore to computationally investigate the effect of out-of-round nozzle extension on the nozzle side loads during an engine startup transient. The rocket engine studied encompasses a regeneratively cooled chamber and nozzle, along with a film cooled nozzle extension. The computational methodology is based on an unstructured-grid, pressure-based computational fluid dynamics formulation, and transient inlet boundary flow properties derived from an engine system simulation. Six three-dimensional cases were performed with the out-of-roundness achieved by three different degrees of ovalization, elongated on lateral y and z axes: one slightly out-of-round, one more out-of-round, and one significantly out-of-round. The results show that the separation line jump was the primary source of the peak side loads. Comparing to the peak side load of the perfectly round nozzle, the peak side loads increased for the slightly and more ovalized nozzle extensions, and either increased or decreased for the two significantly ovalized nozzle extensions. A theory based on the counteraction of the flow destabilizing effect of an exacerbated asymmetrical flow caused by a lower degree of ovalization, and the flow stabilizing effect of a more symmetrical flow, created also by ovalization, is presented to explain the observations obtained in this effort.

Effects of injection nozzle exit width on rotating detonation engine

Science.gov (United States)

Sun, Jian; Zhou, Jin; Liu, Shijie; Lin, Zhiyong; Cai, Jianhua

2017-11-01

A series of numerical simulations of RDE modeling real injection nozzles with different exit widths are performed in this paper. The effects of nozzle exit width on chamber inlet state, plenum flowfield and detonation propagation are analyzed. The results are compared with that using an ideal injection model. Although the ideal injection model is a good approximation method to model RDE inlet, the two-dimensional effects of real nozzles are ignored in the ideal injection model so that some complicated phenomena such as the reflected waves caused by the nozzle walls and the reversed flow into the nozzles can not be modeled accurately. Additionally, the ideal injection model overpredicts the block ratio. In all the cases that stabilize at one-wave mode, the block ratio increases as the nozzle exit width gets smaller. The dual-wave mode case also has a relatively high block ratio. A pressure oscillation in the plenum with the same main frequency with the rotating detonation wave is observed. A parameter σ is applied to describe the non-uniformity in the plenum. σ increases as the nozzle exit width gets larger. Under some condition, the heat release on the interface of fresh premixed gas layer and detonation products can be strong enough to induce a new detonation wave. A spontaneous mode-transition process is observed for the smallest exit width case. Due to the detonation products existing in the premixed gas layer before the detonation wave, the detonation wave will propagate through reactants and products alternately, and therefore its strength will vary with time, especially near the chamber inlet. This tendency gets weaker as the injection nozzle exit width increases.

Manufacturing Process Developments for Regeneratively-Cooled Channel Wall Rocket Nozzles

Science.gov (United States)

Gradl, Paul; Brandsmeier, Will

2016-01-01

Regeneratively cooled channel wall nozzles incorporate a series of integral coolant channels to contain the coolant to maintain adequate wall temperatures and expand hot gas providing engine thrust and specific impulse. NASA has been evaluating manufacturing techniques targeting large scale channel wall nozzles to support affordability of current and future liquid rocket engine nozzles and thrust chamber assemblies. The development of these large scale manufacturing techniques focus on the liner formation, channel slotting with advanced abrasive water-jet milling techniques and closeout of the coolant channels to replace or augment other cost reduction techniques being evaluated for nozzles. NASA is developing a series of channel closeout techniques including large scale additive manufacturing laser deposition and explosively bonded closeouts. A series of subscale nozzles were completed evaluating these processes. Fabrication of mechanical test and metallography samples, in addition to subscale hardware has focused on Inconel 625, 300 series stainless, aluminum alloys as well as other candidate materials. Evaluations of these techniques are demonstrating potential for significant cost reductions for large scale nozzles and chambers. Hot fire testing is planned using these techniques in the future.

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)

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

Analysis Of Feedwater Line Break Of APR1400 By MARS Code

International Nuclear Information System (INIS)

Nguyen Thi Thanh Thuy; Le Dai Dien, Hoang Minh Giang

2011-01-01

This paper will deal with analysis of Feed water Line Break problem (FWLB) of the APR 1400 NPP with initial conditions: operation at 100% of power, double-ended break area of 0.058 m 2 and the break location of the feedwater line between the check valve and the steam generator. The analysis was simulated by MARS code through two step: calculation for steady state and calculation for transient state with initial condition mentioned. Some output result were presented with explanation: sequence of events corresponding to the time of the accident, the system behavior as temperature, pressure, steam generator water levels as well as DNBR, etc. before and after the accident. (author)

Pipe damping

International Nuclear Information System (INIS)

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

1984-01-01

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

Experimental study of subsonic microjet escaping from a rectangular nozzle

Science.gov (United States)

Aniskin, V. M.; Maslov, A. A.; Mukhin, K. A.

2016-10-01

The first experiments on the subsonic laminar microjets escaping from the nozzles of rectangular shape are carried out. The nozzle size is 83.3x3823 microns. Reynolds number calculated by the nozzle height and the average flow velocity at the nozzle exit ranged from 58 to 154. The working gas was air at room temperature. The velocity decay and velocity fluctuations along the center line of the jet are determined. The fundamental difference between the laminar microjets characteristics and subsonic turbulent jets of macro size is shown. Based on measurements of velocity fluctuations it is shown the presence of laminar-turbulent transition in microjets and its location is determined.

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

Thrust Augmentation by Airframe-Integrated Linear-Spike Nozzle Concept for High-Speed Aircraft

Directory of Open Access Journals (Sweden)

Hidemi Takahashi

2018-02-01

Full Text Available The airframe-integrated linear-spike nozzle concept applied to an external nozzle for high-speed aircraft was evaluated with regard to the thrust augmentation capability and the trim balance. The main focus was on the vehicle aftbody. The baseline airframe geometry was first premised to be a hypersonic waverider design. The baseline aftbody case had an external nozzle comprised of a simple divergent nozzle and was hypothetically replaced with linear-spike external nozzle configurations. Performance evaluation was mainly conducted by considering the nozzle thrust generated by the pressure distribution on the external nozzle surface at the aftbody portion calculated by computer simulation at a given cruise condition with zero angle of attack. The thrust performance showed that the proposed linear-spike external nozzle concept was beneficial in thrust enhancement compared to the baseline geometry because the design of the proposed concept had a compression wall for the exhaust flow, which resulted in increasing the wall pressure. The configuration with the boattail and the angled inner nozzle exhibited further improvement in thrust performance. The trim balance evaluation showed that the aerodynamic center location appeared as acceptable. Thus, benefits were obtained by employing the airframe-integrated linear-spike external nozzle concept.

Cause Analysis of Flow Accelerated Corrosion and Erosion-Corrosion Cases in Korea Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Lee, Y. S.; Lee, S. H. [Korea Hydro and Nuclear Power Co., Ltd., Gyeongju (Korea, Republic of); Hwang, K. M. [KEPCO Engineering and Construction Company, Gimcheon (Korea, Republic of)

2016-08-15

Significant piping wall thinning caused by Flow-Accelerated Corrosion (FAC) and Erosion-Corrosion (EC) continues to occur, even after the Mihama Power Station unit 3 secondary pipe rupture in 2004, in which workers were seriously injured or died. Nuclear power plants in many countries have experienced FAC and EC-related cases in steam cycle piping systems. Korea has also experienced piping wall thinning cases including thinning in the downstream straight pipe of a check valve in a feedwater pump line, the downstream elbow of a control valve in a feedwater flow control line, and failure of the straight pipe downstream of an orifice in an auxiliary steam return line. Cause analyses were performed by reviewing thickness data using Ultrasonic Techniques (UT) and, Scanning Electron Microscope (SEM) images for the failed pipe, and numerical simulation results for FAC and EC cases in Korea Nuclear Power Plants. It was concluded that the main cause of wall thinning for the downstream pipe of a check valve is FAC caused by water vortex flow due to the internal flow shape of a check valve, the main cause of wall thinning for the downstream elbow of a control valve is FAC caused by a thickness difference with the upstream pipe, and the main cause of wall thinning for the downstream pipe of an orifice is FAC and EC caused by liquid droplets and vortex flow. In order to investigate more cases, additional analyses were performed with the review of a lot of thickness data for inspected pipes. The results showed that pipe wall thinning was also affected by the operating condition of upstream equipment. Management of FAC and EC based on these cases will focus on the downstream piping of abnormal or unusual operated equipment.

Direct Numerical Simulation of Hypersonic Turbulent Boundary Layer inside an Axisymmetric Nozzle

Science.gov (United States)

Huang, Junji; Zhang, Chao; Duan, Lian; Choudhari, Meelan M.

2017-01-01

As a first step toward a study of acoustic disturbance field within a conventional, hypersonic wind tunnel, direct numerical simulations (DNS) of a Mach 6 turbulent boundary layer on the inner wall of a straight axisymmetric nozzle are conducted and the results are compared with those for a flat plate. The DNS results for a nozzle radius to boundary-layer thickness ratio of 5:5 show that the turbulence statistics of the nozzle-wall boundary layer are nearly unaffected by the transverse curvature of the nozzle wall. Before the acoustic waves emanating from different parts of the nozzle surface can interfere with each other and undergo reflections from adjacent portions of the nozzle surface, the rms pressure fluctuation beyond the boundary layer edge increases toward the nozzle axis, apparently due to a focusing effect inside the axisymmetric configuration. Spectral analysis of pressure fluctuations at both the wall and the freestream indicates a similar distribution of energy content for both the nozzle and the flat plate, with the peak of the premultiplied frequency spectrum at a frequency of [(omega)(delta)]/U(sub infinity) approximately 6.0 inside the free stream and at [(omega)(delta)]/U(sub infinity) approximately 2.0 along the wall. The present results provide the basis for follow-on simulations involving reverberation effects inside the nozzle.

Cleaning the feed-water pipeline internal surfaces

International Nuclear Information System (INIS)

Podkopaev, V.A.

1984-01-01

The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washing by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones

Cleaning the feed-water pipeline internal surfaces

Energy Technology Data Exchange (ETDEWEB)

Podkopaev, V.A.

1984-12-01

The procedure of cleaning the feed-water pipeline internal surfaces at the Chernobylsk-4 power unit is described. Cleaning was conducted in five stages. Pipelines were cleaned from mechanical impurities at the first stage. At the second stage the pipelines were washed by water heated up to 80 deg C. At the third stage nitric acid was added to 95-100 deg C water with the acid concentration in the circuit = 60 mg/l, purification period = 14 h. At the fourth stage hydrogen peroxide was added to the circuit at 95-100 deg C (the solution concentration was equal to 5-6 mg/l, the solution stayed in the circuit for 1 h 20 min). At the fifth stage sodium nitrite concentrated to 20 mg/l was introduced to the circuit in 75 minutes; this promoted strengthening of the oxide layer in the circuit on the base of nitric acid and hydrogen peroxide. Data on the water acidity in the circuit, water electric conductivity and iron concentration after the fourth stage and on completion of the circuit cleaning are presented. The described method of cleaning enables to save scarce reagents and use cheaper ones.

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.

Thermal-Hydraulic Performance of Scrubbing Nozzle Used for CFVS

Energy Technology Data Exchange (ETDEWEB)

Lee, Hyun Chul; Lee, Doo Yong; Jung, Woo Young; Lee, Jong Chan; Kim, Gyu Tae [FNC TECH, Yongin (Korea, Republic of)

2016-05-15

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.

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

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

Heat pipe

International Nuclear Information System (INIS)

Triggs, G.W.; Lightowlers, R.J.; Robinson, D.; Rice, G.

1986-01-01

A heat pipe for use in stabilising a specimen container for irradiation of specimens at substantially constant temperature within a liquid metal cooled fast reactor, comprises an evaporator section, a condenser section, an adiabatic section therebetween, and a gas reservoir, and contains a vapourisable substance such as sodium. The heat pipe further includes a three layer wick structure comprising an outer relatively fine mesh layer, a coarse intermediate layer and a fine mesh inner layer for promoting unimpeded return of condensate to the evaporation section of the heat pipe while enhancing heat transfer with the heat pipe wall and reducing entrainment of the condensate by the upwardly rising vapour. (author)

Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

International Nuclear Information System (INIS)

Boucher, T.J.; Owca, W.A.

1985-05-01

This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

Microalgal cell disruption via ultrasonic nozzle spraying.

Science.gov (United States)

Wang, M; Yuan, W

2015-01-01

The objective of this study was to understand the effect of operating parameters, including ultrasound amplitude, spraying pressure, nozzle orifice diameter, and initial cell concentration on microalgal cell disruption and lipid extraction in an ultrasonic nozzle spraying system (UNSS). Two algal species including Scenedesmus dimorphus and Nannochloropsis oculata were evaluated. Experimental results demonstrated that the UNSS was effective in the disruption of microalgal cells indicated by significant changes in cell concentration and Nile red-stained lipid fluorescence density between all treatments and the control. It was found that increasing ultrasound amplitude generally enhanced cell disruption and lipid recovery although excessive input energy was not necessary for best results. The effect of spraying pressure and nozzle orifice diameter on cell disruption and lipid recovery was believed to be dependent on the competition between ultrasound-induced cavitation and spraying-generated shear forces. Optimal cell disruption was not always achieved at the highest spraying pressure or biggest nozzle orifice diameter; instead, they appeared at moderate levels depending on the algal strain and specific settings. Increasing initial algal cell concentration significantly reduced cell disruption efficiency. In all UNSS treatments, the effectiveness of cell disruption and lipid recovery was found to be dependent on the algal species treated.

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.

EURCYL. A program to generate finite element meshes for pressure vessel nozzles

International Nuclear Information System (INIS)

De Windt, P.; Reynen, J.

1974-12-01

EURCYL is a program dealing with the automatic generation of finite element meshes for pressure vessel nozzles, using isoparametric elements with 8, 20 or 32 nodes. Options exist to generate BWR nozzles as well as PWR nozzles

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.

Crack of reactor vessel upper head penetration nozzles in Korean nuclear plants

Energy Technology Data Exchange (ETDEWEB)

Doh, E.; Lee, T-S.; Kim, J-Y.; Lee, C-H. [KEPCO Plant Service and Engineering Co., Ltd., Busan (Korea, Republic of)

2014-07-01

Since the first CRDM nozzles of reactor vessel head at Kori unit 1 in Korea were inspected in 2003, no CRDM nozzle cracks had been revealed prior to the inspection at Hanbit unit 3 in October 2012, even though many foreign plants had been reporting PWSCC cracks. In October 2012, seven axial cracks from 6 CRDM nozzles at Hanbit unit 3, and in November 2013, six axial cracks from 6 CRDM nozzles at Hanbit unit 4 were detected by TOFD Ultrasonic testing from ID of nozzles. There were confirmed to be PWSCC by Dye penetrant testing and Replica on the surface of J-groove weld of CRDM nozzles. Both plants are OPR-1000 types. All flaws started from the surface of J-groove weld at interface with OD of nozzle, but did not grow up to the top of J-groove weld, and did not make any Leak path up to head outside. The Performance Demonstration Initiative (PDI) system of CRDM nozzle inspection for Westinghouse type plants has been applied in Korea since July 2011. However, its application for OPR-1000 is still under development in Korea. The experience of PDI inspection for Westinghouse type plant contributed greatly to the detection and evaluation of PWSCC of CRDM nozzles at OPR- 1000 of Hanbit unit 3 & 4. The experimentally based procedure of flaw detection and the enhanced detection technique of examiners made it possible to detect and to determine the PWSCC indications. Embedded Flaw Repair process was approved by government authority, and the repair of the 6 CRDM nozzles in each plant was conducted by a consortium of Westinghouse and KPS. (author)