WorldWideScience

Sample records for reactor feedwater nozzle

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

    International Nuclear Information System (INIS)

    Hattori, Tsunekazu.

    1984-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

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

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

    International Nuclear Information System (INIS)

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

    1997-03-01

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

  4. Feedwater recycling system in BWR type reactor

    International Nuclear Information System (INIS)

    Shimamoto, Yoshiharu.

    1980-01-01

    Purpose: To improve the reactor safety by preventing thermal stresses and cracks generated in structural materials due to the fluctuations in the temperature for high temperature water - low temperature water mixture near the feedwater nozzle. Method: Feedwater pipes are connected to a pressure vessel not directly but by way of a flow control valve. While the recycled water is circulated from an inlet nozzle to an outlet nozzle through a recycle pump, flow control valve and recycling pipeways, feedwater is fed from the feedwater pipes to the recycling pipeways by way of the flow control valve. More specifically, since the high temperature recycle water and the low temperature recycle water are mixed within the pipeways, the temperature fluctuations resulted from the temperature difference between the recycle water and the feedwater is reduced to prevent thermal fatigue and generation of cracks thereby securing the reactor safety. (Furukawa, Y.)

  5. 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.)

  6. 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.)

  7. 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.)

  8. 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

  9. BWR feedwater nozzle and control-rod-drive return line nozzle cracking

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    In its 1978 Annual Report to Congress, the Nuclear Regulatory Commission identified as an unresolved safety issue the appearance of cracks in feedwater nozzles at boiling-water reactors (BWRs). Later similar cracking, detected in return water lines for control-rod-drive systems at BWRs, was designated Part II of the issue. This article outlines the resolution of these cracking problems

  10. 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.)

  11. 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.)

  12. 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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  14. Reactor feedwater system

    International Nuclear Information System (INIS)

    Hikabe, Katsumi.

    1978-01-01

    Purpose: In order to prevent thermal stresses of a core of PWR type reactor, described has been a method for feeding heated recirculating water to the core in the case of the reactor start-up or shut-down. Constitution: A recirculating water is degassed, cleaned up and heated in the steam condensers, and then feeds the water to the reactor, characterized in that heaters are provided in the bypasses of the turbine, so that heated water is constantly supplied to the reactor. (Nakamura, S.)

  15. Reactor feedwater control device

    International Nuclear Information System (INIS)

    Koshi, Yuji.

    1993-01-01

    In the device of the present invention, an excess response is not caused in a reactor feed water system even when voids are fluctuated by using an actual water level signal as a reactor water level signal. That is, a standard water level signal and a reactor water level signal are inputted to a comparator. An adder adds water level difference signal outputted from the comparator and mismatch flow rate signal prepared by multiplying the difference between a main steam flow rate signal and a feed water flow rate signal by a mismatch gain. A feed water controller integrates the added signal and outputs flow rate demand signal. A feed water system receives the flow rate demand signal as input. A water level calculation means is disposed to such a device for calculating an actual water level based on the change of coolant possessing amount of the reactor, and the output thereof is defined as a reactor water level signal. With such procedures, excessive elevation of water level of the reactor can be prevented even upon occurrence of void fluctuation phenomenon or the like in the reactor such as upon sole scram operation. Accordingly, plant shut down caused thereby can be avoided safely. (I.S.)

  16. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

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

    International Nuclear Information System (INIS)

    Tsai, Y.L.; Wang, Li. H.; Fan, T.W.; Ranganath, Sam; Wang, C.K.; Chou, C.P.

    2010-01-01

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

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

    International Nuclear Information System (INIS)

    Arthur, D.

    1978-03-01

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

  1. Nozzle for electric dispersion reactor

    Science.gov (United States)

    Sisson, W.G.; Basaran, O.A.; Harris, M.T.

    1995-11-07

    A nozzle for an electric dispersion reactor includes two concentric electrodes, the inner one of the two delivering disperse phase fluid into a continuous phase fluid. A potential difference generated by a voltage source creates a dispersing electric field at the end of the inner electrode. 4 figs.

  2. Development of methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWRs

    International Nuclear Information System (INIS)

    Shvarts, S.; Gerber, D.A.; House, K.; Hirschberg, P.

    1994-01-01

    The objective of this paper is to describe a methodology for evaluating and monitoring steam generator feedwater nozzle cracking in PWR plants. This methodology is based in part on plant test data obtained from a recent Diablo Canyon Power Plant (DCPP) Unit 1 heatup. Temperature sensors installed near the nozzle-to-pipe weld were monitored during the heatup, along with operational parameters such as auxiliary feedwater (AFW) flow rate and steam generator temperature. A thermal stratification load definition was developed from this data. Steady state characteristics of this data were used in a finite element analysis to develop relationship between AFW flow and stratification interface level. Fluctuating characteristics of this data were used to determine transient parameters through the application of a Green's Function approach. The thermal stratification load definition from the test data was used in a three-dimensional thermal stress analysis to determine stress cycling and consequent fatigue damage or crack growth during AFW flow fluctuations. The implementation of the developed methodology in the DCPP and Sequoyah Nuclear Plant (SNP) fatigue monitoring systems is described

  3. Feedwater control system in BWR type reactor

    International Nuclear Information System (INIS)

    Tanji, Jun-ichi; Oomori, Takashi.

    1980-01-01

    Purpose: To improve the water level control performance in BWR type reactor by regulating the water level set to the reactor depending on the rate of change in the recycling amount of coolant to thereby control the fluctuations in the water level resulted in the reactor within an aimed range even upon significant fluctuations in the recycling flow rate. Constitution: The recycling flow rate of coolant in the reactor is detected and the rate of its change with time is computed to form a rate of change signal. The rate of change signal is inputted to a reactor level setter to amend the actual reactor water level demand signal and regulate the water level set to the reactor water depending on the rate of change in the recycling flow rate. Such a regulation method for the set water level enables to control the water level fluctuation resulted in the reactor within the aimed range even upon the significant fluctuation in the recycling flow rate and improve the water level control performance of the reactor, whereby the operationability for the reactor is improved to enhance the operation rate. (Moriyama, K.)

  4. 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

  5. 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

  6. 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...

  7. Reactor vessel nozzle cracks: a photoelastic study

    International Nuclear Information System (INIS)

    Smith, C.W.

    1979-01-01

    A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

  8. 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

  9. 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

  10. Head spray nozzle in reactor pressure vessel

    International Nuclear Information System (INIS)

    Hatano, Shun-ichi.

    1990-01-01

    In a reactor pressure vessel of a BWR type reactor, a head spray nozzle is used for cooling the head of the pressure vessel and, in view of the thermal stresses, it is desirable that cooling is applied as uniformly as possible. A conventional head spray is constituted by combining full cone type nozzles. Since the sprayed water is flown down upon water spraying and the sprayed water in the vertical direction is overlapped, the flow rate distribution has a high sharpness to form a shape as having a maximum value near the center and it is difficult to obtain a uniform flow rate distribution in the circumferential direction. Then, in the present invention, flat nozzles each having a spray water cross section of laterally long shape, having less sharpness in the circumferential distribution upon spraying water to the inner wall of the pressure vessel and having a wide angle of water spray are combined, to make the flow rate distribution of spray water uniform in the inner wall of the pressure vessel. Accordingly, the pressure vessel can be cooled uniformly and thermal stresses upon cooling can be decreased. (N.H.)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. 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.)

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  18. 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.)

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. 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)

  1. 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

  2. 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.)

  3. 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)

  4. 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.)

  5. 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)

  6. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2

    International Nuclear Information System (INIS)

    Tijerina S, F.

    2008-01-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  7. 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

  8. Gas Nozzle Effect on the Deposition of Polysilicon by Monosilane Siemens Reactor

    Directory of Open Access Journals (Sweden)

    Seung Oh Kang

    2012-01-01

    Full Text Available Deposition of polysilicon (poly-Si was tried to increase productivity of poly-Si by using two different types of gas nozzle in a monosilane Bell-jar Siemens (MS-Siemens reactor. In a mass production of poly-Si, deposition rate and energy consumption are very important factors because they are main performance indicators of Siemens reactor and they are directly related with the production cost of poly-Si. Type A and B nozzles were used for investigating gas nozzle effect on the deposition of poly-Si in a MS-Siemens reactor. Nozzle design was analyzed by computation cluid dynamics (CFD. Deposition rate and energy consumption of poly-Si were increased when the type B nozzle was used. The highest deposition rate was 1 mm/h, and the lowest energy consumption was 72 kWh⋅kg-1 in this study.

  9. 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

  10. 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.)

  11. Application of fuzzy logic control system for reactor feed-water control

    International Nuclear Information System (INIS)

    Iijima, T.; Nakajima, Y.

    1994-01-01

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  12. Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

    International Nuclear Information System (INIS)

    Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes

    2017-01-01

    This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

  13. Reliability study of the auxiliary feed-water system of a pressurized water reactor by faults tree and Bayesian Network

    Energy Technology Data Exchange (ETDEWEB)

    Lava, Deise Diana; Borges, Diogo da Silva; Guimarães, Antonio Cesar Ferreira; Moreira, Maria de Lourdes, E-mail: deise_dy@hotmail.com, E-mail: diogosb@outlook.com, E-mail: tony@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    This paper aims to present a study of the reliability of the Auxiliary Feed-water System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10{sup -3}. (author)

  14. Underwater laser beam welding technology for reactor vessel nozzles of PWRs

    International Nuclear Information System (INIS)

    Yoda, Masaki; Tamura, Masataka; Tamura, Masataka

    2010-01-01

    Toshiba has developed an underwater laser beam welding technology for the maintenance of reactor vessel nozzles of pressurized water reactors (PWRs), which eliminates the need for the drainage of water from the reactor vessel. The new welding system makes it possible to both reduce the work period and minimize the radiation exposure of workers compared with conventional technologies for welding in ambient air. We have confirmed the effectiveness of this technology through experiments in which stress corrosion cracking (SCC) was mitigated on the inner surfaces of nozzles. We are promoting its practical application in Japan and overseas in cooperation with Westinghouse Electric Company, a group company of Toshiba. (author)

  15. Experiments on the spray nozzles used in the pressurizer of power reactor

    International Nuclear Information System (INIS)

    Diao Wentang

    1989-04-01

    The spray nozzle, which is used in the pressurizer of pressurized water reactor system, usually uses a less differential pressure between the reactor inlet and outlet as the spray drive pressure, but its flow rate is relatively larger. It is difficult to obtain a optimum spray performance of such a nozzle. The experimental results of five types of twenty seven spray nozzles in different structures and sizes with the range of the spray drive pressure from 0.127 to 0.245 MPa and the flow rates from 5 to 50 t/h are given. The main factors affecting spray performances and their distribution characteristics have been found. And some relatively suitable spray structures have been recommended, which can be used as references for improving the spray nozzles used in the pressurizers of existing PWRs or of the PWRs to be built

  16. 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

  17. Nuclear reactor fuel assembly with a removably top nozzle

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1985-01-01

    The invention relates to a nuclear fuel assembly having an improved attaching structure for removably mounting the top nozzle of the fuel assembly on the upper end of a control-rod guide thimble. The attaching structure comprises an outer socket defined in a portion of the top nozzle, an inner socket extending from the upper end of the guide thimble and removably received in the outer socket for interlocking engagement therewith, and an elongate locking member adapted to be inserted into the inner socket to maintain said interlocking engagement. Removal of the locking member from the inner socket enables the latter to be withdrawn from the outer socket, thereby enabling the top nozzle to be removed from the guide thimble

  18. 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)

  19. Nuclear reactor fuel assembly with a removable top nozzle

    International Nuclear Information System (INIS)

    Shallenberger, J.M.; Ferlan, S.J.

    1986-01-01

    This patent describes a fuel assembly having at least one control rod guide thimble and a top nozzle, the top nozzle including a transversely extending adapter plate. An improved attaching structure is described for removably mounting the top nozzle on the guide thimble comprising: (a) means defining an outer socket in the top nozzle, the outer socket defining means including a passageway extending through the adapter plate and having a first mating element defined in the adapter plate within the passageway; (b) means on an upper end of the guide thimble defining an inner socket, the inner socket defining means including an elongated sleeve having an upper end portion. The upper end portion of the sleeve has a second mating element formed thereon and at least one elongated axial slot defined therein for permitting radial movement of the sleeve upper end portion between a compressed releasing position for removing and inserting the inner socket from and into the outer socket and an expanded locking position for locking the inner and outer sockets together

  20. Nuclear research reactor IEA-R1 heat exchanger inlet nozzle flow - a preliminary study

    International Nuclear Information System (INIS)

    Angelo, Gabriel; Andrade, Delvonei Alves de; Fainer, Gerson; Angelo, Edvaldo

    2009-01-01

    As a computational fluid mechanics training task, a preliminary model was developed. ANSYS-CFX R code was used in order to study the flow at the inlet nozzle of the heat exchanger of the primary circuit of the nuclear research reactor IEA-R1. The geometry of the inlet nozzle is basically compounded by a cylinder and two radial rings which are welded on the shell. When doing so there is an offset between the holes through the shell and the inlet nozzle. Since it is not standardized by TEMA, the inlet nozzle was chosen for a preliminary study of the flow. Results for the proposed model are presented and discussed. (author)

  1. Development of Reactor Vessel Bottom Mount Instrumentation Nozzle Routine Inspection Device

    Energy Technology Data Exchange (ETDEWEB)

    Khaled, Atya Ahmed Abdallah; Ihn, Namgung [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-10-15

    The primary coolant water of pressurized water reactors has created cracks in j-weld of penetrations with Alloy 600 through a process called primary water stress corrosion cracking. On October 6, 2013, BMI nozzle number 3 at Palo Verde Unit 3 (PVNGS-3) exhibited small white de-posits around the annulus. Nozzle attachment to the RV lower head is by J-groove weld to the inside penetration of the nozzle and the weld material is of Alloy 600 material. Above two cases clearly show the necessity of routine inspection of RV lower head penetration during refueling outage. Nondestructive inspection is generally performed to detect fine cracks or defects that may develop during operation. Defects usually occur at weld regions, hence most non-destructive inspection is to scan and check any defects or crack in the weld region. BMI nozzles at the bottom head of a nuclear reactor vessel (RV) are one of such area for inspection. But BMI nozzles have not been inspected during regular refuel outage due to the relative small size of BMI nozzle and limited impact of the consequences of BMI leak. However, there is growing concern since there have been leaks at nuclear power plants (NPPs) as well as recent operating experience. In this study, we propose a system that is conveniently used for nondestructive inspection of BMI nozzles during regular refueling outage without removing all the reactor internals. A 3D model of the inspection system was also developed along with the RV and internals which permits a virtual 3D simulation to check the design concept and usability of the system.

  2. Assessment of pressurized water reactor control rod drive mechanism nozzle cracking

    International Nuclear Information System (INIS)

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

    1994-10-01

    This report surveys the field experience related to cracking of pressurized water reactor (PWR) control rod drive mechanism nozzles (Alloy 600 material); evaluates design, fabrication, and operating conditions for the nozzles in US PWR; and evaluates the safety significance of nozzle cracking. Inspection at 78 overseas and one US PWR has revealed mainly axial cracks in 101 nozzles. The cracking is caused by primary water stress corrosion cracking, which requires the simultaneous presence of high tensile stresses, high operating temperatures, and susceptible microstructure. CRDM nozzle cracking is not a short-term safety issue. An axial crack is not likely to grow above the vessel head to a critical length because the stresses are not high enough to support the growth away from the attachment weld. Primary coolant leaking through an axial crack could cause a short circumferential crack on the outside surface. However, this crack is not likely to propagate through the nozzle wall to cause rupture. Leakage of the primary coolant from a through-wall crack could cause boric acid corrosion of the vessel head and challenge the structural integrity of the head, but it is very unlikely that the accumulated deposits of boric acid crystals resulting from such leakage could remain undetected

  3. Characterization of the full cone pressure swirl spray nozzles for the nuclear reactor containment spray system

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); John, Benny [Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2014-07-01

    Highlights: • Full cone spray pressure swirl nozzle with X-Vane is studied. • Laser illuminated imaging technique is used. • Correlations for coefficient of discharge, spray cone angle and SMD are suggested. • Droplet size and mass fraction distribution is measured. • Inviscid theory predicts the coefficient of discharge. - Abstract: The objective of the present study is to characterize a full cone pressure swirl nozzle for the Containment Spray System (CSS) of Indian Pressurized heavy Water reactors (IPHWR). The influence of Reynolds number and geometric parameters on the coefficient of discharge, spray cone angle, mass flux density distribution, droplet size distribution, Sauter mean diameter (SMD is studied for full cone pressure swirl full cone nozzles. The nozzles of orifice diameter range from 1.3 to 7.2 mm are studied. Experiments are conducted with water at room temperature as the working medium. The nozzles are operated with the pressure ranging from 1 to 8 bar. The measurements of the drop size distributions are performed with laser illuminated imaging technique. The spray cone-angle of the full cone nozzles is measured by the evaluation of images recorded with a camera using IMAGE J software. Correlations for coefficient of discharge, spray cone angle and Sauter mean diameter are suggested on the basis of the experimental results. Rosin–Rammler model and Nukiyama–Tanasawa distributions predict the mass fraction distribution reasonably well. However, the droplet size distribution is predicted by Nukiyama-Tanasawa model only.

  4. Optimization of the size and shape of the set-in nozzle for a PWR reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Murtaza, Usman Tariq, E-mail: maniiut@yahoo.com; Javed Hyder, M., E-mail: hyder@pieas.edu.pk

    2015-04-01

    Highlights: • The size and shape of the set-in nozzle of the RPV have been optimized. • The optimized nozzle ensure the reduction of the mass around 198 kg per nozzle. • The mass of the RPV should be minimized for better fracture toughness. - Abstract: The objective of this research work is to optimize the size and shape of the set-in nozzle for a typical reactor pressure vessel (RPV) of a 300 MW pressurized water reactor. The analysis was performed by optimizing the four design variables which control the size and shape of the nozzle. These variables are inner radius of the nozzle, thickness of the nozzle, taper angle at the nozzle-cylinder intersection, and the point where taper of the nozzle starts from. It is concluded that the optimum design of the nozzle is the one that minimizes the two conflicting state variables, i.e., the stress intensity (Tresca yield criterion) and the mass of the RPV.

  5. Development of ABWR inertia-increased reactor internal pump and thicker sleeve nozzle

    International Nuclear Information System (INIS)

    Takahashi, Shirou; Shiina, Kouji; Matsumura, Seiichi

    2002-01-01

    The conventional reactor internal pumps (RIPs) in the ABWR have an inertia moment coming from the shafts and Motor-Generator sets, enabling the RIPs to continue running for a few seconds, when a trip of all RIPs event occurs. It is possible to simplify the RIPs' power supply system without affecting the core flow supply when the above event occurs by eliminating M-G sets, if the rotating inertia is increased. This inertia increase due to an additional flywheel, which leads to gains in weight and length, requires the larger diameter nozzle with the thicker sleeve. However, too large a nozzle diameter may change the hydraulic performance. In authors' previous study, the optimum nozzle diameter (492 mm) was selected through 1/5-scale test. In this study, the 492 mm nozzle and the inertia-increased RIP were verified through the full-scale tests. The rotating inertia time constant on coastdown characteristics (behavior of the RIP speed in the event of power loss) for the inertia-increased RIP doubled compared with the current RIP. The casing and the shaft vibration were also confirmed to satisfy the design criteria. Moreover, hydraulic performance and heat increase in the motor casing due to the flywheel were evaluated. The inertia increased RIP with the 492 mm nozzle maintained good performance. (author)

  6. Corrosion Damage in Penetration Nozzle and Its Weldment of Reactor Pressure Vessel Head

    International Nuclear Information System (INIS)

    Lim, Yun Soo; Kim, Joung Soo; Kim, Hong Pyo; Hwang, Seong Sik; Yi, Young Sun; Kim, Dong Jin; Jung, Man Kyo

    2003-07-01

    The recent status on corrosion damage of reactor vessel head (RVH) penetration nozzles at primary water reactors (PWRs), including control rod drive mechanism (CRDM) and thermocouple nozzles, was investigated. The studies for primary water stress corrosion cracking (PWSCC) characteristics of Alloy 600 and Alloy 182/82 were reviewed and summarized in terms of the crack initiation and crack growth rate. The studies on the boric acid corrosion (BAC) of low alloy steels were also included in this report. PWSCC was found to be the main failure mechanism of RVH CRDM nozzles, which are constituted with Alloy 600 base metal and Alloy 182 weld filler materials. Alloy 600 and Alloy 182/82 are very susceptible to intergranular SCC in the PWR environments. The PWSCC crack initiation and growth features in the fusion zone of Alloy 182/82 were strongly dependant on solidification anisotropy during welding, test temperature, weld heat, mechanical loading, stress relief heat treatment, cold work and so on. BAC of low alloy steels is a wastage phenomenon due to general corrosion occurring on the over-all surface area of material. Systematic studies, concerned with structural integrity of RVH penetration nozzles as well as improvement of PWSCC resistance of nickel-based weld metals in the simulated PWR environment, are needed

  7. Resolution of concerns in auxiliary feedwater piping

    International Nuclear Information System (INIS)

    Bain, R.A.; Testa, M.F.

    1994-01-01

    Auxiliary feedwater piping systems at pressurized water reactor (PWR) nuclear power plants have experienced unanticipated operating conditions during plant operation. These unanticipated conditions have included plant events involving backleakage through check valves, temperatures in portions of the auxiliary feedwater piping system that exceed design conditions, and the occurrence of unanticipated severe fluid transients. The impact of these events has had an adverse effect at some nuclear stations on plant operation, installed plant components and hardware, and design basis calculations. Beaver Valley Unit 2, a three loop pressurized water reactor nuclear plant, has observed anomalies with the auxiliary feedwater system since the unit went operational in 1987. The consequences of these anomalies and plant events have been addressed and resolved for Beaver Valley Unit 2 by performing engineering and construction activities. These activities included pipe stress, pipe support and pipe rupture analysis, the monitoring of auxiliary feedwater system temperature and pressure, and the modification to plant piping, supports, valves, structures and operating procedures

  8. 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.)

  9. 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

  10. 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

  11. 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

  12. Residual Stress Evaluation of Weld Inlay Process on Reactor Vessel Nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Kihyun; Cho, Hong Seok [KEPCO KPS, Naju (Korea, Republic of)

    2015-10-15

    Weld overlay, weld inlay and stress improvement are mitigation technologies for butt joints. Weld overlay is done on pressurizer nozzles which are the highest potential locations occurring PWSCC due to high temperature in Korea. Reactor vessel nozzles are other big safety concerns for butt joints. Weld overlay and stress improvement should be so difficult to apply to those locations because space is too limited. Weld inlay should be one of the solutions. KEPCO KPS has developed laser welding system and process for reactor nozzles. Welding residual stress analysis is necessary for flaw evaluation. United States nuclear regulatory commission has calculated GTAW(Gas Tungsten Arc Welding) residual stress using ABAQUS. To confirm effectiveness of weld inlay process, welding residual stress analysis was performed. and difference between GTAW and LASER welding process was compared. Evaluation of weld inlay process using ANSYS and ABAQUS is performed. All of the both results are similar. The residual stress generated after weld inlay was on range of 450-500 MPa. Welding residual stresses are differently generated by GTAW and LASER welding. But regardless of welding process type, residual tensile stress is generated on inside surface.

  13. Determination of welding parameters for execution of weld overlayer on PWR nuclear reactor nozzles

    International Nuclear Information System (INIS)

    Ribeiro, Gabriela M.; Lima, Luciana I.; Quinan, Marco A.; Schvartzman, Monica M.

    2009-01-01

    In the PWR reactors, nickel based dissimilar welds have been presented susceptibilities the stress corrosion (S C). For the mitigation the problem a deposition of weld layers on the external surface of the nozzle is an alternative, viewing to provoke the compression of the region subjected to S C. This paper presents a preliminary study on the determination of welding parameters to obtain these welding overlayers. Welding depositions were performed on a test piece welded with nickel 182 alloy, simulating the conditions of a nozzle used in a PWR nuclear power plant. The welding process was the GTAW (Gas Tungsten Arc Welding), and a nickel 52 alloy as addition material. The overlayers were performed on the base metals, carbon steel an stainless steel, changing the welding parameters and verifying the the time of each weld filet. After that, the samples were micro structurally characterized. The macro structures and the microstructures obtained through optical microscopy and Vickers microhardness are presented. The preliminary results make evident the good weld quality. However, a small weld parameters influence used in the base material microstructure (carbon steel and stainless steel). The obtained results in this study will be used as reference in the construction of a mock up which will simulate all the conditions of a pressurizer nozzle of PWR reactor

  14. Feedwater device for nuclear power plant

    International Nuclear Information System (INIS)

    Ikekita, Iwao.

    1980-01-01

    Purpose: To conduct water feeding without using high pressure steam of the reactor and with no radiation exposure by the provision of each feedwater pump driven by each motor controlled from variable frequency thyristor-inverter to a feedwater pipe connecting a condensate pump and the reactor. Constitution: High pressure steams resulted from heat exchange in the reactor core are transferred by way of a main steam check valve in a main steam pipe to a high pressure turbine, drive the high pressure turbine, flow out of the turbine and then drive a low pressure turbine by way of a moisture separator. The steams thus used for the turbine driving are condensed in a condensator and then sent under pressure by way of each condensating pump to a feedwater pipe. Since each of the feedwater pumps provided in the route of the feedwater pipe is driven by each of the motors under the control of the variable frequency thyristor-inverter in starting, shut down and normal operation, water is fed to the reactor. (Horiuchi, T.)

  15. Experiments and analysis of thermal stresses around the nozzle of the reactor vessel

    International Nuclear Information System (INIS)

    Song, D.H.; Oh, J.H.; Song, H.K.; Park, D.S.; Shon, K.H.

    1981-01-01

    This report describes the results of analysis and experiments on the thermal stress around the reactor vessel nozzle performed to establish a capability of thermal stress analysis of pressure vessel subjected to thermal loadings. Firstly, heat conduction analysis during reactor design transients and analysis on the experimental model were performed using computer code FETEM-1 for the purpose of verification of FETEM-1 which was developed in 1979 and will be used to obtain the temperature distribution in a solid body under the steady-state and the transient conditions. The results of the analysis was compared to the results in the Stress Report of Kori-1 reactor vessel and those from experiments on the model, respectively

  16. Removable top nozzle and tool for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

    1986-01-01

    A fuel assembly is described for a nuclear reactor including a bottom nozzle, at least one longitudinally extending control rod guide thimble having an upper end and a lower end being attached to the bottom nozzle and projecting upwardly therefrom, transverse grids axially spaced along-the thimble for supporting an array of upstanding fuel rods, and a top nozzle subassembly removable mounted on the upper end of the guide thimble for obtaining top access to the fuel rods upon removal thereof. The top nozzle subassembly consists of: (a) a section integrally formed on the upper end of the guide thimble and having external threads thereon; (b) a lower adapter plate having a guide thimble hole for receiving the guide thimble so as to mount the adapter plate on the guide thimble for slidable movement therealong; (c) a retainer mounted on the guide thimble for restably supporting and limiting the downward movement of the adapter plate along the guide thimble; (d) an upper hold-down plate having a guide thimble passageway with an internal ledge for receiving the thimble so as to mount the hold-down plate on the thimble for slidable movement therealong; (e) spring means interposed between the upper hold-down plate and the lower adapter plate for biasing the hold-down plate upwardly when a downward force is applied thereon whereby the downward force is yieldably transmitted to the fuel assembly; and (f) a collar disposed within the passageway and in abutment with the ledge, the collar having an internal threaded section engageable with the externally threaded section to move the hold-down plate down against the spring means and thereby mounting of the subassembly on the guide thimble

  17. 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)

  18. 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)

  19. The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    The design features of a once-through steam generator (OTSG) for an integral reactor are significantly different from the commercial U-tube type steam generator from several aspects such as the general arrangement, size, operation conditions, and so on. Therefore a sufficient understanding of the thermal-hydraulic characteristics of the OTSG is essential for the design of the nuclear steam supply system (NSSS) and the power conversion system (PCS). It is also necessary to develop operation procedures complying to the unique design features of the OTSG of interest. The OTSG is sized to produce a sufficiently superheated steam during a normal power operation and therefore the secondary system can be simple relative to that of the other types of steam generators. For the plant adopting the OTSG, the steam pressure in the secondary circuit (tube side of the OTSG) is controlled to be constant during a normal power operation. Constant steam pressure is realized by regulating the control valve on the main steam line dedicated for this purpose. However during a heatup operation, at which the fluid state at the exit of the OTSG is a single phase hot water or two phases, it is not proper to use the control valve on the main steam line due to a control problem at low and multi-phase flow conditions and possibly an erosion problem. For these reasons, another dedicated line called a startup cooling line is used during a heatup condition. There may be several operational conditions for the secondary fluid required to pass through during heatup operation, depending on the design of the PCS. In general, there are two conditions: One is a condition for a vacuum operation for the condenser and another is an entry condition for a steam pressure control operation for an auxiliary power system. In this study, the concept of using a simple startup cooling line with a fixed flow resistance and changing the feedwater flow for the pressure control of the PCS during a heatup period are

  20. 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

  1. 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.)

  2. 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.

  3. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

    2000-01-01

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  4. 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

  5. Method and apparatus for removably mounting a top nozzle on a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Gjertsen, R.K.; Schallenberger, J.M.

    1986-01-01

    In a fuel assembly having a top nozzle and control rod guide thimbles, a method is described of removably mounting the top nozzle on the ends of the guide thimbles, comprising the steps of: (a) releasably mating hollow outer sockets defined in the top nozzle with hollow inner sockets defined on the ends of the guide thimbles. The inner sockets are movable between compressed conditions for removing and inserting the inner sockets from and into the outer sockets in mounting and removing the top nozzle on and from the guide thimbles and expanded conditions for mating the inner and outer sockets together and the top nozzle on the guide thimbles; (b) supporting elongated locking tubes such that end portions thereof extend into the outer sockets defined in the top nozzle; and (c) moving all of the locking tubes at the same time between unlocking and locking positions to displace their end portions axially within the outer sockets between first and second locations

  6. 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

  7. 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.

  8. Underwater cutting of reactor core internals by CO laser using local-dry-zone creating nozzle

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Osa (Mitsubishi Heavy Industries Ltd., Takasago, Hyogo (Japan). Takasago Research and Development Center); Sugihara, Masaaki; Miya, Kenzo

    1992-11-01

    With a view to practical application of the CO laser to underwater cutting of thick steel plates, a nozzle for creating a local dry zone on the workpiece has been developed and tested. The nozzle directed against the workpiece surface discharges a jet of air, which forms the local dry zone, bounded by a cone of high-speed water jet discharged from a concentric annular outlet. Preliminary tests were performed to optimize the nozzle shape and operating conditions. The resulting nozzle was used with a 5 kW CO laser for actual underwater cutting tests on stainless steel plates: Entirely satisfactory cutting performance was confirmed on various workpiece geometries and working positions. (author).

  9. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2; Medidores ultrasonicos en el flujo de agua de alimentacion para recuperar potencia termica en el reactor de la Central Nuclear Laguna Verde U1 and U2

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F. [CFE, Central Laguna Verde, Km. 42.5 Carretera Cardel-Nautla, Veracruz (Mexico)]. e-mail: francisco.tijerina@cfe.gob.mx

    2008-07-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  10. In reactor measurements, modeling and assessments to predict liquid injection shutdown system nozzle to Calandria tube time to contact

    International Nuclear Information System (INIS)

    Kirstein, K.; Kalenchuk, D.

    2011-01-01

    Over the past few years there has been an expanding effort to assess the potential for Calandria Tubes (CTs) coming into contact with Liquid Injection Shutdown System (LISS) Nozzles to ensure continued contact-free operation as required by CSA N285.4. LISS Nozzles (LINs), which run perpendicular to and between rows of fuel channels, sag at a slower rate than the fuel channels. As a result certain LINs may come in contact with CTs above them. The CT/LIN gaps can be predicted from calculated CT sag, LIN sag and a number of component and installation tolerances. This method however results in very conservative predictions when compared to measurements, confirmed with the in reactor measurements initiated in 2000, when gaps were successfully measured the first time using images obtained from a camera-assisted measurement tool inserted into the calandria. To reduce the conservatism of the CT/LIN gap predictions, statistical CT/LIN gap models are used instead. They are derived from a comparison between calculated gaps based on nominal dimensions and the visual image based measured gaps. These reactor specific (typically 95% confidence level) CT/LIN gap models account for all uncertainties and deviations from nominal values. Prediction error margins reduce as more in-reactor gap measurements become available. Each year more measurements are being made using this standardized visual CT/LIN proximity method. The subsequently prepared reactor-specific models have been used to provide time to contact for every channel above the LINs at these stations. In a number of cases it has been used to demonstrate that the reactor can be operated to its end of life before refurbishment with no predicted contact, or specific at-risk channels have been identified for which appropriate remedial actions could be implemented in a planned manner. (author)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  12. 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)

  13. Bottom nozzle for nuclear reactor fuel assembly having an adaptor plate and a coupled filtration plate

    International Nuclear Information System (INIS)

    Verdier, M.; Mortgat, R.

    1992-01-01

    The bottom nozzle includes an adaptor plate with openings to allow the passage of water and a filtration plate with small holes. The openings in the adaptor plate are symmetrical with regard to medians and diagonals. Within each zone, some of the openings are rectangular and some may be circular. The small holes in the filtration plate coincide with the rectangular openings in the adaptor plate

  14. 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)

  15. 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)

  16. Reconstitutable nuclear reactor fuel assembly with unitary removable top nozzle subassembly

    International Nuclear Information System (INIS)

    Shallenberger, J.M.

    1987-01-01

    A reconstitutable fuel assembly is described having at least one control rod guide thimble and a top nozzle, the guide thimble including an upper extension, the top nozzle including at least one hold-down spring, an upper hold-down plate and a lower adapter plate, an improved attaching structure removably mounting the top nozzle as a unitary subassembly on the guide thimble. The attaching structure comprises: (a) a coupling member interfitting the lower adapter plate, the upper hold-down plate and the hold-down spring disposed between the plates so as to capture and retain the plates and spring together as a unitary subassembly in which the upper plate is slidably moveable along the coupling member relative to the lower plate with the spring biasing the upper plate away from the lower plate. The coupling member has spaced apart upper and lower portions with a central passageway extending for slidably receiving the upper extension of the guide thimble in a nonattached relationship in which the coupling member is slidably movable relative to the guide thimble extension for respectively inserting and removing the coupling member on and from the guide thimble extension

  17. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  18. A CFD model for the IEA-R1 reactor neat exchanger inlet nozzle flow

    International Nuclear Information System (INIS)

    Andrade, Delvonei A.; Angelo, Gabriel; Gainer, Gerson; Angelo, Edvaldo; Umbehaun, Pedro E.; Torres, Walmir M.; Sabundjian, Gaiane; Macedo, Luiz A.; Belchior Junior, Antonio; Conti, Thadeu N.; Watanabe, Bruno C.; Sakai, Caio C.

    2011-01-01

    A previous preliminary model of the IEA-R1 heat exchanger inlet nozzle flow was developed and published in the International Nuclear Atlantic Conference-INAC-2009. A new model was created based on the preliminary one. It was improved concerning the actual heat exchanger tube bundle geometry. This became a very special issue. Difficulties with the size of the numerical mesh came out pointing to our computational system limits. New CFD calculations with this improved model were performed using ANSYS-CFX. In this paper, we present this model and discuss the results. (author)

  19. A CFD model for the IEA-R1 reactor neat exchanger inlet nozzle flow

    Energy Technology Data Exchange (ETDEWEB)

    Andrade, Delvonei A.; Angelo, Gabriel; Gainer, Gerson; Angelo, Edvaldo; Umbehaun, Pedro E.; Torres, Walmir M.; Sabundjian, Gaiane; Macedo, Luiz A.; Belchior Junior, Antonio; Conti, Thadeu N.; Watanabe, Bruno C.; Sakai, Caio C., E-mail: delvonei@ipen.b, E-mail: gfainer@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    A previous preliminary model of the IEA-R1 heat exchanger inlet nozzle flow was developed and published in the International Nuclear Atlantic Conference-INAC-2009. A new model was created based on the preliminary one. It was improved concerning the actual heat exchanger tube bundle geometry. This became a very special issue. Difficulties with the size of the numerical mesh came out pointing to our computational system limits. New CFD calculations with this improved model were performed using ANSYS-CFX. In this paper, we present this model and discuss the results. (author)

  20. 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

  1. An interface redesign for the feed-water system of the advanced boiling water reactor in a nuclear power plant in Taiwan

    International Nuclear Information System (INIS)

    Hsieh Minchih; Chiu Mingchuan; Hwang Sheueling

    2014-01-01

    A well-designed human-computer interface for the visual display unit in the control room of a complex environment can enhance operator efficiency and, thus, environmental safety. In fact, a cognitive gap often exists between an interface designer and an interface user. Therefore, the issue of the cognitive gap of interface design needs more improvement and investigation. This is an empirical study that presents the application of an ecological interface design (EID) using three cases and demonstrates that an EID framework can support operators in various complex situations. Specifically, it analyzes different levels of automation and emergency condition response at the Lungmen Nuclear Power Plant in Taiwan. A simulated feed-water system was developed involving two interface styles. This study uses the NASA Task Load Index to objectively evaluate the mental workload of the human operators and the Situation Awareness Rating Technique to subjectively assess operator understanding and response, and is a pilot study investigating EID display format use at nuclear power plants in Taiwan. Results suggest the EID-based interface has a remarkable advantage over the original interface in supporting operator performance in the areas of response time and accuracy rate under both normal and emergency situations and provide supporting evidence that an EID-based interface can effectively enhance monitoring tasks in a complex environment. (author)

  2. 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%

  3. 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)

  4. Modeling of primary water stress corrosion cracking at control rod drive mechanism nozzles of pressurized water reactors

    International Nuclear Information System (INIS)

    Aly, Omar Fernandes

    2006-01-01

    One of the main failure mechanisms that cause risks to pressurized water reactors is the primary water stress corrosion cracking (PWSCC) occurring in alloys. It can occurs, besides another places, at the control reactor displacement mechanism nozzles. It is caused by the joint effect of tensile stress, temperature, susceptible metallurgical microstructure and environmental conditions of the primary water. These cracks can cause accidents that reduce nuclear safety by blocking the rod's displacement and may cause leakage of primary water, reducing the reactor's life. In this work it is proposed a study of the existing models and a modeling proposal to primary water stress corrosion cracking in these nozzles in a nickel based Alloy 600. It is been superposed electrochemical and fracture mechanics models, and validated using experimental and literature data. The experimental data were obtained at CDTN-Brazilian Nuclear Technology Development Center, in a recent installed slow strain rate testing equipment. In the literature it is found a diagram that indicates a thermodynamic condition for the occurrence of some PWSCC sub modes in Alloy 600: it was used potential x pH diagrams (Pourbaix diagrams), for Alloy 600 in high temperature primary water (300 deg C till 350 deg C). Over it, were located the PWSCC sub modes, using experimental data. It was added a third parameter called 'stress corrosion strength fraction'. However, it is possible to superpose to this diagram, other parameters expressing PWSCC initiation or growth kinetics from other models. Here is the proposition of the original contribution of this work: from an original experimental condition of potential versus pH, it was superposed, an empiric-comparative, a semi-empiric-probabilistic, an initiation time, and a strain rate damage models, to quantify respectively the PWSCC susceptibility, the failure time, and in the two lasts, the initiation time of stress corrosion cracking. It was modeling from our

  5. 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)

  6. Recommendations of the MRP-139: Inspection of Welds dissimilar in Nozzles PWR reactor vessel in Spain

    International Nuclear Information System (INIS)

    Gadea, J. R.; Willke, A.; Regidor, J. J.; Tecnatom, S. A.

    2010-01-01

    The guide EPRI MRP-139, which provides the way forward for the inspection and evaluation of dissimilar butt welds, the primary system of PWR reactors, indicating the type of nondestructive testing to be done in these areas, based on discovered several cases of default in lnconel alloys 600 and 182 in American and European plants. The phenomenon of cracking.

  7. Simulation of the fault transitory of the feedwater controller in a Boiling water reactor with the Ramona-3B code; Simulacion del transitorio de falla del controlador de agua de alimentacion en un reactor de agua en ebullicion con el codigo RAMONA-3B

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, J.L.; Ortiz V, J. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-07-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  9. 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.)

  10. 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

  11. 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

  12. 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

  13. Removable top nozzle and tool for a nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    Wilson, J.F.; Cerni, S.; Gjertsen, R.K.

    1987-01-01

    A tool is described used in a nuclear reactor fuel assembly for connecting and disconnecting an internally threaded collar threadably engagable with an externally threaded upper end of a control rod guide thimble. The tool consists of: (a) rotation means engagable with the collar and operable to apply a torque in one direction to threadably connect the internally threaded collar on the thimble's externally threaded upper end and in an opposite direction to disconnect the collar from its threaded connection with the thimble; and (b) gripper means adapted to be inserted into the upper end of the guide thimble and operable to prevent the thimble from rotating about its longitudinal axis as the rotation means applies torque to the collar in connecting and disconnecting the collar on and off the guide thimble; (c) the gripper means being disposed within the rotation means and having a portion thereof projecting outwardly beyond the rotation means for insertion into the thimble when the rotation means is engaged with the collar

  14. 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)

  15. 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

  16. 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

  17. 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)

  18. 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

  19. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  20. 49 CFR 230.57 - Injectors and feedwater pumps.

    Science.gov (United States)

    2010-10-01

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

  1. 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

  2. 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.

  3. 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

  4. 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.

  5. 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

  6. 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)

  7. Realistic thermal transient margin analysis of 'MONJU' based on plant performance measurements. Reactor vessel outlet nozzle and evaporator feed water inlet tube sheet of the manual reactor plant trip

    International Nuclear Information System (INIS)

    Yamada, Fumiaki; Mori, Takero

    2005-01-01

    In order to develop technologies and achieve safe and stable operation of Monju' as well as realize optimized design and construction of safe and economically competitive fast breeder reactors, the authors are evaluating design approach applied to 'Monju' based on actually measured behavioral data during plant operations. This report uses actual measured characteristic data of 'Monju' during a plant trip test obtained at a commissioning stage with up to 40% power output and introduces plant thermal hydraulic behavior analysis in a representative thermal transient event, i.e. a manual plant trip. Thermal transient driven loads incurred by the reactor vessel outlet nozzle and by the evaporator feed water inlet tube sheet were further derived by structural analyses and were compared with the previously derived values in the design stage and with the limit values. Though the reactor vessel outlet nozzle was exposed to larger temperature change in the trip test than the analytical prediction, the newly calculated mechanical load was about 50% of the previous evaluation in the design stage. Also, the newly analyzed mechanical load incurred by the evaporator feed water inlet tube sheet in this event had a large margin against the limit value of cumulative damage cycle fraction, although the observed temperature disturbance in a steam blow test was wilder than the analytical prediction. Thus we concluded that the Monju' plant has an assured safety margin against thermal transient in plant trip events. (author)

  8. 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.)

  9. 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.)

  10. 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)

  11. 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)

  12. Sodium cleaning device for nuclear reactor equipments

    International Nuclear Information System (INIS)

    Fujisawa, Morio.

    1985-01-01

    Purpose: To enable sodium cleaning over the entire length of large size equipments such as control rod drives and primary coolant recycling pumps for use in FBR type reactors. Constitution: A plurality of warm water supply nozzles each having a valve are connected at varying height on the side of a cleaning tank, to which an exhaust line is connected. These nozzles are connected with an exhaust port at the bottom of the tank to constitute a pipeway for cleaning warm water recycling line including a water feed pump and a feedwater heater. The water level in the tank is changed stepwise by successively selecting the warm water feed nozzles. Further, the warm water in the tank is recyclically fed through the nozzles selected at each step of the water level through the recycle line while warming. On the other hand, the pressure inside the tank is reduced through the exhaust line, whereby the warm water in the tank is boiled at low temperature to clean-up sodium on the equipments to be cleaned over the entire length. (Horiuchi, T.)

  13. 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

  14. Development of Preemptive Repair Technology for Alloy 600 J-Groove Welds of Reactor Vessel Upper Head CEDM Nozzles

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Kwang Woon; Lee, Jang Wook; Cho, Ki Hyun; Choi, Kwang Min; Choi, Dong Chul; Cho, Sang Beum; Cho, Hong Seok [KEPCO, Daejeon (Korea, Republic of)

    2016-05-15

    After 2000, PWSCC in numerous NPPs around the world has been generated, and recently, PWSCC in several CEDM nozzles of domestic NPP Hanbit Unit 3 and 4 was founded and repaired with embedded flaw repair(EFR) welding method by Westinghouse. In this study, development status of EFR equipment and basic experimental results for preventive PWSCC of RVUH CEDM nozzles will be introduced. The development of EFR seal welding equipment and welding process for the preemptive repair with original Alloy 600 J-Groove welds of RVUHP was conducted. The EFR welding equipment was tested to be possible seal welding to track J-Groove welds with three dimensional curved surfaces and OD penetration with vertical welding position. Through several BOP and overlay welding experiments, it was verified that good weld beads with no defects, such as cracks, spatter, undercut at the stable welding conditions with heat input of 27.4-32.5 KJ/in were well produced. Consequently, it is expected that the EFR seal welding technique will be applicable on the site.

  15. 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

  16. 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.)

  17. 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)

  18. Simulation of a passive auxiliary feedwater system with TRACE5

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-07-01

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

  19. 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

  20. 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

  1. Boundary element analysis of stress due to thermal shock loading or reactor pressure vessel nozzle; Napetostna analiza pri nestacionarni termicni obremenitvi cevnega prikljucka reaktorske tlacne posode z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Kramberger, J; Potrc, I [Tehniska fakulteta, Maribor (Yugoslavia)

    1989-07-01

    Apart from being exposed to the primary loading of internal pressure and steady temperature field, the reactor pressure vessel is also subject to various thermal transients (thermal shocks). Theoretical and experimental stress analyses show that severe material stresses occur in the nozzle area of the pressure vessel which may lead to defects (cracks). It has been our aim to evaluate these stresses by the use of the Boundary Element method. (author)

  2. On-line fatigue monitoring system for reactor pressure vessel

    International Nuclear Information System (INIS)

    Tokunaga, K.; Sakai, A.; Aoki, T.; Ranganath, S.; Stevens, G.L.

    1994-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to an operating boiling water reactor (BWR), Tsuruga Unit-1, is described. The system uses the influence function approach and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computed fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification No.501. Fatigue usage results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension. (author)

  3. 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.

  4. 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

  5. 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)

  6. 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

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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)

  12. 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)

  13. Evaluation of reflooding effects on an overheated boiling water reactor core in a small steam-line break accident using MAAP, MELCOR, and SCDAP/RELAP5 computer codes

    International Nuclear Information System (INIS)

    Lindholm, I.; Pekkarinen, E.; Sjoevall, H.

    1995-01-01

    Selected core reflooding situations were investigated in the case of a Finnish boiling water reactor with three severe accident analysis computer codes (MAAP, MELCOR, and SCDAP/RELAP5). The unmitigated base case accident scenario was a 10% steam-line break without water makeup to the reactor pressure vessel initially. The pumping of water to the core was started with the auxiliary feed water system when the maximum fuel cladding temperature reached 1,500 K. The auxiliary feedwater system pumps water (temperature 303 K) through the core spray spargers (core spray) on the top of the core and through feedwater nozzles to the downcomer (downcomer injection). The scope of the study was restricted to cases where the overheated core was still geometrically intact at the start of the reflooding. The following different core reflooding situations were investigated: (1) auxiliary feedwater injection to core spray (45 kg/s); (2) auxiliary feedwater injection to downcomer (45 kg/s); (3) auxiliary feedwater injection to downcomer (45 kg/s) and to core spray (45 kg/s); (4) no reflooding of the core. All the three codes predicted debris formation after the water addition, and in all MAAP and MELCOR reflooding results the core was quenched. The major difference between the code predictions was in the amount of H 2 produced, though the trends in H 2 production were similar. Additional steam production during the quenching process accelerated the oxidation in the unquenched parts of the core. This result is in accordance with several experimental observations

  14. 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

  15. 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

  16. Reactor core cooling device

    International Nuclear Information System (INIS)

    Kobayashi, Masahiro.

    1986-01-01

    Purpose: To safely and effectively cool down the reactor core after it has been shut down but is still hot due to after-heat. Constitution: Since the coolant extraction nozzle is situated at a location higher than the coolant injection nozzle, the coolant sprayed from the nozzle, is free from sucking immediately from the extraction nozzle and is therefore used effectively to cool the reactor core. As all the portions from the top to the bottom of the reactor are cooled simultaneously, the efficiency of the reactor cooling process is increased. Since the coolant extraction nozzle can be installed at a point considerably higher than the coolant injection nozzle, the distance from the coolant surface to the point of the coolant extraction nozzle can be made large, preventing cavitation near the coolant extraction nozzle. Therefore, without increasing the capacity of the heat exchanger, the reactor can be cooled down after a shutdown safely and efficiently. (Kawakami, Y.)

  17. 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

  18. 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

  19. Application of TRAC-BD1/MOD1 to a BWR/4 feedwater control failure ATWS

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Giles, M.M.; Mohr, C.M. Jr.; Weaver, W.L. III.

    1984-01-01

    This paper begins with a short description of the Transient Reactor Analysis Code for Boiling Water Reactors (TRAC-BWR), briefly mentioning some of its main features such as specific BWR models and input structure. Next, an input model of a BWR/4 is described, and, the assumptions used in performing an analysis of the loss of a feedwater controller without scram are listed. The important features of the calculated trends in flows, pressure, reactivity, and power are shown graphically and commented in the text. A comparison of some of the main predicted trends with the calculated results from a similar study by General Electric is also presented

  20. 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)

  1. Level controlling system in BWR type reactors

    International Nuclear Information System (INIS)

    Joge, Toshio; Higashigawa, Yuichi; Oomori, Takashi.

    1981-01-01

    Purpose: To reasonably attain fully automatic water level control in the core of BWR type nuclear power plants. Constitution: A feedwater flow regulation valve for reactor operation and a feedwater flow regulation valve for starting are provided at the outlet of a motor-driven feedwater pump in a feedwater system, and these valves are controlled by a feedwater flow rate controller. While on the other hand, a damp valve for reactor clean up system is controlled either in ''computer'' mode or in ''manual'' mode selected by a master switch, that is, controlled from a computer or the ON-OFF switch of the master switch by way of a valve control analog memory and a turn-over switch. In this way, the water level in the nuclear reactor can be controlled in a fully automatic manner reasonably at the starting up and shutdown of the plant to thereby provide man power saving. (Seki, T.)

  2. 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.

  3. Iron concentration controller in feedwater in nuclear plant

    International Nuclear Information System (INIS)

    Aizawa, Motohiro; Isaka, Yoshitaka

    1990-01-01

    The purpose of the present invention is to prevent chlorine ions from flowing into a reactor when sea water leakage accident should occur in a condenser upon control of Fe concentration in feedwater. That is, a sensor is disposed for detecting the leakage of the sea water at the exit of the condenser. The controller receives a detection signal as the input and delivers a control signal as the output. A control system receives the control signal and actuates valves in bypass systems. In view of the above, the electroconductivity or chlorine ion concentration of the condensate, which varies upon occurrence of sea water leakages in the condenser, is detected by the sensor, and then the controller closes a valve dispposed in the bypass systems in a processing device for filtering and desalting the condensates. Accordingly, the chlorine ions mixed into the condensates are removed by a desalting device without flowing into the reactor. In view of the above, an effect capable of keeping integrity of the plant is obtainable. (I.S.)

  4. 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.

  5. Thermal and stability considerations for a supercritical water-cooled fast reactor during power-raising phase of plant startup

    International Nuclear Information System (INIS)

    Cai, Jiejin; Ishiwatari, Yuki; Oka, Yoshiaki; Ikejiri, Satoshi

    2009-01-01

    This paper describes thermal analyses and linear stability analyses of the Supercritical Water-cooled Fast Reactor with 'two-path' flow scheme during the power-raising phase of plant startup. For thermal consideration, the same criterion of the maximum cladding surface temperature (MCST) as applied to the normal operating condition is used. For thermal-hydraulic stability consideration, the decay ratio of 0.5 is applied, which is taken from BWRs. Firstly, we calculated the flow rate distribution among the parallel flow paths from the reactor vessel inlet nozzles to the mixing plenum below the core using a system analysis code. The parallel flow paths consist of the seed fuel assemblies cooled by downward flow, the blanket fuel assemblies cooled by downward flow and the downcomer. Then, the MCSTs are estimated for various reactor powers and feedwater flow rates with system analyses. The decay ratios are estimated with linear stability analyses. The available range of the reactor power and feedwater flow rate to satisfy the thermal and stability criteria is obtained. (author)

  6. 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.

  7. 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.

  8. 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.)

  9. Seismic qualification of PWR plant auxiliary feedwater systems

    International Nuclear Information System (INIS)

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

    1983-08-01

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

  10. Analysis of limit cycling on a boiler feedwater control system

    International Nuclear Information System (INIS)

    Thomas, P.J.; Harrison, T.A.; Hollywell, P.D.

    1986-01-01

    During operation of the UKAEA Prototype Fast Reactor, it was found that oscillations sometimes occurred in the boiler feedwater systems. These were normally of relatively low amplitude, but led to the adoption of low controller gains so that control was rather slack. While control performance proved generally adequate for steady running, the lack of tight control of steam drum levels sometimes led to difficulties during periods when plant conditions were undergoing major change. The paper discusses the methods used to gain a full understanding of the phenomena occurring, and describes how that knowledge is being used to improve the control system so as to eliminate the limit cycling modes and ensure good control of steam drum levels. A noteworthy feature of the study was the use of two independent representations of plant behaviour: (i) a frequency response model, FWRFREQ, and (ii) a time-domain simulation model, PFRTDM. The simplified analysis of FWRFREQ proved to be of enormous value in identifying modes of system behaviour; PFRTDM was used as a detailed check on the accuracy and validity of the results obtained. (author)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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)

  16. 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)

  17. 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)

  18. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  19. 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.

  20. 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.

  1. 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...

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

    International Nuclear Information System (INIS)

    Sanchez B, A.

    2003-01-01

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

  3. 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.

  4. 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

  5. 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...

  6. Implementation of automated, on-line fatigue monitoring in a boiling water reactor

    International Nuclear Information System (INIS)

    Sakai, Takeshi; Tokunaga, Katsumi; Stevens, G.L.; Ranganath, S.

    1993-01-01

    A workstation-based, on-line fatigue monitoring system for tracking fatigue usage applied to a Japanese operating boiling water reactor (BWR), Tsuruga Unit 1, is described. The system uses the influence function approach and rainflow cycle counting methodology, operates on a workstation computer, and determines component stresses using temperature, pressure, and flow rate data that are made available via signal taps from previously existing plant sensors. Using plant-unique influence functions developed specifically for the feedwater nozzle location, the system calculates stresses as a function of time and computes the fatigue usage. The analysis method used to compute fatigue usage complies with MITI Code Notification number-sign 501. Fatigue values are saved automatically on files at times defined by the user for use at a later time. Of particular note, this paper describes some of the details involved with implementing such a system from the utility perspective. Utility installation details, as well as why such a system was chosen for implementation are presented. Fatigue results for an entire fuel cycle are presented and compared to assumed design basis events to confirm that actual plant thermal duty is significantly less severe than originally estimated in the design basis stress report. Although the system is specifically set up to address fatigue duty for the feedwater nozzle location, a generic shell structure was implemented so that any other components could be added at a future time without software modifications. As a result, the system provides the technical basis to more accurately evaluate actual reactor conditions as well as the justification for plant life extension

  7. The effect of steam cycle conditions upon the economics and design of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Philpott, E.F.; Pounder, F.; Willby, C.R.

    1978-01-01

    The paper studies the effect of variation of steam and feedwater conditions upon the economics, design and layout of a sodium-cooled fast reactor. The parameters investigated are steam temperature and pressure, feedwater temperature, and boiler recirculation ratio. The paper also includes an assessment of the effects of associating the fast reactor with saturated steam cycle conditions. (author)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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.)

  14. Use of on-line fatigue monitoring of nuclear reactor components as a tool for plant life extension

    International Nuclear Information System (INIS)

    Stevens, G.L.; Ranganath, S.

    1991-01-01

    In this paper the application of an on-line fatigue monitoring system for tracking fatigue usage in operating power plants is described. The system, like several others which have been developed, uses the influence function approach, operates on a microcomputer, and determines component stresses using temperature, pressure, and flow rate data that are typically available from plant computers. Using plant-unique influence functions developed specifically for each component location, the system calculates stresses as a function of time and computes the fatigue usage. Stress values are calculated at time internals defined by the user and the fatigue values are saved on files for use at a later time. The application of the GE Fatigue Monitoring System (GEFMS) to calculate fatigue usage in the feedwater nozzle of a GE boiling Water Reactor is described in this paper

  15. 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

  16. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  17. Mathematical modeling of a fast-breeder-reactor generating unit

    International Nuclear Information System (INIS)

    Kim, V.E.; Golovach, E.A.; Senkin, V.I.

    1984-01-01

    Dynamics equations are given for a reactor, intermediate heat exchanger, steam generator, and turbogenerator. The dynamic characteristics of the generating unit are described when perturbations occur in grid frequency, turbine valves, and feedwater consumption

  18. 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

  19. 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.)

  20. 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.)

  1. 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

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

    International Nuclear Information System (INIS)

    Meier, J.K.

    1983-01-01

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

  3. 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.)

  4. Investigation of accident management procedures related to loss of feedwater and station blackout in PSB-VVER integral test facility

    Energy Technology Data Exchange (ETDEWEB)

    Bucalossi, A. [EC JRC, (JRC F.5) PO Box 2, 1755 ZG Petten (Netherlands); Del Nevo, A., E-mail: alessandro.delnevo@enea.it [ENEA, C.R. Brasimone, 40032 Camugnano (Italy); Moretti, F.; D' Auria, F. [GRNSPG, Universita di Pisa, via Diotisalvi 2, 56100 Pisa (Italy); Elkin, I.V.; Melikhov, O.I. [Electrogorsk Research and Engineering Centre, Electrogorsk, Moscow Region (Russian Federation)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Four integral test facility experiments related to VVER-1000 reactor. Black-Right-Pointing-Pointer TH response of the VVER-1000 primary system following total loss of feedwater and station blackout scenarios. Black-Right-Pointing-Pointer Accident management procedures in case of total loss of feedwater and station blackout. Black-Right-Pointing-Pointer Experimental data represent an improvement of existing database for TH code validation. - Abstract: VVER 1000 reactors have some unique and specific features (e.g. large primary and secondary side fluid inventory, horizontal steam generators, core design) that require dedicated experimental and analytical analyses in order to assess the performance of safety systems and the effectiveness of possible accident management strategies. The European Commission funded project 'TACIS 2.03/97', Part A, provided valuable experimental data from the large-scale (1:300) PSB-VVER test facility, investigating accident management procedures in VVER-1000 reactor. A test matrix was developed at University of Pisa (responsible of the project) with the objective of obtaining the experimental data not covered by the OECD VVER validation matrix and with main focus on accident management procedures. Scenarios related to total loss of feed water and station blackout are investigated by means of four experiments accounting for different countermeasures, based on secondary cooling strategies and primary feed and bleed procedures. The transients are analyzed thoroughly focusing on the identification of phenomena that will challenge the code models during the simulations.

  5. 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

  6. 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

  7. 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

  8. Analysis of loss of normal feedwater transient using RELAP5/MOD1/NSC; KNU1 plant simulation

    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)

  9. 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

  10. Duplex tab exhaust nozzle

    Science.gov (United States)

    Gutmark, Ephraim Jeff (Inventor); Martens, Steven (nmn) (Inventor)

    2012-01-01

    An exhaust nozzle includes a conical duct terminating in an annular outlet. A row of vortex generating duplex tabs are mounted in the outlet. The tabs have compound radial and circumferential aft inclination inside the outlet for generating streamwise vortices for attenuating exhaust noise while reducing performance loss.

  11. Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

    OpenAIRE

    Coules, Harry; Smith, David

    2018-01-01

    Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel's structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile...

  12. 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

  13. 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

  14. 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

  15. 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.

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  18. Fundamentals of boiling water reactor systems

    International Nuclear Information System (INIS)

    Mattern, J.

    1976-01-01

    The reactor assembly consists of the reactor vessel, its internal components of the core, shroud, steam separator, dryer assemblies, feedwater spargers, internal recirculation pumps and control rod drive housings. Connected to the steam lines are the pressure relief valves which protect the pressure boundary from damage due to overpressure. (orig./TK) [de

  19. 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

  20. Premixed direct injection nozzle

    Science.gov (United States)

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

    2011-02-15

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

  1. Limit loads in nozzles

    International Nuclear Information System (INIS)

    Zouain, N.

    1983-01-01

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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.)

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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)

  13. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  14. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  15. 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

  16. Reactor water quality degradation suppressing method upon reactor start up

    International Nuclear Information System (INIS)

    Maeda, Katsuharu.

    1993-01-01

    Preceding to reactor start-up, vacuum degree in a condenser is increased, and after the vacuum degree has been increased sufficiently, a desalting tower is inserted. Then, water feed to the reactor is started and the reactor is operated so that water is supplied gradually. Thus, dissolved oxygen in the feedwater and condensates is kept low and an entire organic carbon leaching rate from resins in the condensate desalting tower is reduced. Further, since feedwater is gradually supplied after the start-up, the entire organic carbon brought into the reactor is decomposed by heat and radiation and efficiently removed by a reactor coolant cleanup system. As a result, corrosion of stainless steel or the like is suppressed, as well as integrity of fuels can be maintained. Further, degradation of water quality can be suppressed effectively not by additionally putting the condensate desalting towers to in-service in accordance with the increase of the feedwater flow rate accompanying the power up but by previously putting the condensate desalting towers to in-service. (N.H.)

  17. 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)

  18. 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.

  19. 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.

  20. Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  1. Single-tube condensation experiment in Passive Auxiliary Feedwater System of APR1400+

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Chang Wook; No, Hee Cheon; Yun, Bong Yo; Jeon, Byong Guk [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2012-05-15

    Conventional Korean nuclear power plants, Advanced Power Reactors (APR), are characterized by an active cooling system. However, Active cooling system may not prevent significant damage without any AC power source available for its operation as vividly illustrated through the recent Fukushima incident. In the APR1400+ to be designed, an independent passive cooling system was added in order to overcome the aforementioned shortcomings. In the Passive Auxiliary Feedwater System (PAFS), gravity force and density difference between steam and water are used. The system comprises of 240 condensation tubes to efficiently remove decay heat. Before applying the PAFS to APR1400+, the system's safety and heat removal performance must be verified. The present study experimentally evaluates the heat removal performance of a single tube in the PAFS. The objectives of SCOP (Single-tube Condensation experiment facility of PAFS) are the evaluation of the heat removal performance in the tube of the PAFS and database construction under various tube designs and test conditions. Reaching these objectives, we developed advanced measurement techniques for the amount of moisture, heat flux, and water film thickness.

  2. Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

    1993-02-01

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

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

    International Nuclear Information System (INIS)

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

    1994-05-01

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

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

    International Nuclear Information System (INIS)

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

    1993-12-01

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

  5. Auxiliary feedwater system risk-based inspection guide for the Maine Yankee Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    1992-10-01

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

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

    International Nuclear Information System (INIS)

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

    1991-07-01

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

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

    International Nuclear Information System (INIS)

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

    1993-08-01

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

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

    International Nuclear Information System (INIS)

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

    1993-02-01

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

  9. Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    1991-09-01

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

  10. Operating data monitoring and fatigue evaluation systems and findings for boiling water reactors in Japan

    International Nuclear Information System (INIS)

    Maekawa, O.; Kanazawa, Y.; Takahashi, Y.; Tani, M.

    1995-01-01

    The reactor pressure vessel (RPV) is one of the most critical components of a boiling water reactor (BWR) when utilities think about plant life extension (PLEX). Design stress analysis sometimes reports very high fatigue usage factors for such portions of RPVs as stud bolt, feedwater nozzle and support skirt.In order to evaluate design margin and to eliminate excessive conservatism in this design analysis to pave the way for PLEX, Japanese BWR utilities jointly with BWR manufacturers in Japan established a programme (1) to acquire plant operational data on line for specific parameters used in stress analysis, (2) to evaluate margin in the design using measured plant data best estimate boundary conditions for stress analysis, and (3) to establish a simplified fatigue analysis method for BWR RPV.A plant data acquisition system, named OPEDAS, has been developed and installed in Tokyo Electric Power Company's 1100MWe BWR at its Kashiwazaki-Kariwa nuclear power plant. Best estimate stress analysis using measured in-plant data has been carried out and the results show considerable margin in fatigue usage factor over the design. A simplified fatigue analysis method using in-plant data has been developed with the Green's function, although some limitations have been identified for its use. ((orig.))

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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)

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

    International Nuclear Information System (INIS)

    Lee, Song Kyu; Kim, Eun Kee; Heo, Gyun Young; An, Sang Ha

    2009-01-01

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

  17. '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)

  18. 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

  19. 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

  20. Pressure tube reactors

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1981-01-01

    Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

  1. 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

  2. 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...

  3. 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.).

  4. 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.).

  5. Thermodynamic analysis of a supercritical water reactor

    International Nuclear Information System (INIS)

    Edwards, M.

    2007-01-01

    A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

  6. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  7. 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)

  8. 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)

  9. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

  10. Modeling and simulation of the feedwater system, associated controller and interface with the user for the SUN-RAH nucleo electric plants university student simulator; Modelado y simulacion del sistema de agua de alimentacion, controlador asociado e interfaz con el usuario para el simulador universitario de nucleoelectricas SUN-RAH

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez B, A. [Laboratorio de Analisis de Ingenieria de Reactores Nucleares, DEPFI Campus Morelos, en IMTA Jiutepec, Morelos (Mexico)]. e-mail: alitet@eresmas.com

    2003-07-01

    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)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  12. Cracking at nozzle corners in the nuclear pressure vessel industry

    International Nuclear Information System (INIS)

    Smith, C.W.

    1986-01-01

    Cracks in nozzle corners at the pressure boundary of nuclear reactors have been frequently observed in service. These cracks tend to form with radial orientations with respect to the nozzle central axis and are believed to be initiated by thermal shock. However, their growth is believed to be primarily due to a steady plus a fluctuating internal pressure. Due to the impracticality of fracture testing of full-scale models, the Oak Ridge National Laboratory instituted the use of an intermediate test vessel (ITV) for use in fracture testing which had the same wall thickness and nozzle size as the prototype but significantly reduced overall length and diameter. In order to determine whether or not these ITVs could provide realistic data for full-scale reactor vessels, laboratory models of full-scale boiling water reactors and ITVs were constructed and tested. After briefly reviewing the laboratory testing and correlating results with service experience, results obtained will be used to draw some general conclusions regarding the stable growth of nonplanar cracks with curved crack fronts which are the most common precursors to fracture of pressure vessel components near junctures. Use of linear elastic fracture mechanics is made in determining stress-intensity distribution along the crack fronts

  13. 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.

  14. 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

  15. Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

    International Nuclear Information System (INIS)

    Ha, Gayeon; Heo, Gyunyoung; Song, Seok Yoon

    2016-01-01

    Since a performance of Feedwater Heater (FWH) is directly related to the thermodynamic efficiency of Nuclear Power Plants (NPPs), performance degradation of FWH results in loss of thermal power and ultimately business benefit. Nevertheless, it is difficult to diagnose its degradation of performance during normal operation due to its minor changes in process parameters, for instance, pressure, temperature, and flowrate. In this paper, six degradation modes have been analyzed and the performance indices for FWH such as Terminal Temperature Difference (TTD) and Drain Cooling Approach (DCA) have been used to diagnose degradation modes. PEPSE (Performance Evaluation of Power System Efficiencies) simulation, which is a plant simulation software simulating plant static characteristic and building energy balance model, has been used to generate the data of performance indices of FWH and actual measurements of FWH from NPPs was used to validate the classification model. In this paper, six degradation modes have been analyzed and the performance indices for FWH have been used to diagnose what degradation mode occurs. The RSM was proposed as a trend identifier of variables. Using RSM, it is possible to obtain appropriate information of the variables in noise environment since noise can be compressed while the original information is being converted to a trend. The SVC has been performed to classify the degradation mode of FWH, and then actual measurements of FWH from NPPs was used to validate the classification model. Performance indices under various leakage conditions show different patterns. In further study, tube leakage simulations for the various cases will be needed

  16. Classification of Feedwater Heater Performance Degradation Using Residual Sign Matrix

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Gayeon; Heo, Gyunyoung [Kyung Hee University, Seoul (Korea, Republic of); Song, Seok Yoon [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Since a performance of Feedwater Heater (FWH) is directly related to the thermodynamic efficiency of Nuclear Power Plants (NPPs), performance degradation of FWH results in loss of thermal power and ultimately business benefit. Nevertheless, it is difficult to diagnose its degradation of performance during normal operation due to its minor changes in process parameters, for instance, pressure, temperature, and flowrate. In this paper, six degradation modes have been analyzed and the performance indices for FWH such as Terminal Temperature Difference (TTD) and Drain Cooling Approach (DCA) have been used to diagnose degradation modes. PEPSE (Performance Evaluation of Power System Efficiencies) simulation, which is a plant simulation software simulating plant static characteristic and building energy balance model, has been used to generate the data of performance indices of FWH and actual measurements of FWH from NPPs was used to validate the classification model. In this paper, six degradation modes have been analyzed and the performance indices for FWH have been used to diagnose what degradation mode occurs. The RSM was proposed as a trend identifier of variables. Using RSM, it is possible to obtain appropriate information of the variables in noise environment since noise can be compressed while the original information is being converted to a trend. The SVC has been performed to classify the degradation mode of FWH, and then actual measurements of FWH from NPPs was used to validate the classification model. Performance indices under various leakage conditions show different patterns. In further study, tube leakage simulations for the various cases will be needed.

  17. 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

  18. 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.

  19. Reactors

    DEFF Research Database (Denmark)

    Shah, Vivek; Vaz Salles, Marcos António

    2018-01-01

    The requirements for OLTP database systems are becoming ever more demanding. Domains such as finance and computer games increasingly mandate that developers be able to encode complex application logic and control transaction latencies in in-memory databases. At the same time, infrastructure...... engineers in these domains need to experiment with and deploy OLTP database architectures that ensure application scalability and maximize resource utilization in modern machines. In this paper, we propose a relational actor programming model for in-memory databases as a novel, holistic approach towards......-level function calls. In contrast to classic transactional models, however, reactors allow developers to take advantage of intra-transaction parallelism and state encapsulation in their applications to reduce latency and improve locality. Moreover, reactors enable a new degree of flexibility in database...

  20. 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)

  1. 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.

  2. 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.

  3. SINGLE PHASE ANALYTICAL MODELS FOR TERRY TURBINE NOZZLE

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua; Zhang, Hongbin; Zou, Ling; O' Brien, James

    2016-11-01

    All BWR RCIC (Reactor Core Isolation Cooling) systems and PWR AFW (Auxiliary Feed Water) systems use Terry turbine, which is composed of the wheel with turbine buckets and several groups of fixed nozzles and reversing chambers inside the turbine casing. The inlet steam is accelerated through the turbine nozzle and impacts on the wheel buckets, generating work to drive the RCIC pump. As part of the efforts to understand the unexpected “self-regulating” mode of the RCIC systems in Fukushima accidents and extend BWR RCIC and PWR AFW operational range and flexibility, mechanistic models for the Terry turbine, based on Sandia National Laboratories’ original work, has been developed and implemented in the RELAP-7 code to simulate the RCIC system. RELAP-7 is a new reactor system code currently under development with the funding support from U.S. Department of Energy. The RELAP-7 code is a fully implicit code and the preconditioned Jacobian-free Newton-Krylov (JFNK) method is used to solve the discretized nonlinear system. This paper presents a set of analytical models for simulating the flow through the Terry turbine nozzles when inlet fluid is pure steam. The implementation of the models into RELAP-7 will be briefly discussed. In the Sandia model, the turbine bucket inlet velocity is provided according to a reduced-order model, which was obtained from a large number of CFD simulations. In this work, we propose an alternative method, using an under-expanded jet model to obtain the velocity and thermodynamic conditions for the turbine bucket inlet. The models include both adiabatic expansion process inside the nozzle and free expansion process out of the nozzle to reach the ambient pressure. The combined models are able to predict the steam mass flow rate and supersonic velocity to the Terry turbine bucket entrance, which are the necessary input conditions for the Terry Turbine rotor model. The nozzle analytical models were validated with experimental data and

  4. Safety aspects of water chemistry in light water reactors

    International Nuclear Information System (INIS)

    1988-12-01

    The goals of the water chemistry control programmes are to maximize operational safety and the availability and operating life of primary system components, to maximize fuel integrity, and to control radiation buildup. To achieve these goals an effective corporate policy should be developed and implemented. Essential management responsibilities are: Recognizing of the long-term benefits of avoiding or minimizing: a) system corrosion; b) fuel failure; and c) radiation buildup. The following control or diagnostic parameters are suitable performance indicators: for PWR primary coolant circuits: pH of reactor water (by operating temperature); Concentration of chlorides in reactor water; Hydrogen (or oxygen) in reactor water. For PWR secondary coolant circuits: pH in feedwater; Cation productivity in steam generator blowdown; Iron concentration in feedwater; Oxygen concentration in condensate. And BWR coolant circuits: Conductivity of reactor water; Concentration of chlorides in reactor water; Iron concentration in feedwater; Copper concentration in feedwater. The present document represents a review of the developments in some Member States on how to implement a reasonable water chemistry programme and how to assess its effectiveness through numerical indicators. 12 figs, 20 tabs

  5. 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.

  6. 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...

  7. 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.

  8. Process for manufacturing separating nozzles

    International Nuclear Information System (INIS)

    Bier, W.; Linder, G.; Mayer, E.

    1979-01-01

    The final form of the basic body and the unit consisting of the nozzle and peeling orifice provides immovable fixing of these parts. Surfaces of various components can then be milled, using milling tools, in one operation. Assembly can be made automatic. (DG) [de

  9. Main feedwater valve diagnostics at Waterford 3 nuclear generating station

    International Nuclear Information System (INIS)

    Fitzgerald, W.V.

    1991-01-01

    Pneumatically-operated control valves are coming under increasing scrutiny in nuclear power plants because of their relatively high incident rate. The theory behind a device that could make performance evaluation of these valves simpler and more effective was first described at the original EPRI Power Plant Valve Symposium. The development of this Diagnostic System was completed in 1989, and it was recently used to troubleshoot two main feedwater valves at Louisiana Power and Light's Waterford 3 Power Station. During a cold snap last December, these valves failed to respond to the input signal and, as a result, the plant came off line. An incident report had to be filed, and the plant chose to contact the original equipment manufacturer (OEM) for assistance. This paper describes the original incident involving these valves and then gives a brief description of the diagnostic system and how it works. The balance of the paper then reviews how the OEM and plant personnel utilized the system to evaluate each component of the control valve assembly (I/P transducer, positioner, volume boosters, actuator, and valve body assembly). By simply stroking the valve and monitoring pneumatic signals and valve position, the problem was traced to a malfunctioning positioner and a volume booster that was leaking. The problems were corrected and new performance signatures run for the valves using the system to document their improved operation. This case study demonstrates how new Diagnostic Technology along with OEM involvement can effectively address problems with pneumatically-operated control valves so that root-cause solutions can be implemented

  10. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  11. 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

  12. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  13. Fastener locking device for attaching guide thimble to fuel assembly bottom nozzle

    International Nuclear Information System (INIS)

    Widener, W.H.

    1987-01-01

    This patent describes a nuclear reactor fuel assembly including an end nozzle and at least one longitudinally-extending guide thimble projecting away from the end nozzle. The end nozzle has at least one passageway defined therethrough and a ledge defined within the passageway so as to face away from the guide thimble and divide the passageway into a first portion extending from the ledge toward the guide thimble. A second portion extends from the ledge away from the guide thimble. The second passageway portion has a larger cross-sectional size than the first passageway portion, the end nozzle also having recess means defined thereon in the second portion of the passageway. The guide thimble has an end disposed adjacent to the first portion of the passageway with threaded means defined thereon and a fastener locking device

  14. 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.)

  15. 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.

  16. 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.

  17. 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)

  18. 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.

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

    International Nuclear Information System (INIS)

    Papp, L.

    1995-01-01

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

  20. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. Diagnosis of Feedwater Heater Performance Degradation using Fuzzy Approach

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonmin; Kang, Yeon Kwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Song, Seok Yoon [Korea Hydro and Nuclear Power, Daejeon (Korea, Republic of)

    2014-05-15

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

  2. Method of nuclear reactor control using a variable temperature load dependent set point

    International Nuclear Information System (INIS)

    Kelly, J.J.; Rambo, G.E.

    1982-01-01

    A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

  3. Detection of a regulating valve closure failure during review of recorded data after an automatic reactor shut down. Incident at the NPP Beznau-1, 27 April 1995

    International Nuclear Information System (INIS)

    Deutschmann, H.

    1996-01-01

    After recognizing a leak in the oil system of the running main feedwater pump 1 during rated power operation of the plant the operator changed feedwater supply manually to the stand-by pump 2. A short time later pump 2 was automatically tripped by the signal ''low oil pressure''. Immediate reduction of the reactor power by the operator was not successful because the scram signal ''low steam generator level and mismatch of steam/feedwater flow'' occurred and scram was actuated. In this plant a special operating feature, actuated by the scram signal, is implemented to reduce steam release to atmosphere in case of scram. The signal ''scram and average primary Temperature >287 deg. C opens the feedwater regulating valves, and later, if the average primary temperature decreases to <287 deg. C, they reclose by a redundant signal. In the experienced event, after the scram actuation, in the steam generator A a feedwater overfill occurred. The overfill protection tripped the operating feedwater pumps (main feedwater pump 3 and two auxiliary feedwater pumps). The large injection of water produced an overcooling of the primary with isolation of the volume control system outlet of the primary. The operator repaired the defective oil coolers of the feedwater pumps and restarted the plant. At that time, he had not recognized, that the plant response, which caused the steam generator overfill, was wrong. One day later, as all the recorded data were reviewed in more detail, it was found that the closure time of the feedwater regulating valve to steam generator A was much longer than designed (19 s instead 7 s). The operator requested an LCO for continued operation in spite of the fact, that the closure time was not fixed in the Technical specification. 3 figs

  4. 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.

  5. 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

  6. 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)

  7. 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)

  8. Stress analysis of LOFT containment vessel attachments for the mainsteam and feedwater piping support structures

    International Nuclear Information System (INIS)

    Finicle, D.P.

    1977-01-01

    The LOFT Containment Vessel attachments for the Mainsteam and Feedwater Piping Support Structures have been analyzed for operating and faulted loading conditions. This report contains the analysis of the connections to the containment vessel for the most current design and loading. Also contained in this report is the analysis of the piping supports

  9. 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.)

  10. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  11. 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)

  12. 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.

  13. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  14. Pressure tube reactor

    International Nuclear Information System (INIS)

    Matsumoto, Tomoyuki; Fujino, Michihira.

    1980-01-01

    Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

  15. 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.

  16. 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.

  17. Auxiliary feedwater system risk-based inspection guide for the Beaver Valley, Units 1 and 2 nuclear power plants

    International Nuclear Information System (INIS)

    Lloyd, R.C.; Vehec, T.A.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Rossbach, L.W.; Sena, P.P. III

    1993-02-01

    In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Beaver Valley Units 1 and 2 were selected as two of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at Beaver Valley Units 1 and 2

  18. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    International Nuclear Information System (INIS)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

    2012-01-01

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described

  19. Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

    Energy Technology Data Exchange (ETDEWEB)

    Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

  20. Auxiliary feedwater system risk-based inspection guide for the J.M. Farley Nuclear Power Plant

    International Nuclear Information System (INIS)

    Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G.

    1990-10-01

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

  1. Improved reactor cavity

    International Nuclear Information System (INIS)

    Katz, L.R.; Demarchais, W.E.

    1984-01-01

    A reactor pressure vessel disposed in a cavity has coolant inlet or outlet pipes extending through passages in the cavity walls and welded to pressure nozzles. The cavity wall has means for directing fluid away from a break at a weld away from the pressure vessel, and means for inhibiting flow of fluid toward the vessel. (author)

  2. Applicability of fan spray nozzles to stripping insoluble gases from viscous liquids

    International Nuclear Information System (INIS)

    Tseng, H.H.; Johnson, E.F.

    1983-08-01

    Fan spray nozzle stripping appears to be a practical technique for separating dilute volatile solutes from nonvolatile solvents. In particular this technique can be used to strip molecular tritium and tritium fluoride at extremely small concentration (in the parts per million range) from molten salts used as blanket materials in a fusion reactor. Under adjusted operating conditions of the fan spray as it leaves the nozzle, a high percentage of the theoretically maximum achievable stripping would take place from the expanding sheet of the fan spray as it leaves the nozzle and before it breaks up. Although the only available experimental data are for aqueous solutions, a new theoretical analysis of the fan spray sheet demonstrates the applicability of this technique to nonaqueous liquids. The equation derived from this analysis relates the theoretically achievable mass transfer efficiency to the properties of the liquid flowing through the fan spray nozzle and to the operating conditions of the nozzle. Any fluid with viscosity higher than or equal to that of water would be expected to follow this equation as long as a fan-shaped sheet is formed under the operating conditions of the nozzle

  3. 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

  4. RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

    International Nuclear Information System (INIS)

    Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

    1984-01-01

    Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

  5. Steam generation: fossil-fired systems: utility boilers; industrial boilers; boiler auxillaries; nuclear systems: boiling water; pressurized water; in-core fuel management; steam-cycle systems: condensate/feedwater; circulating water; water treatment

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    A survey of development in steam generation is presented. First, fossil-fired systems are described. Progress in the design of utility and industrial boilers as well as in boiler auxiliaries is traced. Improvements in coal pulverizers, burners that cut pollution and improve efficiency, fans, air heaters and economisers are noted. Nuclear systems are then described, including the BWR and PWR reactors, in-core fuel management techniques are described. Finally, steam-cycle systems for fossil-fired and nuclear power plants are reviewed. Condensate/feedwater systems, circulating water systems, cooling towers, and water treatment systems are discussed

  6. Through an Annular Turbine Nozzle

    Directory of Open Access Journals (Sweden)

    Rainer Kurz

    1995-01-01

    is located in the gas turbine. The experiments were performed using total pressure probes and wall static pressure taps. The pitch variation modifies the flow field both upstream and downstream of the nozzle, although the experiments show that the effect is localized to the immediate neighborhood of the involved blades. The effects on the wakes and on the inviscid flow are discussed separately. The mean velocities show a strong sensitivity to the changes of the pitch, which is due to a potential flow effect rather than a viscous effect.

  7. Axisymmetric nozzles with chamfered contraction

    Czech Academy of Sciences Publication Activity Database

    Tesař, Václav

    2017-01-01

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

  8. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  9. 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...

  10. 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...

  11. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  12. 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

  13. Cavitation phenomena in a fuel injection nozzle of a diesel engine by neutron radiography

    International Nuclear Information System (INIS)

    Takenaka, N.; Kawabata, Y.; Miyata, D.; Kawabata, Y.; Sim, C. M.; Lim, I. C.

    2005-01-01

    Visualization of cavitation phenomena in a Diesel engine fuel injection nozzle was carried out by using neutron radiography system in Research Reactor Institute in Kyoto University and HANARO in Korea Atomic Energy Research Institute. A neutron chopper was synchronized to the engine rotation for high shutter speed exposures. A multi exposure method was applied to obtain a clear image as an ensemble average of the synchronized images. Some images were successfully obtained and suggested new understanding of the cavitation phenomena in a Diesel engine fuel injection nozzle

  14. 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

  15. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  16. 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

  17. 77 FR 55877 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

    Science.gov (United States)

    2012-09-11

    ... backfitting as defined in 10 CFR 50.109 (the Backfit Rule) and is not otherwise inconsistent with the issue... licenses or combined licenses. This regulatory guide may be applied to applications for operating licenses... may access information related to this document, which the NRC possesses and are publicly available...

  18. A fuzzy logic controller for feedwater regulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Eryuerek, E.E.; Upadhyaya, B.R.; Alguindigue, I.E.

    1994-01-01

    Fuzzy control refers to the application of fuzzy logic theory to control systems. In this paper fuzzy controllers for steam generator water level control and pump speed control are presented, and their performance in the presence of perturbations is discussed. In order to test the robustness of the controllers, their performance is compared with the performance of model based adaptive controllers and traditional PID controllers. The control actions calculated by the fuzzy controllers is have the characteristic of quick and smooth control compared to the others

  19. 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

  20. Core access system for nuclear reactor

    International Nuclear Information System (INIS)

    Andrea, C.

    1977-01-01

    Disclosed is an improved nuclear reactor arrangement to facilitate both through-the-head instrumentation and insertion and removal of assemblies from the nuclear core. The arrangement is of the type including a reactor vessel head comprising a large rotatable cover having a plurality of circular openings therethrough, a plurality of upwardly extending nozzles mounted on the upper surface of a large cover, and a plurality of upwardly extending skirts mounted on a large cover about the periphery or boundary of the circular openings; a plurality of small plugs for each of the openings in the large cover, the plugs also having nozzles mounted on the upper surface thereof, and drive mechanisms mounted on top of some of the nozzles and having means extending therethrough into the reactor vessel, the drive mechanisms and nozzles extending above the elevation of the upwardly extending skirts

  1. 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

  2. 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

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. 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

  8. 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

  9. 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.

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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

  15. Development of technique to apply induction heating stress improvement to recirculation inlet nozzle

    International Nuclear Information System (INIS)

    Chiba, Kunihiko; Nihei, Kenichi; Ootaka, Minoru

    2009-01-01

    Stress corrosion cracking (SCC) have been found in the primary loop recirculation (PLR) systems of boiling water reactors (BWR). Residual stress in welding heat-affected zone is one of the factors of SCC, and the residual stress improvement is one of the most effective methods to prevent SCC. Induction heating stress improvement (IHSI) is one of the techniques to improve reduce residual stress. However, it is difficult to apply IHSI to the place such as the recirculation inlet nozzle where the flow stagnates. In this present study, the technique to apply IHSI to the recirculation inlet nozzle was developed using water jet which blowed into the crevice between the nozzle safe end and the thermal sleeve. (author)

  16. Study for discharge coefficient of flow nozzles. Prediction by using numerical simulation

    International Nuclear Information System (INIS)

    Ikeda, Hiroshi; Sakai, Norio; Yamamoto, Yasushi; Arai, Kenji; Matsumoto, Masaaki

    2008-01-01

    In nuclear plant, as water feeding into reactor have much effect on thermal power of plant, it is important to measure accurately the flow rate of water. Flow nozzle is on of typical differential pressure type flow meters and the discharge coefficient is used to calculate the flow rate. This coefficient is given by actual experiment and theory. We studied the theoretical assumption of the discharge coefficient curve using numerical simulation and evaluated the effect of flow nozzle configuration on the coefficient numerically and experimentally. As the result, numerical simulation can predict the discharge coefficient of theoretical curve within 0.3%. And we found that the throat length and throat tapping location of flow nozzle have much effect on the coefficient. (author)

  17. 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.)

  18. 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.).

  19. 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.).

  20. Comparison of ASME Code NB-3200 and NB-3600 results for fatigue analysis of B31.1 branch nozzles

    International Nuclear Information System (INIS)

    Nitzel, M.E.; Ware, A.G.; Morton, D.K.

    1996-01-01

    Fatigue analyses wre conducted on two reactor coolant system branch nozzles in an operating PWR designed to the B31.1 Code, for which no explicit fatigue analysis was required by the licensing basis. These analyses were performed as part of resolving issues connected with NRC's Fatigue Action Plan to determine if the cumulative usage factor (CUF) for these nozzles, using the 1992 ASME Code and representative PWR transients, were comparable to nozzles designed and analyzed to the ASME Code. Both NB-3200 and NB-3600 ASME Code methods were used. NB-3200 analyses included the development of finite element models for each nozzle. Although detailed thermal transients were not available for the plant analyzed, representative transients from similar PWRs were applied in each method. CUFs calculated using NB-3200 methods were significantly less than using NB-3600. The paper points out differences in analysis methods and highlights difficulties and unknowns in performing more detailed analyses to reduce conservative assumptions

  1. The Status and Inspection of Bottom Mounted Instrumentation Nozzle in Korea

    International Nuclear Information System (INIS)

    Doh, Euisoon; Kim, Yoonwon; Kim, Jaeyoon; Lee, Tacksu; Lee, Changhun

    2012-01-01

    The PWSCC Cracking of Alloy 600 material has been issued since CRDM Penetration cracking of Bugey in France in 1990's. And J-groove weld cracking of CRDM at Oconee and PCR Nozzle cracking at Wolf Creek in USA were raising concern of the integrity for Dissimilar Metal Weld of Alloy 600. BMI(Bottom Mounted Instrumentation) Nozzle cracks were found at Takahama unit 1 in Japan and South Texas Project unit 1 in USA in 2003. And recent cracks of Reactor Head Vent line at Yonggwang unit 3 in Korea are enough to cause worry about the integrity for BMI Nozzles in Korea. BMI inspections of Westinghouse type plant were performed by KPS for Kori unit 1 in 2006, Ulchin unit 2 in 2007, and Kori unit 3 in 2008. The first inspection of OCR-1000 plant was carried out on May 2011 at Yonggwang unit 3. KPS developed the inspection technique of OCR-1000 plant for End Effector Module and controller, a quarterly actual sized Bottom head Mock up, Inspection probes meeting the regulatory guide lines and typical configuration of OCR-1000 plant. Two specimens with actual PWSCC cracks were used to demonstrate the Inspection technique of Detection and Sizing. and the quarterly actual sized Bottom head Mock up was very meaningful to check the Interference during the inspection by narrow gap between newly developments led to a successful inspection of the BMI Inspection. And the inspection was concurrently performed with 10 year Reactor Vessel ICI without hurting any critical path of the outage. This BMI inspection is contributing to keep Operational Safety of plants by prevention of Leakage at BMI nozzle and weld. And performing 10 Year ISI for BMI nozzle is very effective to prevent BMI nozzle Break by detecting PWSCC Initiation per PFM Sensitivity study

  2. BWR type reactors

    International Nuclear Information System (INIS)

    Yano, Ryoichi; Sato, Takashi; Osaki, Masahiko; Hirayama, Fumio; Watabe, Atsushi.

    1980-01-01

    Purpose: To effectively eliminate radioactive substances released upon loss of coolant accidents in BWR type reactors. Constitution: A high pressure gas jetting device having a plurality of small aperture nozzles is provided above a spray nozzle, that is, at the top of a dry well. The jetting device is connected to a vacuum breaker provided in a pressure suppression chamber. Upon loss of coolant accident, coolants are sprayed from the spray nozzle and air or nitrogen is jetted from the gas jetting device as well. Then, the gases in the dry well are disturbed, whereby radioactive iodine at high concentration liable to be accumulated in the dry well is forced downwardly, dissolved in the spray water and eliminated. (Ikeda, J.)

  3. Prediction of the Inlet Nozzle Velocity Profiles for the CANDU-6 Moderator Analysis

    International Nuclear Information System (INIS)

    Yoon, Churl; Park, Joo Hwan

    2006-01-01

    For the moderator analysis of the CANDU reactors in Korea, predicting local moderator subcooling in the Calandria vessels is one of the main concerns for the estimation of heat sink capability of moderator under LOCA transients. The moderator circulation pattern is determined by the combined forces of the inlet jet momentum and the buoyancy flow. Even though the inlet boundary condition plays an important role in determining the moderator circulations, no measured data of detailed inlet velocity profiles is available. The purpose of this study is to produce the velocity profiles at the inlet nozzles by a CFD simulation. To produce the velocity vector fields at the inlet nozzle surfaces, the internal flows in the nozzle assembly were simulated by using a commercial CFD code, CFX-5.7. In the reference, the analytical capability of CFX-5.7 had been estimated by a validation of the CFD code against available experimental data for separate flow phenomena. Various turbulence models and grid spacing had been also tested. In the following section, the interface treatment between the computational domains would be explained. In section 3, the inlet nozzle flow through the CANDU moderator nozzle assembly was predicted by using the obtained technology of the CFD simulation

  4. An analytical evaluation for the pressure drop characteristics of bottom nozzle flow holes

    International Nuclear Information System (INIS)

    Yang, S. G.; Kim, H. J.; Lim, H. T.; Park, E. J.; Jeon, K. L.

    2002-01-01

    An analytical evaluation for the bottom nozzle flow holes was performed to find a best design concept in terms of pressure drop. For this analysis, Computational Fluid Dynamics (CFD), FLUENT 5.5, code was selected as an analytical evaluation tool. The applicability of CFD code was verified by benchmarking study with Vibration Investigation of Small-scale Test Assemblies (VISTA) test data in several flow conditions and typical flow hole shape. From this verification, the analytical data were benchmarked roughly within 17% to the VISTA test data. And, overall trend under various flow conditions looked very similar between both cases. Based on the evaluated results using CFD code, it is concluded that the deburring and multiple chamfer hole features at leading edge are the excellent design concept to decrease pressure drop across bottom nozzle plate. The deburring and multiple chamfer hole features at leading edge on the bottom nozzle plate have 12% and 17% pressure drop benefit against a single chamfer hole feature on the bottom nozzle plate, respectively. These design features are meaningful and applicable as a low pressure drop design concept of bottom nozzle for Pressurized Water Reactor (PWR) fuel assembly

  5. Top Nozzle Holddown Spring Optimization of KSNP Fuel Assembly

    International Nuclear Information System (INIS)

    Lee, Seong Ki; Park, Nam Kyu; Kim, Hyeong Koo; Lee, Joon Ro; Kim, Jae Won

    2002-01-01

    Nuclear fuel assembly for Korea Standard Nuclear Power (KSNP) Plant has 4 helical compression springs at the upper end of it. The springs, in conjunction with the fuel assembly weight, apply a holddown force against excess of buoyancy forces and the upward hydraulic forces due to the reactor coolant flow. Thus the holddown spring is to be designed such that the positive net downward force will be maintained for all normal and anticipated transient flow and temperature conditions in the nuclear reactor. With satisfying these in-reactor requirements of the fuel assembly holddown spring. Under the assumption that spring density is constant, the volume nozzle holddown spring. Under the assumption that spring density is constant, the volume minimization is executed by using the design variables, viz., wire diameter, mean coil diameter, minimization is executed by using the design variables, viz., wire diameter, mean coil diameter are within the compatible range of the fuel assembly structural components. Based on these conditions, the optimum design of the holddown spring is obtained considering the reactor operating condition and by using ANSYS code. The optimized spring has the properties that are a decreased volume and increased stiffness, compared with the existing one even if the absolute values are very similar each other. The holddown spring design features and the algorithm developed in this study could be directly applicable to the current commercial production. Therefore, it could be used to enhance the design efficiency and the functional performance of the spring, and to reduce a material cost a little

  6. Integrated Composite Rocket Nozzle Extension, Phase I

    Data.gov (United States)

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

  7. Aerospike Nozzle for Rotating Detonation Engine Application

    Data.gov (United States)

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

  8. RNL automated ultrasonic inspection of the PISC II PWR inlet nozzle (Plate 3)

    International Nuclear Information System (INIS)

    Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.G.

    1987-01-01

    In June 1984, Risley Nuclear Laboratories (RNL) performed an automated ultrasonic inspection of the Pressurized Water Reactor (PWR) inlet nozzle (plate 3) from the international Programme of Inspection of Steel Components (PISC II) round-robin inspection programme. High-sensitivity pulse-echo detection and predominantly time-of-flight diffraction sizing techniques were employed from the clad inner surface of the nozzle using digital data collection, analysis, and display facilities developed at RNL. RNL detected 30 out of 31 intended weld flaws, achieved one hundred per cent correct acceptance of all acceptable flaws and had a correct rejection frequency on all rejectable flaws of 0.86. The results confirm that well-conceived automated inspection procedures, similar to those used by RNL in this nozzle inspection, could form the basis of a PSI/ISI procedure for reactor pressure vessel nozzle regions. Analysis of the RNL results with regard to the influence of flaw characteristics on inspection performance lends strong support to the general conclusions drawn by the PISC Data Analysis Group. In particular, the most difficult flaws to accurately size were circular smooth and rough flaws. Examination of the RNL results on individual flaws reveals valuable information on the strengths and weaknesses of the adopted procedures and points towards procedural changes that would improve inspection performance. This report describes the procedures adopted by RNL, in the inspection, and reviews the results in the light of definitive flaw information. (author)

  9. Auxiliary feedwater system risk-based inspection guide for the Diablo Canyon Unit 1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Harrison, D.G.

    1990-08-01

    This document presents a compilation of auxiliary feedwater (AFW) system failure information which has been screened for risk significance in terms of failure frequency and degradation of system performance. It is a risk-prioritized listing of failure events and their causes that are significant enough to warrant consideration in inspection planning at Diablo Canyon. This information is presented to provide inspectors with increased resources for inspection planning at Diablo Canyon. The risk importance of various component failure modes was identified by analysis of the results of probabilistic risk assessments (PRAs) for many pressurized water reactors (PWRs). However, the component failure categories identified in PRAs are rather broad, because the failure data used in the PRAs is an aggregate of many individual failures having a variety of root causes. In order to help inspectors to focus on specific aspects of component operation, maintenance and design which might cause these failures, an extensive review of component failure information was performed to identify and rank the root causes of these component failures. Both Diablo Canyon and industry-wide failure information was analyzed. Failure causes were sorted on the basis of frequency of occurrence and seriousness of consequence, and categorized as common cause failures, human errors, design problems, or component failures. This information permits an inspector to concentrate on components important to the prevention of core damage. Other components which perform essential functions, but which are not included because of high reliability or redundancy, must also be addressed to ensure that degradation does not increase their failure probabilities, and hence their risk importances. 23 refs., 1 fig., 1 tab

  10. 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

  11. Fracture mechanics evaluation of LOFT lower plenum injection nozzle

    International Nuclear Information System (INIS)

    Nagata, P.K.; Reuter, W.G.

    1977-01-01

    An analysis to establish whether or not a leak-before-break concept would apply to the LOFT lower plenum injection nozzle is described. The analysis encompassed the structure from the inlet side of valve V-2170 to the lower plenum nozzle-to-reactor vessel weld on the left side of the emergency core cooling system (ECCS). The defect that was assumed to exist was of such a size that the probability of its being missed by the applicable inspection technique was near zero. The Inconel 600 nozzle forging with an initial assumed defect size of 0.64 cm (0.25 in.) deep would behave as follows: (1) the axially oriented defect would result in leak before rupture (the number of cycles to rupture was 11,000), (2) the circumferentially oriented defect would result in a rupture before leak. The number of cycles to failure would be in excess of 14,000. Based on the conservative assumption that the thermal stresses were membrane stresses as opposed to a bending stress, the following were found. For the Inconel 82 weld metal (thickness of 1.3 cm [0.53 in.]) and AISI 316 SST valve body, with an initial assumed defect of 0.25 cm (0.1 in.), the crack would grow through the thickness in a minimum of 3950 cycles and to a critical rupture crack length of 5.1 cm (2.0 in.) in an additional 80 cycles. The Inconel 82 weld metal at the shell body (thickness of 9.7 cm or 3.8 in.) with an assumed defect 1.3 cm (0.5 in.) deep would fail in 334 cycles. Calculations made assuming a linear stress gradient instead of the above-mentioned flat distribution through the wall indicated that the number of stress cycles increased to 2200

  12. 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.

  13. Online monitoring of steam/water chemistry of a fast breeder test reactor

    International Nuclear Information System (INIS)

    Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

    2005-01-01

    Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

  14. 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

  15. 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

  16. 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.

  17. Three dimensional stress analysis of nozzle-to-shell intersections by the finite element method and a auto-mesh generation program

    International Nuclear Information System (INIS)

    Fujihara, Hirohiko; Ueda, Masahiro

    1975-01-01

    In the design of chemical reactors or nuclear pressure vessels it is often important to evaluate the stress distribution in nozzle-to-shell intersections. The finite element method is a powerful tool for stress analysis, but it has a defects to require troublesome work in preparing input data. Specially, the mesh data of oblique nozzles and tangential nozzles, in which stress concentration is very high, are very difficult to be prepared. The authors made a mesh generation program which can be used to any nozzle-to-shell intersections, and combining this program with a three dimensional stress analysis program by the finite element method they made the stress analysis of nozzle-to-shell intersections under internal pressure. Consequently, stresses, strains and deformations of nozzles nonsymmetrical to spherical shells and nozzles tangential to cylindrical shells were made clear and it was shown that the curvature of the inner surface of the nozzle corner was a controlling factor in reducing stress concentration. (auth.)

  18. 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

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

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges

    1998-01-01

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

  20. 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

  1. 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

  2. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-01

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

  4. 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

  5. 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.

  6. Analysis of unprotected overcooling events in the Integral Fast Reactor

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1989-01-01

    Simple analytic models are developed for predicting the response of a metal fueled, liquid-metal cooled reactor to unprotected overcooling events in the balance of plant. All overcooling initiators are shown to fall into two categories. The first category contains these events for which there is no final equilibrium state of constant overcooling, as in the case for a large steam leak. These events are analyzed using a non-flow control mass approach. The second category contains those events which will eventually equilibrate, such as a loss of feedwater heaters. A steady flow control volume analysis shows that these latter events ultimately affect the plant through the feedwater inlet to the steam generator. The models developed for analyzing these two categories provide upper bounds for the reactor's passive response to overcooling accident initiators. Calculation of these bounds for a prototypic plant indicate that failure limits -- eutectic melting, sodium boiling, fuel pin failure -- are not exceeded in any overcooling event. 2 refs

  7. Dominant accident sequences in Oconee-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Dearing, J.F.; Henninger, R.J.; Nassersharif, B.

    1985-04-01

    A set of dominant accident sequences in the Oconee-1 pressurized water reactor was selected using probabilistic risk analysis methods. Because some accident scenarios were similar, a subset of four accident sequences was selected to be analyzed with the Transient Reactor Analysis Code (TRAC) to further our insights into similar types of accidents. The sequences selected were loss-of-feedwater, small-small break loss-of-coolant, loss-of-feedwater-initiated transient without scram, and interfacing systems loss-of-coolant accidents. The normal plant response and the impact of equipment availability and potential operator actions were also examined. Strategies were developed for operator actions not covered in existing emergency operator guidelines and were tested using TRAC simulations to evaluate their effectiveness in preventing core uncovery and maintaining core cooling

  8. Emergency cooling system for a gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Cook, R.K.; Burylo, P.S.

    1975-01-01

    The site of the gas-cooled reactor with direct-circuit gas turbine is preferably the sea coast. An emergency cooling system with safety valve and emergency feed-water addition is designed which affects at least a part of the reactor core coolant after leaving the core. The emergency cooling system includes a water emergency cooling circuit with heat exchanger for the core coolant. The safety valve releases water or steam from the emergency coolant circuit when a certain temperature is exceeded; this is, however, replaced by the emergency feed-water. If the gas turbine exhibits a high and low pressure turbine stage, which are flowed through by coolant one behind another, a part of the coolant can be removed in front of each part turbine by two valves and be added to the haet exchanger. (RW/LH) [de

  9. Report of the ad hoc committee 'medium-flux reactor' of the Federal Minister for Research and Technology. Pt. 1

    International Nuclear Information System (INIS)

    Concepts for a new jet nozzle reactor as a thermal neutron source for beam tube experiments in fundamental and applied research in West Germany (reactor and infrastructure, organisation criteria, siting). (HP) [de

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

    International Nuclear Information System (INIS)

    Basdekas, D.L.

    1988-09-01

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

  11. Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, Susan L.; Cinson, Anthony D.; MacFarlan, Paul J.; Hanson, Brady D.; Mathews, Royce

    2012-08-01

    The objective of this investigation was to evaluate the efficacy of ultrasonic testing (UT) for primary water leak path assessments of reactor pressure vessel (RPV) upper head penetrations. Operating reactors have experienced leakage when stress corrosion cracking of nickel-based alloy penetrations allowed primary water into the annulus of the interference fit between the penetration and the low-alloy steel RPV head. In this investigation, UT leak path data were acquired for an Alloy 600 control rod drive mechanism nozzle penetration, referred to as Nozzle 63, which was removed from the North Anna Unit 2 reactor when the RPV head was replaced in 2002. In-service inspection prior to the head replacement indicated that Nozzle 63 had a probable leakage path through the interference fit region. Nozzle 63 was examined using a phased-array UT probe with a 5.0-MHz, eight-element annular array. Immersion data were acquired from the nozzle inner diameter surface. The UT data were interpreted by comparing to responses measured on a mockup penetration with known features. Following acquisition of the UT data, Nozzle 63 was destructively examined to determine if the features identified in the UT examination, including leakage paths and crystalline boric acid deposits, could be visually confirmed. Additional measurements of boric acid deposit thickness and low-alloy steel wastage were made to assess how these factors affect the UT response. The implications of these findings for interpreting UT leak path data are described.

  12. Reconstitutable fuel assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Ferlan, S.J.; Kmonk, S.; Schallenberger, J.M.

    1982-01-01

    A reconstitutable fuel assembly for a nuclear reactor which includes a mechanical, rather than metallurgical, arrangement for connecting control rod guide thimbles to the top and bottom nozzles of a fuel assembly. Multiple sleeves enclosing control rod guide thimbles interconnect the top nozzle to the fuel assembly upper grid. Each sleeve is secured to the top nozzle by retaining rings disposed on opposite sides of the nozzle. Similar sleeves enclose the lower end of control rod guide thimbles and interconnect the bottom nozzle with the lowermost grid on the assembly. An end plug fitted in the bottom end of each sleeve extends through the bottom nozzle and is secured thereto by a retaining ring. Should it be necessary to remove a fuel rod from the assembly, the retaining rings in either the top or bottom nozzles may be removed to release the nozzle from the control rod guide thimbles and thus expose either the top or bottom ends of the fuel rods to fuel rod removing mechanisms

  13. Improving accuracy of ET measurement of LISS nozzle to calandria tube clearance

    International Nuclear Information System (INIS)

    Craig, S.T.; Krause, T.W.; Schankula, J.J.

    2006-01-01

    The AECL Fuel Channel Inspection System (AFCIS) has been used in an in-reactor field trial to successfully measure the clearance between Liquid Injection Shutdown System (LISS) nozzles and calandria tubes. Each measurement over the full length of a channel added only 15 minutes to the on-channel inspection time. No changes were required to the inspection heads. The only equipment changes made were the addition of a Remote Field Eddy Current (RFEC) module to the eddy current instrument, and minor wiring changes, at the instrument, to achieve a RFEC configuration. With the experience gained from the field trial, factors potentially limiting accuracy were identified. These, and other factors, were investigated and are discussed herein. The RFEC probe is delivered inside the pressure tube. Magnetic fields from the RFEC probe extend through the conducting walls of the pressure tube and calandria tube to interact with the LISS nozzle. Data acquired during the field trial showed the LISS nozzle signal is distinct and the signal-to-noise ratio is very favourable. Nevertheless, comparison of the RFEC measurements to a visual examination, made during the same outage, had the RFEC method underestimating the clearance by 2.5 mm on average. By way of laboratory tests, the following factors were investigated as potential sources of error: resistivity and geometry of LISS nozzle reference/calibration pieces, pressure-tube wall thickness, diameter and resistivity variations, pressure-tube to calandria-tube gap, and radial offsets of the probe within the pressure-tube. The sensitivity to these various noise sources was established. A model, based on fundamental electromagnetic principles, was developed and was used to normalize the effects of LISS nozzle conductivity and geometry. This enabled compensation for various sources of error, and made it possible to produce a correction factor for the field trial data, reducing the average difference from the visual inspection of LISS

  14. 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,

  15. Mixing-Structure Relationship in Jet-Stirred Reactors

    KAUST Repository

    Ayass, Wassim W.

    2016-05-26

    In this study, measurements were performed to assess the overall mixing in jet-stirred reactors (JSRs) passively agitated by feed nozzles. The reactor diameter, nozzle shape, and nozzle diameter were varied to determine the effects of these geometrical parameters on mixing. The mixing was studied at ambient conditions using laser absorption spectroscopy to follow the exit concentration of a tracer gas, carbon dioxide, after a step change in its input flow. The results indicate that the use of a JSR of diameter D = 40 mm, having inclined or crossed nozzles of diameter d = 1 mm is recommended for low residence times up to 0.4 sec, while at moderate/high residence times 0.5-5 sec the use of a JSR of D = 56 mm and d = 0.3 mm having crossed nozzles is suggested.

  16. 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

  17. Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-07-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  18. Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

    International Nuclear Information System (INIS)

    Moreno Chamorro, P.; Gallego Diaz, C.

    2011-01-01

    The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

  19. 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

  20. 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.

  1. 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)

  2. 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)

  3. 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)

  4. 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.

  5. 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.

  6. Reducing scram frequency by modifying/eliminating steam generator low-low level reactor trip setpoint for Maanshan nuclear power plant

    International Nuclear Information System (INIS)

    Yuann, R.Y.; Chiang, S.C.; Hsiue, J.K.; Chen, P.C.

    1987-01-01

    The feasibility of modification/elimination of steam generator low-low level reactor trip setpoint is evaluated by using RETRAN-02 code for the purpose of reducing scram frequency in Maanshan 3-loop pressurized water reactor. The ANS Condition II event loss of normal feedwater and condition IV event feedwater system line break are the basis for steam generator low-low level reactor trip setpoint sensitivity analysis, including various initial reactor power levels, reactivity feedback coefficients, and system functions assumptions etc., have been performed for the two basis events with steam generator low-low level reactor trip setpoint at 0% narrow range and without this trip respectively. The feasibility of modifying/eliminating current steam generator low-low level reactor trip setpoint is then determined based on whether the analysis results meet with the ANS Condition II and IV acceptance criteria or not

  7. 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.

  8. BWR fuel assembly bottom nozzle with one-way coolant flow valve

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

  9. PWR type reactor plant

    International Nuclear Information System (INIS)

    Matsuoka, Tsuyoshi.

    1993-01-01

    A water chamber of a horizontal U-shaped pipe type steam generator is partitioned to an upper high temperature water chamber portion and a lower low temperature water chamber portion. An exit nozzle of a reactor container containing a reactor core therein is connected to a suction port of a coolant pump by way of first high temperature pipelines. The exit port of the coolant pump is connected to the high temperature water chamber portion of the steam generator by way of second high temperature pipelines. The low temperature water chamber portion of the steam generator is connected to an inlet nozzle of the reactor container by way of the low temperature pipelines. The low temperature water chamber portion of the steam generator is positioned lower than the high temperature water chamber portion, but upper than the reactor core. Accordingly, all of the steam generator for a primary coolant system, coolant pumps as well as high temperature pipelines and low temperature pipelines connecting them are disposed above the reactor core. With such a constitution, there is no worry of interrupting core cooling even upon occurrence of an accident, to improve plant safety. (I.N.)

  10. Reactor container spray device

    International Nuclear Information System (INIS)

    Yanai, Ryoichi.

    1980-01-01

    Purpose: To enable decrease in the heat and the concentration of radioactive iodine released from the reactor vessel into the reactor container in the spray device of BWR type reactors. Constitution: A plurality of water receiving trays are disposed below the spray nozzle in the dry well and communicated to a pressure suppression chamber by way of drain pipeways passing through a diaphragm floor. When the recycling system is ruptured and coolants in the reactor vessel and radioactive iodine in the reactor core are released into the dry well, spray water is discharged from the spray nozzle to eliminate the heat and the radioactive iodine in the dry well. In this case, the receiving trays collect the portions of spray water whose absorption power for the heat and radioactive iodine is nearly saturated and falls them into the pool water of the pressure suppression chamber. Consequently, other portions of the spray water that still possess absorption power can be jetted with no hindrance, to increase the efficiency for the removal of the heat and iodine of the spray droplets. (Horiuchi, T.)

  11. 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.

  12. 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.

  13. Supercritical Water Reactor Cycle for Medium Power Applications

    International Nuclear Information System (INIS)

    BD Middleton; J Buongiorno

    2007-01-01

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  14. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  15. 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.

  16. 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

  17. Shock wave fabricated ceramic-metal nozzles

    NARCIS (Netherlands)

    Carton, E.P.; Stuivinga, M.E.C.; Keizers, H.L.J.; Verbeek, H.J.; Put, P.J. van der

    1999-01-01

    Shock compaction was used in the fabrication of high temperature ceramic-based materials. The materials' development was geared towards the fabrication of nozzles for rocket engines using solid propellants, for which the following metal-ceramic (cermet) materials were fabricated and tested: B4C-Ti

  18. New atomization nozzle for spray drying

    NARCIS (Netherlands)

    Deventer, H.C. van; Houben, R.J.; Koldeweij, R.B.J.

    2013-01-01

    A new atomization nozzle based on ink jet technology is introduced for spray drying. Application areas are the food and dairy industry, in the first instance, because in these industries the quality demands on the final powders are high with respect to heat load, powder shape, and size distribution.

  19. Clamp and Gas Nozzle for TIG Welding

    Science.gov (United States)

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

    1982-01-01

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

  20. 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.

  1. 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.

  2. Design criteria for piping and nozzles program

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  3. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  5. Study of the reliability of the Auxiliary Feedwater System of a LWR nuclear power plant through the Fault Tree and Bayesian Network

    International Nuclear Information System (INIS)

    Lava, Deise Diana

    2016-01-01

    This paper aims to present a study of the reliability of the Auxiliary Feedwater System (AFWS) through the methods of Fault Tree and Bayesian Network. Therefore, the paper consists of a literature review of the history of nuclear energy and the methodologies used. The AFWS is responsible for providing water system to cool the secondary circuit of nuclear reactors of the PWR type when normal feeding water system failure. How this system operates only when the primary system fails, it is expected that the AFWS failure probability is very low. The AFWS failure probability is divided into two cases: the first is the probability of failure in the first eight hours of operation and the second is the probability of failure after eight hours of operation, considering that the system has not failed within the first eight hours. The calculation of the probability of failure of the second case was made through the use of Fault Tree and Bayesian Network, that it was constructed from the Fault Tree. The results of the failure probability obtained were very close, on the order of 10 -3 . (author)

  6. Oxygen suppression in boiling water reactors. Phase 2. Annual report 1981, December 2, 1980-December 31, 1981

    International Nuclear Information System (INIS)

    Burley, E.L.

    1982-07-01

    A hydrogen addition test will be performed in the Dresden-2 reactor of Commonwealth Edison Company during 1982. Up to 2 ppM hydrogen will be added to and dissolved in the reactor feedwater to reverse the radiolysis reaction in the reactor core and suppress oxgen concentration in the primary coolant. At low oxygen levels the propensity of stressed and sensitized 304 stainless steel toward intergranular stress corrosion cracking is greatly reduced. The test will answer outstanding questions and uncertainties in the areas of water chemistry, equipment design and materials performance. Nine special sample facilities will be prepared in the primary coolant, main stream, feedwater/condensate, and offgas systems. Instrumentation will be available to measure hydrogen, oxygen, conductivity, pH, soluble and insoluble corrosion products, and electrochemical potentials. In addition, an autoclave in which confirming constant extension rate tests can be conducted in reactor water will be provided

  7. Some aspects of primary and secondary water chemistry in CANDU reactors

    International Nuclear Information System (INIS)

    LeSurf, J.E.

    1978-09-01

    A brief review of the water chemistry in various circuits of CANDU reactors is given. Then, five particular aspects of recent work are highlighted: (i) Radiation Field Growth: in-reactor and out-reactor studies have related water chemistry to corrosion product deposition on fuel sheaths and subsequent contamination of out-core surfaces. (ii) Metal Oxide Solubility: novel techniques are being used to measure the solubilities of metal oxides at primary circuit conditions. (iii) Decontamination: the use of heavy water as coolant in CANDU reactors led to the development of a unique decontamination strategy and technique, called CAN-DECON, which has attracted the attention of operators of light-water reactors. (iv) Steam Generator Corrosion: mathematical modelling of the water chemistry in the bulk and crevice regions of nuclear steam generators, supported by chemical experiments, has shown why sea water ingress from leaking condensers can be damaging, and has provided a rapid way to evaluate alternative boiler water chemistries. (v) Automatic Control of Feedwater Chemistry: on-line automatic chemical analysis and computer control of feedwater chemistry provides All Volatile Treatment for normal operation with pure feedwater, and carefully controlled sodium phosphate addition when there is detectable sea-water ingress from leaking condensers. (author)

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

    International Nuclear Information System (INIS)

    Marquardt, K.; Dengler, H.

    1976-01-01

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

  9. 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

  10. 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

  11. Evaluation of total loss of feedwater accident/recovery phase and investigation of the associated EOP

    International Nuclear Information System (INIS)

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

    1993-01-01

    To evaluate the sequence of event and the thermohydraulic behavior during total loss of feedwater accident and recovery procedure, a RELAP5/MOD3 calculation is performed and compared with the LOFT L9-1/L3-3 experiment. Also, the predictability of the code for the major thermohydraulic phenomena following the accident is assessed. As a result, it is found that a pressure control using the spray until the time the water level reaches the top of the pressurizer, an overpressure protection by pressurizer PORV, a recovery of the secondary heat removal capability by refilling steam generator, and an effective cooldown by the continued natural circulation can be perfomed without core uncovery. It is also found that the plantspecific evaluation is necessary to confirm the effectiveness of the current symptom-oriented emergency operating procedure, especially in an overpressure protection performance and steam generator recovery performance. (Author)

  12. 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)

  13. FBR type reactor

    International Nuclear Information System (INIS)

    Nagai, Fumio.

    1979-01-01

    Purpose: To unify the temperature distribution in a nuclear reactor vessel by the provision of a gas recycle path for pressurizing a cover gas to recycle the cover gas and thus stir the gas in a cover gas chamber. Constitution: A plurality of gas inlet tubes and gas discharge tubes are provided to the wall of a cover gas chamber above the liquid level of coolants in a nuclear reactor vessel and the cover gas is recycled through the tubes. The plurality of gas inlet tubes are each provided at their tops with nozzles opening circumferentially and communicated to the outlet of a compressor. While on the other hand, the plurality of gas discharge tubes are communicated to the inlet of a compressor. Upon operation of the compressor, the pressurized cover gas is jetted out from the nozzles, swirls along the inner circumferential surface of the vessel and interrupts and stirs the vertical thermal convection. The gas, after swirling one-half of the inner circumferential surface of the vessel, automatically flows out of the gas discharging tubes opening behind the nozzles and then flows into the inlet of the compressor. (Seki, T.)

  14. Use of oxygen dosing to prevent flow accelerated corrosion in British Energy's Advanced Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Quirk, G.P.; Woolsey, I.S.; Rudge, A.

    2010-01-01

    Flow accelerated corrosion (FAC) was recognized as major threat to the carbon steel feed and economizer tubing of the once-through boilers of the UK's Advanced Gas-cooled Reactors (AGRs) following the observation of FAC damage of the boiler inlet orifice assemblies at two plants in 1977, and subsequent review of the likelihood of further damage elsewhere within the boilers of all AGRs. In most cases, replacement of susceptible tubing was not feasible; due to the inaccessibility of the boiler components within the reactor concrete pressure vessel. Preventing further FAC damage within the boilers therefore had to rely largely on changes to the boiler feedwater chemistry. Following extensive research programs carried out in the late 1970s and early 1980s two main feedwater chemistry regimes were adopted to suppress FAC in different AGRs. The four units found to be at greatest risk of FAC damage adopted an oxygen dosed All Volatile Treatment (AVT) regime during commissioning, while four other units retained the original deoxygenated ammonia dosed AVT regime, but with an increased feedwater pH. The deoxygenated ammonia dosed chemistry regime was also adopted in four AGR units subsequently built, which used 1%Cr0.5%Mo feed and economizer tubing in their once-through boilers. The oxygen dosed AVT chemistry regime adopted in four units having helical once-through boilers has proved highly effective in preventing FAC, with no evidence of damage after around 150,000 hours of operation. However, FAC damage was eventually found in some of the other units operating with a deoxygenated feedwater chemistry regime, in spite of having adopted an elevated feedwater pH. These units have now successfully converted to an oxygen dosed AVT feedwater chemistry regime to prevent further FAC damage, with the result that all 14 AGR reactors now operate with variants of the original oxygen dosed feedwater chemistry regime developed during the 1980s. The paper outlines the development of

  15. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  16. Data report of ROSA/LSTF experiment TR-LF-07. Loss-of-feedwater transient with primary feed-and-bleed operation

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-07-01

    An experiment TR-LF-07 was conducted on June 23, 1992 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-IV (ROSA-IV) Program. The ROSA/LSTF experiment TR-LF-07 simulated a loss-of-feedwater transient in a pressurized water reactor (PWR) under assumptions of primary feed-and-bleed operation and total failure of auxiliary feedwater system. A safety injection (SI) signal was generated when steam generator (SG) secondary-side collapsed liquid level decreased to 3 m. Primary depressurization was initiated by fully opening a power-operated relief valve (PORV) of pressurizer (PZR) 30 min after the SI signal. High pressure injection (HPI) system was started in loop with PZR 12 s after the SI signal, while it was initiated in loop without PZR when the primary pressure decreased to 10.7 MPa. The primary and SG secondary pressures were kept almost constant because of cycle opening of the PZR PORV and SG relief valves. The PZR liquid level began to drop steeply following the PORV full opening, which caused liquid level formation at the hot leg. The PZR and hot leg liquid levels recovered due to the HPI actuation in both loops. The primary pressure became lower than the SG secondary pressure, which resulted in the actuation of accumulator (ACC) system in both loops. The PZR and hot legs became full of liquid again after the ACC actuation. The primary feed-and-bleed operation by use of the PORV, HPI and ACC systems was effective to core cooling because of no core uncovery. The experiment was terminated when the continuous core cooling was confirmed due to the successive coolant injection from the HPI system even after the ACC termination. The obtained data would be useful to study operator actions and procedures in the PWR multiple fault events which behaviors in the PZR affect. This report summarizes the test procedures, conditions and major observation in the ROSA/LSTF experiment TR-LF-07. (author)

  17. 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

  18. Fracture analyses and test of regions with nozzle and hole and curvature influence in nuclear vessel

    International Nuclear Information System (INIS)

    Wang Baisong; Xu Dinggen; Ye Weijuan; Hu Yinbiao; Liang Xingyun; Gu Shaode; Zhou Peiying

    1993-08-01

    For the calculations of stress intensity factor K 1 of surface crack in the regions with nozzle and hole and the curvature influence on nuclear vessel, a improved 3-D collapsed isoparametric singular element with quarter-points was presented. The square root singularity in the vertical planes of crack was derived. The methods of transitional element and calculating K 1 from displacements were extensively used in 3- D case. The SIF K 1 of the corner crack in inner wall of the nozzle of RPV (reactor pressure vessel) for a typical 300 MW nuclear plant was calculated, and it was verified by 3-D photo-elastic test and diffusion of light test. The engineering fracture analysis and evaluation of the outside surface crack in the circular are transitional region of the head flange of RPV are also completed

  19. Testing and qualification of CIRCE venturi-nozzle flow meter for large scale experiments

    International Nuclear Information System (INIS)

    Ambrosini, W.; Forgione, N.; Oriolo, F.; Tarantino, M.; Agostini, P.; Benamati, G.; Bertacci, G.; Elmi, N.; Alemberti, A.; Cinotti, L.; Scaddozzo, G.

    2005-01-01

    This paper is focused on the tests carried out at the ENEA Brasimone Centre for the qualification of a large Venturi-Nozzle flow meter operating in Lead Bismuth Eutectic (LBE). Such flow meter has been selected to provide flow rate measurements during the thermal-hydraulic tests that will be performed on the experimental facility CIRCE. This large-scale facility is installed at the ENEA Brasimone Centre for studying the fluid-dynamics and operating behaviour of ADS reactor plants, as well as to qualify several components intended to be used in the LBE technology. The Venturi-Nozzle flow meter has been supplied by the Euromisure s.r.l., together with the calculated theoretical characteristic equation. The results obtained by the tests performed allowed to qualify this theoretical curve supplied by the manufacturer, that presents a very good agreement especially at high flow rate values. (authors)

  20. 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)

  1. 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%

  2. Fluiddynamic effects in the fuel element top nozzle area during refilling and reflooding

    International Nuclear Information System (INIS)

    Hawighorst, A.; Kroening, H.; Mewes, D.; Spatz, R.; Mayinger, F.

    1985-01-01

    During the refilling and reflooding phase following a hypothetical loss of coolant accident in lightwater cooled nuclear reactors, there will be countercurrent flow between discharging steam and the feed of emergency core cooling water. It was the objective of this research project to contribute to a better physical understanding of the fluiddynamic processes in the area of the fuel element top nozzle and so to improve emergency core cooling calculations. Therefore, experimental and theoretical investigations about the entrainment and countercurrent behaviour of gas/liquid flows have been implemented within this project. Fluiddynamic processes in the fuel element top nozzle area were simulated during the reflooding and refilling phase. Based on special internals as single and multiple-hole orifices, basic phenomena of fluidynamics were studied first with air-water. Subsequently, investigations of the system steam/water were conducted. The reactor geometry was approximated step by step, until a complete reactor fuel assembly top nozzle was constituted. The system pressure was 4.8 bars (abs), in accordance with the conditions in the reactor pressure vessel at the end of the blowdown phase. The water was initially fed in at saturation temperature, then, as a second step, fed in at subcooled condition relative to the steam temperature, in order to be able to study condensation effects as well. First, investigations on gas/liquid countercurrent flows in the fluid system air/water are presented. Then one studies countercurrent flow in the system steam/water, including the investigation of condensation effects. Finally, a detailed description of the research on droplet size determination is given

  3. Coolant cleanup system for a nuclear reactor

    International Nuclear Information System (INIS)

    Shiina, Atsushi; Usui, Naoshi; Yamamoto, Michiyoshi; Osumi, Katsumi.

    1983-01-01

    Purpose: To maintain the electric conductivity of reactor water lower and to minimize the heat loss in the cleanup system by providing a low temperature cleanup system and a high temperature cleanup system together. Constitution: A low temperature cleanup system using ion exchange resins as filter aids and a high temperature cleanup system using inorganic ion exchange materials as filter aids are provided in combination. A part of the reactor water in a reactor pressure vessel is passed through a conductivity meter, one portion of which flows into the high temperature cleanup system having no heat exchanger and filled with inorganic ion exchange materials by way of a first flow rate control valve and the other portion of which flows into the low temperature cleanup system having heat exchangers and filled with the ion exchange materials by way of a second control valve. The first control valve is adjusted so as to flow, for example, about more than 15% of the feedwater flow rate to the high temperature cleanup system and the second control valve is adjusted with its valve opening degree depending on the indication of the conductivity meter so as to flow about 2 - 7 % of the feedwater flow rate into the low temperature cleanup system, to thereby control the electric conductivity to between 0.055 - 0.3 μS/cm. (Moriyama, K.)

  4. 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.

  5. 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)

  6. Safety design/analysis and scenario for prevention of CDA with ECCS in lead-bismuth-cooled fast reactor

    International Nuclear Information System (INIS)

    Minoru, Takahashi; Vaclav, Dostal; Abu Khalid, Rivai; Novitrian; Yumi, Yamada

    2007-01-01

    Safety design has been developed to show safety feature of Pb-Bi-cooled direct contact boiling water small fast reactor (PBWFR). The core is designed to have negative void reactivity even if the entire core and upper plenum are voided by steam intrusion from above. In-vessel type control rod driving mechanisms are used to prevent control rods from accidental ejection due to high pressure in the reactor vessel. In cases of coolant leakage from reactor vessel and feed water pipes, Pb-Bi coolant level in the reactor vessel is kept at the required level for decay heat removal by means of closed type guard vessel. Dual pipes are adopted to avoid leak of water in the feedwater system. Pump trip in feedwater systems initiates loss of coolant flow (LOF) event, although there is no concern of loss of flow accident due to primary pump trip. Injection of high pressure water slows down the flow-coast-down of feedwater at the LOF event. It has been evaluated that the fuel temperature is kept lower than safety limits at the unprotected loss of flow and heat sink (ATWS). A scenario for prevention of the core disruptive accident (CDA) with the emergency core cooling system (ECCS) is examined. The reactor becomes super-critical when the reactor vessel is filled with water. It is necessary to use water with boric acid for the ECC system, and additional backup rods for sub-critical core in water injection. (authors)

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  8. Analysis of film cooling in rocket nozzles

    Science.gov (United States)

    Woodbury, Keith A.

    1992-01-01

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

  9. Nozzle flow calculation for real gases

    International Nuclear Information System (INIS)

    Bier, K.; Ehrler, F.; Hartz, U.; Kissau, G.

    1977-01-01

    The flow of CHF 2 Cl vapor (refrigerant R 22) through a Laval nozzle of annular geometry has been investigated in the region near the saturation line with stagnation pressures up to 85 per cent of the critical pressure. Static pressure profiles measured along the nozzle axis were found in good agreement with profiles calculated for one-dimensional isentropic flow of the real gas the thermal properties of which were derived from an equation of state proposed previously by Rombusch. Minor deviations between measured and calculated static pressure curves occur in the supersonic part of the mozzle, especially when supersaturated states of the vapour are passed. These deviations can be attributed to uncertainties in the calculation of the enthalpy and to a small influence of the static pressure probe. An additional investigation was concerned with an approximate calculation of the nozzle flow of real gases. In this approximation the well known relations of ideal gas dynamics are applied, the ratio of specific heats for the ideal gas being replaced, however, by a suitably adapted isentropic exponent, which can be determined e.g. from measured values of the Laval pressure or of the mass flow. For pressure ratios p/po between 1 and approximately 0.1, corresponding to Mach numbers up to approximately 2.2, all the interesting properties of the investigated flow of CHF 2 Cl vapour are approximated within a few per cent. (orig.) [de

  10. Coherent structures in a supersonic complex nozzle

    Science.gov (United States)

    Magstadt, Andrew; Berry, Matthew; Glauser, Mark

    2016-11-01

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

  11. Stress analysis of PCV nozzle junction

    International Nuclear Information System (INIS)

    Uchiyama, Shoichi; Oikawa, Tsuneo; Hoshino, Seizo

    1976-01-01

    Most of various pressure vessels comprise each one cylindrical shell and one or more nozzles. In this study, in order to analyze the stress in the structures of this type as minutely and exactly as possible, the program for stress analysis by the finite element method was made, which is required for the strength analysis for three-dimensional structures. Especially, the problem of the stress distribution around nozzle junctions was solved theoretically with the program. The program for the analysis developed in this study is provided with various functions, such as the input generator for cylindrical, conical and spherical shells, and plotter, and is very covenient. The accuracy of analysis is very good. The method of analysis and the calculation of the rigidity matrices for the deformation in plane and bending are explained. The result of the stress analysis around the nozzle junctions of a containment vessel with this program was in good agreement with experimental data and the result with SAP-4 code, therefore the propriety of the calculated result with this program was proved. Also calculations were carried out on three cases, namely a flat plate fixed at one end with distributed load, a cylinder fixed at one end with internal pressure, and an I-beam fixed at one end with concentrated load. The calculated results agreed well with theoretical solutions in all cases. (Kako, I.)

  12. Flow energy piezoelectric bimorph nozzle harvester

    Science.gov (United States)

    Sherrit, Stewart; Lee, Hyeong Jae; Walkemeyer, Phillip; Hasenoehrl, Jennifer; Hall, Jeffrey L.; Colonius, Tim; Tosi, Luis Phillipe; Arrazola, Alvaro; Kim, Namhyo; Sun, Kai; Corbett, Gary

    2014-04-01

    There is a need for a long-life power generation scheme that could be used downhole in an oil well to produce 1 Watt average power. There are a variety of existing or proposed energy harvesting schemes that could be used in this environment but each of these has its own limitations. The vibrating piezoelectric structure is in principle capable of operating for very long lifetimes (decades) thereby possibly overcoming a principle limitation of existing technology based on rotating turbo-machinery. In order to determine the feasibility of using piezoelectrics to produce suitable flow energy harvesting, we surveyed experimentally a variety of nozzle configurations that could be used to excite a vibrating piezoelectric structure in such a way as to enable conversion of flow energy into useful amounts of electrical power. These included reed structures, spring mass-structures, drag and lift bluff bodies and a variety of nozzles with varying flow profiles. Although not an exhaustive survey we identified a spline nozzle/piezoelectric bimorph system that experimentally produced up to 3.4 mW per bimorph. This paper will discuss these results and present our initial analyses of the device using dimensional analysis and constitutive electromechanical modeling. The analysis suggests that an order-of-magnitude improvement in power generation from the current design is possible.

  13. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  14. 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.

  15. VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

    International Nuclear Information System (INIS)

    Doshi, P.K.; DeMario, E.E.; Knott, R.P.

    1993-01-01

    Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

  16. 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)

  17. 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)

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

    International Nuclear Information System (INIS)

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

    1993-12-01

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

  19. 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.

  20. 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)

  1. 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)

  2. TMI-2 instrument nozzle examinations at Argonne National Laboratory, February 1991--June 1993

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-06-01

    The accident at the Three Mile Island Unit 2 (TMI-2) reactor in March 1979 resulted in the relocation of approximately 19,000 kg of molten core material to the lower head of the reactor vessel. This material caused extensive damage to the instrument guide tubes and nozzles and was suspected of having caused significant metallurgical changes in the condition of the lower head itself. These changes and their effect on the margin-to-failure of the lower head became the focal point of an investigation co-sponsored by the United States Nuclear Regulatory Commission (NRC) and the Organization for Economic Co-operation and Development (OECD). The TMI-2 Vessel Investigation Project (VIP) was formed to determine the metallurgical state of the vessel at the lower head and to assess the margin-to-failure of the vessel under the conditions existing during the accident. This report was prepared under the auspices of the OECD/NEA Three Mile Island Vessel Investigation Project. Under the auspices of the VIP, specimens of the reactor vessel were removed in February 1990 by MPR Associates, Inc. In addition to these specimens, fourteen instrument nozzle segments and two segments of instrument guide tubes were retrieved for metallurgical evaluation. The purpose of this evaluation was to provide additional information on the thermal conditions on the lower head that would influence the margin-to-failure, and to provide insight into the progression of the accident scenario, specifically the movement of the molten fuel across the lower head.

  3. Generic evaluation of feedwater transients and small break loss-of-coolant accidents in combustion engineering designed operating plants

    International Nuclear Information System (INIS)

    1980-01-01

    The purpose of this report is to summarize the results of a generic evaluation of feedwater transients, small break loss-of-coolant accidents (LOCAs), and other TMI-2-related events in the Combustion Engineering (CE)-designed operating plants and to establish or confirm the bases for their continued operation. The results of this evaluation are presented in this report in the form of a set of findings and recommendations in each of the principal review areas

  4. An evaluation of the Davis-Besse loss of feedwater event (June 1985) from an accident management perspective

    International Nuclear Information System (INIS)

    Di Salvo, R.; Leonard, M.T.; Wreathall, J.

    1986-01-01

    An accident management perspective is used to analyze events associated with a total loss-of-feedwater at the Davis-Besse nuclear power plant in June 1985. The relationships of accident management to the closely associated concepts of risk management and emergency management are delineated. The analysis shows that the principal contributors to the event's occurrence were shortcomings in risk management. Successful performance by the operators in accident management was principally responsible for terminating the event without consequence to public health

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  6. 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.

  7. 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.

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

    KAUST Repository

    Dehwah, Abdullah

    2015-03-01

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

  9. 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

  10. 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.

  11. 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.

  12. Fluidized-bed calciner with combustion nozzle and shroud

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  13. Atmospheric-pressure small-scale thermal-hydraulic experiment of a PIUS-type reactor

    International Nuclear Information System (INIS)

    Tasaka, Kanji; Tamaki, Masayoshi; Imai, Satoshi; Kohketsu, Hideto; Anoda, Yoshinari; Murata, Hideo; Kukita, Yutaka.

    1992-01-01

    An experimental small-scale low-pressure setup of a PIUS (Process Inherent Ultimate Safety)-type reactor was used for the examination of the stability during normal operation such as startup and load following operation and of the safety during accidents such as loss-of-feedwater and pump runaway. Automatic feedback pump control system based on differential pressure at lower honeycomb density lock was quite effective to maintain the stratified interface between primary and pool water in the honeycomb density lock during normal operation. The process inherent ultimate safety characteristics of the PIUS-type reactor was confirmed with pump-trip scram at the pump speed limit for the various simulated accidents such as a loss-of-feedwater and pump runaway. (author)

  14. 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.)

  15. 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.

  16. 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

  17. 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.

  18. 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)

  19. Analysis of plume backflow around a nozzle lip in a nuclear rocket

    International Nuclear Information System (INIS)

    Chung, C.H.; Kim, S.C.; Stubbs, R.M.; De Witt, K.J.

    1993-06-01

    The structure of the flow around a nuclear thermal rocket nozzle lip has been investigated using the direct simulation Monte Carlo method. Special attention has been paid to the behavior of a small amount of harmful particles that may be present in the rocket exhaust gas. The harmful fission product particles are modeled by four inert gases whose molecular weights are in a range of 4 131. Atomic hydrogen, which exists in the flow due to the extremely high nuclear fuel temperature in the reactor, is also included. It is shown that the plume backflow is primarily determined by the thin subsonic fluid layer adjacent to the surface of the nozzle lip, and that the inflow boundary in the plume region has negligible effect on the backflow. It is also shown that a relatively large amount of the lighter species is scattered into the backflow region while the amount of the heavier species becomes negligible in this region due to extreme separation between the species. Results indicate that the backscattered molecules are very energetic and are fast-moving along the surface in the backflow region near the nozzle lip. 22 refs

  20. A recycling molecular beam reactor

    International Nuclear Information System (INIS)

    Prada-Silva, G.; Haller, G.L.; Fenn, J.B.

    1974-01-01

    In a Recycling Molecular Beam Reactor, RMBR, a beam of reactant gas molecules is formed from a supersonic free jet. After collision with a target the molecules pass through the vacuum pumps and are returned to the nozzle source. Continuous recycling permits the integration of very small reaction probabilities into measurable conversions which can be analyzed by gas chromatography. Some preliminary experiments have been carried out on the isomerization of cyclopropane

  1. EDF field experience of 182 J-Groove welds on CRDMs and SG channel head nozzles

    International Nuclear Information System (INIS)

    Duisabeau, L.; Deforge, D.; Thebault, Y.; Stindel, M.; Lemaire, E.

    2011-01-01

    The Reactor Pressure Vessel Head (RPVH) replacement program, which began after a leak occurrence in a vessel head nozzle in Alloy 600 at Bugey Unit 3, was a unique opportunity to perform an extended inspection program on the welds from the decommissioned RPV heads. This paper presents the actual results of this program. More than 800 CRDM J groove welds from 18 decommissioned RPV heads were inspected by automatic dye penetrant testing. Detected indications were characterized by viewing tools specifically developed and in some specific cases, by destructive investigations in hot lab. Some welding defects were observed but no indication corresponding to stress corrosion cracking (SCC) was detected at the welds wet surface nor propagation from welding manufacturing defects, including the weld with the longest operating time on EDF power plants (170 000 h). Very few cases of SCC propagation from Alloy 600 to Alloy 182 are reported. One case of initiation at the weld root pass was observed. From design, the weld root pass (mechanically loaded) of CRDM (Control Rod Drive Mechanism) nozzles is not in contact with primary water and the cracking observed occurred after a through wall cracking of the Alloy 600 tube, enabling primary water to wet the root pass. Concerning the steam generator (SG) drain nozzle, the alloy 182 weld root is directly in contact with primary water. In June 2008, a primary water leakage was suspected on a steam generator bowl drain while conducting a bare metal visual examination during the plant's outage. Dye penetrant testing of the weld and metallographic replica were implemented during the 2008 and 2009 refuelling outages to confirm a leakage by SCC. Manufacturing reports analyses revealed that the drain nozzle weld was repaired and had not been stress relieved during manufacturing. EDF has decided to plug this nozzle and to enforce the maintenance policy for similar components with the same manufacturing specificity. Regarding national and

  2. 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

  3. Bottom nozzle of a LWR fuel assembly

    International Nuclear Information System (INIS)

    Leroux, J.C.

    1991-01-01

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

  4. Airfoil shape for a turbine nozzle

    Science.gov (United States)

    Burdgick, Steven Sebastian; Patik, Joseph Francis; Itzel, Gary Michael

    2002-01-01

    A first-stage nozzle vane includes an airfoil having a profile according to Table I. The annulus profile of the hot gas path is defined in conjunction with the airfoil profile and the profile of the inner and outer walls by the Cartesian coordinate values given in Tables I and II, respectively. The airfoil is a three-dimensional bowed design, both in the airfoil body and in the trailing edge. The airfoil is steam and air-cooled by flowing cooling mediums through cavities extending in the vane between inner and outer walls.

  5. Application of a power plant simplification methodology: The example of the condensate feedwater system

    International Nuclear Information System (INIS)

    Seong, P.H.; Manno, V.P.; Golay, M.W.

    1988-01-01

    A novel framework for the systematic simplification of power plant design is described with a focus on the application for the optimization of condensate feedwater system (CFWS) design. The evolution of design complexity of CFWS is reviewed with emphasis upon the underlying optimization process. A new evaluation methodology which includes explicit accounting of human as well as mechanical effects upon system availability is described. The unifying figure of merit for an operating system is taken to be net electricity production cost. The evaluation methodology is applied to the comparative analysis of three designs. In the illustrative examples, the results illustrate how inclusion in the evaluation of explicit availability related costs leads to optimal configurations. These are different from those of current system design practices in that thermodynamic efficiency and capital cost optimization are not overemphasized. Rather a more complete set of design-dependent variables is taken into account, and other important variables which remain neglected in current practices are identified. A critique of the new optimization approach and a discussion of future work areas including improved human performance modeling and different optimization constraints are provided. (orig.)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. Probabilistic common cause failure modeling for auxiliary feedwater system after the introduction of flood barriers

    International Nuclear Information System (INIS)

    Zheng, Xiaoyu; Yamaguchi, Akira; Takata, Takashi

    2013-01-01

    Causal inference is capable of assessing common cause failure (CCF) events from the viewpoint of causes' risk significance. Authors proposed the alpha decomposition method for probabilistic CCF analysis, in which the classical alpha factor model and causal inference are integrated to conduct a quantitative assessment of causes' CCF risk significance. The alpha decomposition method includes a hybrid Bayesian network for revealing the relationship between component failures and potential causes, and a regression model in which CCF parameters (global alpha factors) are expressed by explanatory variables (causes' occurrence frequencies) and parameters (decomposed alpha factors). This article applies this method and associated databases needed to predict CCF parameters of auxiliary feedwater (AFW) system when defense barriers against internal flood are introduced. There is scarce operation data for functionally modified safety systems and the utilization of generic CCF databases is of unknown uncertainty. The alpha decomposition method has the potential of analyzing the CCF risk of modified AFW system reasonably based on generic CCF databases. Moreover, the sources of uncertainty in parameter estimation can be studied. An example is presented to demonstrate the process of applying Bayesian inference in the alpha decomposition process. The results show that the system-specific posterior distributions for CCF parameters can be predicted. (author)

  8. Reliability analysis of 2 types of auxiliary feedwater system for PWR

    International Nuclear Information System (INIS)

    Ekariansyah, Andi Sofrany

    2002-01-01

    This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-01

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

  10. 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.

  11. Propagation of temperature disturbances in bounded flows downstream of a nozzle block

    International Nuclear Information System (INIS)

    Krebs, L.

    1979-12-01

    The early detection of cooling disturbances in a fuel element of a sodium cooled reactor is a must for safety reasons. One possibility of achieving this goal is by measuring and analyzing the coolant temperature at the fuel element outlet. Assessment of the potential of this method requires knowledge of the flow phenomena downstream of the fuel element. As a fluid dynamics model of a fuel element a nozzle block is used, the bores of which correspond to the subchannels between the fuel rods. The studies are conducted in water which has kinematic properties comparable to those of sodium. The velocity and temperature fields downstream of the nozzle block are examined for two REYNOLDS numbers. To simulate a disturbed cooling condition, water with a temperature higher by ΔT anti T = 10 K is injected through one subchannel of the nozzle block. At the same time, the volume injected is varied. The central channel and one side channel close to the wall are selected as injection sites. Statisticl analysis of the measured velocity and temperature signals covers the following parameters: Linear averages, intensities, probability densities, spectral power densities, autocorrelation functions, integral turbulence lengths, dissipation lengths, dissipation, skewness and flatness values. On the basis of FOURIER's differential equation of heat conduction a theoretical model is developed to describe both the average temperature field and the intensity field in the flow downstream of the nozzle block. Comparison of measurements and calculations furnishes good agreement and indicates that extrapolation of the model to sodium as a fluid is possible. Supplementary to the measurements and calculations details of the water test rig and the anemometer measuring system used for velocity and temperature measurements are shown in the Appendix. (orig.) 891 GL/orig. 892 KN [de

  12. Mixing In Jet-Stirred Reactors With Different Geometries

    KAUST Repository

    Ayass, Wassim W.

    2013-12-01

    This work offers a well-developed understanding of the mixing process inside Jet- Stirred Reactors (JSR’s) with different geometries. Due to the difficulty of manufacturing these JSR’s made in quartz, existing JSR configurations were assessed with certain modifications and optimal operating conditions were suggested for each reactor. The effect of changing the reactor volume, the nozzle diameter and shape on mixing were both studied. Two nozzle geometries were examined in this study, a crossed shape nozzle and an inclined shape nozzle. Overall, six reactor configurations were assessed by conducting tracer experiments - using the state-of-art technologies of high-speed cameras and laser absorption spectroscopy- and Computational Fluid Dynamics (CFD) simulations. The high-speed camera tracer experiment gives unique qualitative information – not present in the literature – about the actual flow field. On the other hand, when using the laser technique, a more quantitative analysis emerges with determining the experimental residence time distribution (RTD) curves of each reactor. Comparing these RTD curves with the ideal curve helped in eliminating two cases. Finally, the CFD simulations predict the RTD curves as well as the mixing levels of the JSR’s operated at different residence times. All of these performed studies suggested the use of an inclined nozzle configuration with a reactor diameter D of 40mm and a nozzle diameter d of 1mm as the optimal choice for low residence time operation. However, for higher residence times, the crossed configuration reactor with D=56mm and d=0.3mm gave a nearly perfect behavior.

  13. 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

  14. Passive safety system of a super fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sutanto, E-mail: sutanto@fuji.waseda.jp [Cooperative Major in Nuclear Energy, Waseda University, Tokyo (Japan); Polytechnic Institute of Nuclear Technology—National Nuclear Energy Agency, Yogyakarta (Indonesia); Oka, Yoshiaki [The University of Tokyo, Tokyo (Japan)

    2015-08-15

    Highlights: • Passive safety system of a Super FR is proposed. • Total loss of feedwater flow and large LOCA are analyzed. • The criteria of MCST and core pressure are satisfied. - Abstract: Passive safety systems of a Super Fast Reactor are studied. The passive safety systems consist of isolation condenser (IC), automatic depressurization system (ADS), core make-up tank (CMT), gravity driven cooling system (GDCS), and passive containment cooling system (PCCS). Two accidents of total loss of feedwater flow and 100% cold-leg break large LOCA are analyzed by using the passive systems and the criteria of maximum cladding surface temperature (MCST) and maximum core pressure are satisfied. The isolation condenser can be used for mitigation of the accident of total loss of feedwater flow at both supercritical and subcritical pressures. The ADS is used for depressurization leading to a loss of coolant during line switching to operation of the isolation condenser at subcritical pressure. Use of CMT during line switching recovers the lost coolant. In case of large LOCA, GDCS can be used for core reflooding. Coolant vaporization in the core released to containment through the break is condensed by passive containment cooling system. The condensate flows to the GDCS pool by gravity force. The maximum cladding surface temperature (MCST) of the accident satisfies the criterion.

  15. Analysis of film cooling in rocket nozzles

    Science.gov (United States)

    Woodbury, Keith A.

    1993-01-01

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

  16. Grit blasting nozzle fabricated from mild tool steel proves satisfactory

    Science.gov (United States)

    Mc Farland, J. E.; Turbitt, B.

    1966-01-01

    Dry blasting with glass beads through a nozzle assembly descales both the outside and inside surfaces of tubes of Inconel 718 used for the distribution of gaseous oxygen. The inside of the nozzle is coated with polyurethane and the deflector with a commercially available liquid urethane rubber.

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

    Indian Academy of Sciences (India)

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

  18. 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.

  19. 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.

  20. Combustor nozzle for a fuel-flexible combustion system

    Science.gov (United States)

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

    2011-03-22

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