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Sample records for generation pressurized water

  1. Steam generator for pressurized-water reactors

    International Nuclear Information System (INIS)

    Michel, E.

    1971-01-01

    In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

  2. 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

  3. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  4. Culinary and pressure irrigation water system hydroelectric generation

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, Cory [Water Works Engineers, Pleasant Grove City, UT (United States)

    2016-01-29

    Pleasant Grove City owns and operates a drinking water system that included pressure reducing stations (PRVs) in various locations and flow conditions. Several of these station are suitable for power generation. The City evaluated their system to identify opportunities for power generation that can be implemented based on the analysis of costs and prediction of power generation and associated revenue. The evaluation led to the selection of the Battle Creek site for development of a hydro-electric power generating system. The Battle Creek site includes a pipeline that carries spring water to storage tanks. The system utilizes a PRV to reduce pressure before the water is introduced into the tanks. The evaluation recommended that the PRV at this location be replaced with a turbine for the generation of electricity. The system will be connected to the utility power grid for use in the community. A pelton turbine was selected for the site, and a turbine building and piping system were constructed to complete a fully functional power generation system. It is anticipated that the system will generate approximately 440,000 kW-hr per year resulting in $40,000 of annual revenue.

  5. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  6. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  7. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  8. Water plasma generation under atmospheric pressure for HFC destruction

    International Nuclear Information System (INIS)

    Watanabe, Takayuki; Tsuru, Taira

    2008-01-01

    The purpose of this paper is to investigate the decomposition process of hydrofluoroethylene (HFC-134a) by water plasmas. The water plasma was generated by DC arc discharge with a cathode of hafnium embedded into a copper rod and a nozzle-type copper anode. The advantage of the water plasma torch is the generation of 100%-water plasma by DC discharge. The distinctive steam generation leads to the portable light-weight plasma generation system that does not require the gas supply unit, as well as the high energy efficiency owing to the nonnecessity of the additional water-cooling. HFC-134a was injected into the water plasma jet to decompose it in the reaction tube. Neutralization vessel was combined to the reaction tube to absorb F 2 and HF generated from the HFC-134a decomposition. The decomposition was performed with changing the feed rate of HFC-134a up to 185 mmol/min. The decomposition efficiency of 99.9% can be obtained up to 0.43 mmol/kJ of the ratio of HFC-134a feed rate to the arc power, hence the maximum feed rate was estimated to be 160 g/h at 1 kW of the arc power

  9. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-01

    The use of internally and externally cooled annular fuel rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and economic assessment. The investigation was conducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperature. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasibility issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density

  10. High Performance Fuel Desing for Next Generation Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mujid S. Kazimi; Pavel Hejzlar

    2006-01-31

    The use of internally and externally cooled annular fule rods for high power density Pressurized Water Reactors is assessed. The assessment included steady state and transient thermal conditions, neutronic and fuel management requirements, mechanical vibration issues, fuel performance issues, fuel fabrication methods and econmic assessment. The investigation was donducted by a team from MIT, Westinghouse, Gamma Engineering, Framatome ANP, and AECL. The analyses led to the conclusion that raising the power density by 50% may be possible with this advanced fuel. Even at the 150% power level, the fuel temperature would be a few hundred degrees lower than the current fuel temperatre. Significant economic and safety advantages can be obtained by using this fuel in new reactors. Switching to this type of fuel for existing reactors would yield safety advantages, but the economic return is dependent on the duration of plant shutdown to accommodate higher power production. The main feasiblity issue for the high power performance appears to be the potential for uneven splitting of heat flux between the inner and outer fuel surfaces due to premature closure of the outer fuel-cladding gap. This could be overcome by using a very narrow gap for the inner fuel surface and/or the spraying of a crushable zirconium oxide film at the fuel pellet outer surface. An alternative fuel manufacturing approach using vobropacking was also investigated but appears to yield lower than desirable fuel density.

  11. Study on low pressure evaporation of fresh water generation system model

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Han Shik; Wibowo, Supriyanto; Shin, Yong Han; Jeong, Hyo Min [Gyeongsang National University, Tongyeong (Korea, Republic of); Fajar, Berkah [University of Diponegoro, Semarang (Indonesia)

    2012-02-15

    A low pressure evaporation fresh water generation system is designed for converting brackish water or seawater into fresh water by distillation in low pressure and temperature. Distillation through evaporation of feed water and subsequent vapor condensation as evaporation produced fresh water were studied; tap water was employed as feed water. The system uses the ejector as a vacuum creator of the evaporator, which is one of the most important parts in the distillation process. Hence liquid can be evaporated at a lower temperature than at normal or atmospheric conditions. Various operating conditions, i.e. temperature of feed water and different orifice diameters, were applied in the experiment to investigate the characteristics of the system. It was found that these parameters have a significant effect on the performance of fresh water generation systems with low pressure evaporation.

  12. Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

    Science.gov (United States)

    McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

    1994-01-01

    The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

  13. Fuzzy logic control of steam generator water level in pressurized water reactors

    International Nuclear Information System (INIS)

    Kuan, C.C.; Lin, C.; Hsu, C.C.

    1992-01-01

    In this paper a fuzzy logic controller is applied to control the steam generator water level in a pressurized water reactor. The method does not require a detailed mathematical mode of the object to be controlled. The design is based on a set of linguistic rules that were adopted from the human operator's experience. After off-line fuzzy computation, the controller is a lookup table, and thus, real-time control is achieved. Shrink-and-swell phenomena are considered in the linguistic rules, and the simulation results show that their effect is dramatically reduced. The performance of the control system can also be improved by changing the input and output scaling factors, which is convenient for on-line tuning

  14. Remote-controlled television for locating leaking tubes in pressurized-water reactor steam generators

    International Nuclear Information System (INIS)

    Cormault, P.; Denis, J.

    1978-01-01

    The Scarabee system is designed for observation of the tubes in water boxes of pressurized-water reactor nuclear-power-station steam generators. It consists essentially of a camera and a projector used as a marker, both of which swivel freely. The whole unit is housed in a water-tight container which can easily be decontaminated. Remote control of camera and marker movement is carried out from a console. (author)

  15. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  16. Pump selection and application in a pressurized water reactor electric generating plant

    International Nuclear Information System (INIS)

    Kitch, D.M.

    1985-01-01

    Various pump applications utilized in a nuclear pressurized water reactor electric generating plant are described. Emphasis is on pumps installed in the auxiliary systems of the primary nuclear steam supply system. Hydraulic and mechanical details, the ASME Code (Nuclear Design), materials, mechanical seals, shaft design, seismic qualification, and testing are addressed

  17. Evaluation of anticipatory signal to steam generator pressure control program for 700 MWe Indian pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Pahari, S.; Hajela, S.; Rammohan, H. P.; Malhotra, P. K.; Ghadge, S. G.

    2012-01-01

    700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) is horizontal channel type reactor with partial boiling at channel outlet. Due to boiling, it has a large volume of vapor present in the primary loops. It has two primary loops connected with the help of pressurizer surge line. The pressurizer has a large capacity and is partly filled by liquid and partly by vapor. Large vapor volume improves compressibility of the system. During turbine trip or load rejection, pressure builds up in Steam Generator (SG). This leads to pressurization of Primary Heat Transport System (PHTS). To control pressurization of SG and PHTS, around 70% of the steam generated in SG is dumped into the condenser by opening Condenser Steam Dump Valves (CSDVs) and rest of the steam is released to the atmosphere by opening Atmospheric Steam Discharge Valves (ASDVs) immediately after sensing the event. This is accomplished by adding anticipatory signal to the output of SG pressure controller. Anticipatory signal is proportional to the thermal power of reactor and the proportionality constant is set so that SG pressure controller's output jacks up to ASDV opening range when operating at 100% FP. To simulate this behavior for 700 MWe IPHWR, Primary and secondary heat transport system is modeled. SG pressure control and other process control program have also been modeled to capture overall plant dynamics. Analysis has been carried out with 3-D neutron kinetics coupled thermal hydraulic computer code ATMIKA.T to evaluate the effect of the anticipatory signal on PHT pressure and over all plant dynamics during turbine trip in 700 MWe IPHWR. This paper brings out the results of the analysis with and without considering anticipatory signal in SG pressure control program during turbine trip. (authors)

  18. Pressure transients resulting from sodium-water reaction following a large leak in LMFBR steam generator

    International Nuclear Information System (INIS)

    Rajput, A.K.

    1984-01-01

    The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases

  19. Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

    International Nuclear Information System (INIS)

    Woo, H.H.; Lu, S.C.

    1981-01-01

    Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

  20. Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

    International Nuclear Information System (INIS)

    Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

    1976-01-01

    Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

  1. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1985-01-01

    An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

  2. Estabilishing requirements for the next generation of pressurized water reactors--reducing the uncertainty

    International Nuclear Information System (INIS)

    Chernock, W.P.; Corcoran, W.R.; Rasin, W.H.; Stahlkopf, K.E.

    1987-01-01

    The Electric Power Research Institute is managing a major effort to establish requirements for the next generation of U.S. light water reactors. This effort is the vital first step in preserving the viability of the nuclear option to contribute to meet U.S. national electric power capacity needs in the next century. Combustion Engineering, Inc. and Duke Power Company formed a team to participate in the EPRI program which is guided by a Utility Steering committee consisting of experienced utility technical executives. A major thrust of the program is to reduce the uncertainties which would be faced by the utility executives in choosing the nuclear option. The uncertainties to be reduced include those related to safety, economic, operational, and regulatory aspects of advanced light water reactors. This paper overviews the Requirements Document program as it relates to the U.S. Advanced Light Water Reactor (ALWR) effort in reducing these uncertainties and reports the status of efforts to establish requirements for the next generation of pressurized water reactors. It concentrates on progress made in reducing the uncertainties which would deter selection of the nuclear option for contributing to U.S. national electric power capacity needs in the next century and updates previous reports in the same area. (author)

  3. 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

  4. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  5. Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

    International Nuclear Information System (INIS)

    Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

    1978-07-01

    In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

  6. Generation of reactive species in atmospheric pressure dielectric barrier discharge with liquid water

    Science.gov (United States)

    Zelong, ZHANG; Jie, SHEN; Cheng, CHENG; Zimu, XU; Weidong, XIA

    2018-04-01

    Atmospheric pressure helium/water dielectric barrier discharge (DBD) plasma is used to investigate the generation of reactive species in a gas-liquid interface and in a liquid. The emission intensity of the reactive species is measured by optical emission spectroscopy (OES) with different discharge powers at the gas-liquid interface. Spectrophotometry is used to analyze the reactive species induced by the plasma in the liquid. The concentration of OH radicals reaches 2.2 μm after 3 min of discharge treatment. In addition, the concentration of primary long-lived reactive species such as H2O2, {{{{NO}}}3}- and O3 are measured based on plasma treatment time. After 5 min of discharge treatment, the concentration of H2O2, {{{{NO}}}3}-, and O3 increased from 0 mg · L-1 to 96 mg · L-1, 19.5 mg · L-1, and 3.5 mg · L-1, respectively. The water treated by plasma still contained a considerable concentration of reactive species after 6 h of storage. The results will contribute to optimizing the DBD plasma system for biological decontamination.

  7. Fresh Water Generation from Aquifer-Pressured Carbon Storage: Annual Report FY09

    Energy Technology Data Exchange (ETDEWEB)

    Wolery, T; Aines, R; Hao, Y; Bourcier, W; Wolfe, T; Haussman, C

    2009-11-25

    This project is establishing the potential for using brine pressurized by Carbon Capture and Storage (CCS) operations in saline formations as the feedstock for desalination and water treatment technologies including reverse osmosis (RO) and nanofiltration (NF). The aquifer pressure resulting from the energy required to inject the carbon dioxide provides all or part of the inlet pressure for the desalination system. Residual brine is reinjected into the formation at net volume reduction, such that the volume of fresh water extracted balances the volume of CO{sub 2} injected into the formation. This process provides additional CO{sub 2} storage capacity in the aquifer, reduces operational risks (cap-rock fracturing, contamination of neighboring fresh water aquifers, and seismicity) by relieving overpressure in the formation, and provides a source of low-cost fresh water to offset costs or operational water needs. This multi-faceted project combines elements of geochemistry, reservoir engineering, and water treatment engineering. The range of saline formation waters is being identified and analyzed. Computer modeling and laboratory-scale experimentation are being used to examine mineral scaling and osmotic pressure limitations. Computer modeling is being used to evaluate processes in the storage aquifer, including the evolution of the pressure field. Water treatment costs are being evaluated by comparing the necessary process facilities to those in common use for seawater RO. There are presently limited brine composition data available for actual CCS sites by the site operators including in the U.S. the seven regional Carbon Sequestration Partnerships (CSPs). To work around this, we are building a 'catalog' of compositions representative of 'produced' waters (waters produced in the course of seeking or producing oil and gas), to which we are adding data from actual CCS sites as they become available. Produced waters comprise the most common

  8. Denting of Inconel 600 steam generator tubes in pressurized water reactors

    International Nuclear Information System (INIS)

    Van Rooyen, D.; Weeks, J.R.

    1976-10-01

    Rapid, localized corrosion of carbon steel tube support plates (TSP) has led to cases of denting of steam generator tubes, due to the pressure of corrosion products formed in crevices between the tubes and TSP holes. The corrosion product is mainly magnetite (Fe 3 O 4 ), formed in ''run-away'' fashion as a result of local chemistry changes when an extended operation with phosphate (PO 4 ) treatment of the secondary coolant is followed by an all volatile treatment (AVT). The rate of the ''run-away'' magnetite formation, and therefore, the extent of damage will probably vary with the amounts of the harmful chemicals present and with temperature. Leaky condensers are felt to be responsible for the presence of Cl - ions, and for the observation that denting is more extensive in plants with salt water cooled condensers. It is possible that thermal cycles assist the denting process, both by mechanical and chemical ratchetting mechanisms. Recommendations are presented concerning the continued operation of plants with observed denting

  9. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  10. Initial pressure spike and its propagation phenomena in sodium-water reaction tests for MONJU steam generators

    International Nuclear Information System (INIS)

    Sato, M.; Hiroi, H.; Tanaka, N.; Hori, M.

    1977-01-01

    With the objective of demonstrating the safe design of steam generators for prototype LMFBR MONJU against the postulated large-leak accident, a number of large-leak sodium-water reaction tests have been conducted using the SWAT-1 and SWAT-3 rigs. Investigation of the potential effects of pressure load on the system is one of the major concerns in these tests. This paper reports the behavior of initial pressure spike in the reaction vessel, its propagation phenomena to the simulated secondary cooling system, and the comparisons with the computer code for one-dimensional pressure wave propagation problems. Both rigs used are the scaled-down models of the helically coiled steam generators of MONJU. The SWAT-1 rig is a simplified model and consists of a reaction vessel (1/8 scale of MONJU evaporator with 0.4 m dia. and 2.5 m height) and a pressure relief system i.e., a pressure relief line and a reaction products tank. On the other hand, the SWAT-3 rig is a 1/2.5 scale of MONJU SG system and consists of an evaporator (reaction vessel with 1.3 m dia. and 6.35 m height), a superheater, an intermediate heat exchanger (IHX), a piping system simulating the secondary cooling circuit and a pressure relief system. The both water injection systems consist of a water injection line with a rupture disk installed in front of injection hole and an electrically heated water tank. Choice of water injection rates in the scaled-down models is made based on the method of iso-velocity modeling. Test results indicated that the characteristics of the initial pressure spike are dominated by those of initial water injection which are controlled by the conditions of water heater and the size of water injection hole, etc

  11. A concept of passive safety pressurized water reactor system with inherent matching nature of core heat generation and heat removal

    International Nuclear Information System (INIS)

    Murao, Yoshio; Araya, Fumimasa; Iwamura, Takamichi; Okumura, Keisuke

    1995-01-01

    The reduction of manpower in operation and maintenance by simplification of the system are essential to improve the safety and the economy of future light water reactors. At the Japan Atomic Energy Research Institute (JAERI), a concept of a simplified passive safety reactor system JPSR was developed for this purpose and in the concept minimization of developing work and conservation of scale-up capability in design were considered. The inherent matching nature of core heat generation and heat removal rate is introduced by the core with high reactivity coefficient for moderator density and low reactivity coefficient for fuel temperature (Doppler effect) and once-through steam generators (SGs). This nature makes the nuclear steam supply system physically-slave for the steam and energy conversion system by controlling feed water mass flow rate. The nature can be obtained by eliminating chemical shim and adopting in-vessel control rod drive mechanism (CRDM) units and a low power density core. In order to simplify the system, a large pressurizer, canned pumps, passive residual heat removal systems with air coolers as a final heat sink and passive coolant injection system are adopted and the functions of volume and boron concentration control and seal water supply are eliminated from the chemical and volume control system (CVCS). The emergency diesel generators and auxiliary component cooling system of 'safety class' for transferring heat to sea water as a final heat sink in emergency are also eliminated. All of systems are built in the containment except for the air coolers of the passive residual heat removal system. The analysis of the system revealed that the primary coolant expansion in 100% load reduction in 60 s can be mitigated in the pressurizer without actuating the pressure relief valves and the pressure in 50% load change in 30 s does not exceed the maximum allowable pressure in accidental conditions in regardless of pressure regulation. (author)

  12. Pressurized-water reactors

    International Nuclear Information System (INIS)

    Bush, S.H.

    1983-03-01

    An overview of the pressurized-water reactor (PWR) pressure boundary problems is presented. Specifically exempted will be discussions of problems with pumps, valves and steam generators on the basis that they will be covered in other papers. Pressure boundary reliability is examined in the context of real or perceived problems occurring over the past 5 to 6 years since the last IAEA Reliability Symposium. Issues explicitly covered will include the status of the pressurized thermal-shock problem, reliability of inservice inspections with emphasis on examination of the region immediately under the reactor pressure vessel (RPV) cladding, history of piping failures with emphasis on failure modes and mechanisms. Since nondestructive examination is the topic of one session, discussion will be limited to results rather than techniques

  13. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  14. General empirical model for 60Co generation in pressurized water reactors with continuous refueling

    International Nuclear Information System (INIS)

    Urrutia, G.A.; Blesa, M.A.; Fernandez-Prini, R.; Maroto, A.J.G.

    1984-01-01

    A simplified model is presented that permits one to calculate the average activity on the fuel elements of a reactor that operates under continuous refueling, based on the assumption of crud interchange between fuel element surface and coolant in the form of particulate material only and using the crud specific activity as an empirical parameter determined in plant. The net activity flux from core to out-of-core components is then calculated in the form of parametric curves depending on crud specific activity and rate of particulate release from fuel surface. In pressure vessel reactors, contribution to out-ofcore radionuclide inventory arising in the release of activated materials from core components must be taken into account. The contribution from in situ activation of core components is calculated from the rates of release and the specific activities corresponding to the exposed surface of the component (calculated in a straightforward way on the basis of core geometry and neutron fluxes). The rates of release can be taken from the literature, or in the case of cobalt-rich alloys, can be calculated from experimentally determined cobalt contents of structural components and crud. For pressure vessel reactors operating under continuous refueling, activation of deposited crud and release of activated materials are compared; the latter, in certain cases, may represent a sizable (and even the largest) fraction of the total cobalt activity. It is proposed that the ratio of activities of 59 Fe to 54 Mn may be used as a diagnostic tool for in situ activation of structural materials; available data indicate ratios close to unity for pressure tube heavy water reactors (no in situ activation) and ratios around 4 to 10 for pressure vessel, heavy water reactors

  15. Denting of Inconel steam generator tubes in pressurized water reactors. Third informal report

    International Nuclear Information System (INIS)

    van Rooyen, D.; Weeks, J.R.

    1977-08-01

    The recent plant operating experience and laboratory test results on the phenomenon of denting in recirculating PWR steam generators is reviewed. Although denting was first reported only in plants that were converted from phosphate to AVT, it has now also been observed in plants still on phosphate, as well as in some that started on AVT. In some units, slightly abnormal eddy current signals have been observed at the top of the tube sheets. The degree of denting in operating steam generators may be related to the levels and duration of chloride inleakage. Chloride, however, is not the only active ingredient, and does not seem to give denting until local acid conditions arise; consequently, it may be necessary for soluble copper and/or nickel ions to be present to promote the denting reaction. Chloride concentrations in actively corroding crevices can increase by several orders of magnitude over the bulk coolant. It is thus difficult to develop a basis for Cl - specifications for secondary water. Maintaining Cl - low enough to prevent denting may be unmanageable without full flow condensate demineralization in coastal plants with copper alloy condensors and feedwater lines. Cathodic depolarization by oxidizing species are thought to promote the formation of acid chlorides in crevices and trigger the denting reactions; some ions may also catalyze the rapid formation of magnetite. These, and other mechanistic aspects of denting are discussed. The implications of the Inconel 600 tube defects at Ginna in non-dented areas, originating from the primary side, are also discussed

  16. Hybrid Pressure Retarded Osmosis−Membrane Distillation (PRO−MD) Process for Osmotic Power and Clean Water Generation

    KAUST Repository

    Han, Gang

    2015-05-20

    A novel pressure retarded osmosis−membrane distillation (PRO−MD) hybrid process has been experimentally conceived for sustainable production of renewable osmotic power and clean water from various waters. The proposed PRO−MD system may possess unique advantages of high water recovery rate, huge osmotic power generation, well controlled membrane fouling, and minimal environmental impacts. Experimental results show that the PRO−MD hybrid process is promising that not only can harvest osmotic energy from freshwater but also from wastewater. When employing a 2 M NaCl MD concentrate as the draw solution, ultrahigh power densities of 31.0 W/m2 and 9.3 W/m2 have been demonstrated by the PRO subsystem using deionized water and real wastewater brine as the feeds, respectively. Simultaneously, high purity potable water with a flux of 32.5−63.1 L/(m2.h) can be produced by the MD subsystem at 40−60 °C without any detrimental effects of fouling. The energy consumption in the MD subsystem might be further reduced by applying a heat exchanger in the hybrid system and using low-grade heat or solar energy to heat up the feed solution. The newly developed PRO−MD hybrid process would provide insightful guidelines for the exploration of alternative green technologies for renewable osmotic energy and clean water production.

  17. Experiments and modeling of discharge characteristics in water-mist sprays generated by pressure-swirl atomizers

    Science.gov (United States)

    Santangelo, Paolo E.

    2012-12-01

    Pressure-swirl atomizers are often employed to generate a water-mist spray, typically employed in fire suppression. In the present study, an experimental characterization of dispersion (velocity and cone angle) and atomization (drop-size axial evolution) was carried out following a previously developed methodology, with specific reference to the initial region of the spray. Laser-based techniques were used to quantitatively evaluate the considered phenomena: velocity field was reconstructed through a Particle Image Velocimetry analysis; drop-size distribution was measured by a Malvern Spraytec device, highlighting secondary atomization and subsequent coalescence along the spray axis. Moreover, a comprehensive set of relations was validated as predictive of the involved parameters, following an inviscid-fluid approach. The proposed model pertains to early studies on pressure-swirl atomizers and primarily yields to determine both initial velocity and cone angle. The spray thickness is also predicted and a classic correlation for Sauter Mean Diameter is shown to provide good agreement with experimental results. The analysis was carried out at the operative pressure of 80 bar; two injectors were employed featuring different orifice diameters and flow numbers, as a sort of parametric approach to this spray typology.

  18. A fuzzy decision tree method for fault classification in the steam generator of a pressurized water reactor

    International Nuclear Information System (INIS)

    Zio, Enrico; Baraldi, Piero; Popescu, Irina Crenguta

    2009-01-01

    This paper extends a method previously introduced by the authors for building a transparent fault classification algorithm by combining the fuzzy clustering, fuzzy logic and decision trees techniques. The baseline method transforms an opaque, fuzzy clustering-based classification model into a fuzzy logic inference model based on linguistic rules which can be represented by a decision tree formalism. The classification model thereby obtained is transparent in that it allows direct interpretation and inspection of the model. An extension in the procedure for the development of the fuzzy logic inference model is introduced to allow the treatment of more complicated cases, e.g. splitted and overlapping clusters. The corresponding computational tool developed relies on a number of parameters which can be tuned by the user to optimally compromise the level of transparency of the classification process and its efficiency. A numerical application is presented with regards to the fault classification in the Steam Generator of a Pressurized Water Reactor.

  19. Bursting-protection configuration for cylindrical steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Mutzl, J.

    1979-01-01

    The bursting-protection jacket consists of cylinder courses, being joined together in axial direction, and of a bottom and a cover, being connected by means of axial prestressing tendons. For absorption and transmission of the steam generator weight and the bursting forces the bottom consists of a conical shell, tapered towards the side of the steam generator, and a support ring supporting the bottom circle of the cone. This support ring is built in sandwich construction and is connected with the axial tendons. The conical shell may be reinforced by radial ribs. If a primary coolant pump is built in there is provided for a rocking bearing between its pump casing flange and the bottom. (DG) [de

  20. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    Ahrens, G.; Haury, G.; Lahner, K.; Schatz, A.

    1983-01-01

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.) [de

  1. Physical aspects of the Canadian generation IV supercritical water-cooled pressure tube reactor plant design

    Energy Technology Data Exchange (ETDEWEB)

    Gaudet, M.; Yetisir, M.; Haque, Z. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    The form of the containment building is a function of the requirements imposed by various systems. In order to provide sufficient driving force for naturally-circulated emergency cooling systems, as well as providing a gravity-driven core flooding pool function, the Canadian SCWR reactor design relies on elevation differences between the reactor and the safety systems. These elevation differences, the required cooling pool volumes and the optimum layout of safety-related piping are major factors influencing the plant design. As a defence-in-depth, the containment building and safety systems also provide successive barriers to the unplanned release of radioactive materials, while providing a path for heat flow to the ultimate heat sink, the atmosphere. Access to the reactor for refuelling is from the top of the reactor, with water used as shielding during the refuelling operations. The accessibility to the reactor and protection of the environment are additional factors influencing the plant design. This paper describes the physical implementation of the major systems of the Canadian SCWR within the reactor building, and the position of major plant services relative to the reactor building. (author)

  2. Pressurized water reactor systems

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1975-01-01

    Design and mode of operation of the main PWR components are described: reactor core, pressure vessel and internals, cooling systems with pumps and steam generators, ancillary systems, and waste processing. (TK) [de

  3. Tachometric flowmeters for measuring circulation water parameters in steam generators of the NPPs running on pressurized water reactors

    International Nuclear Information System (INIS)

    Ageev, A.G.; Korolkov, B.M.; Nigmatulin, B.I.; Belov, V.I.; Vasileva, R.V.; Trubkin, N.I.

    1997-01-01

    Tachometric flowmeters used in steam generators for determining the velocity and direction of the flow have a limited service life owing to the use of corundum for the bearing assembly components. Various materials were investigated for the feasibility of using them as alternatives for replacing the corundum bearing and guide bushing under conditions close to the conditions in steam generators: 7 MPa, 260 degC. Good results were obtained with bearing assemblies fabricated from corrosion-resistant steel. Testing of the transducer design and optimization of the technique was accomplished in the course of testing steam generators of the WWER-1000 reactor at the Balakovskaya nuclear power plant. The velocity and direction of flow in the steam generator were measured within a wide range of unit power ratings up to the values corresponding to full power output. The service life of the transducers proved to be not less than 720 hours. The transducer parameters remained unchanged over the entire operation period. (M.D.)

  4. Hybrid Pressure Retarded Osmosis−Membrane Distillation (PRO−MD) Process for Osmotic Power and Clean Water Generation

    KAUST Repository

    Han, Gang; Zuo, Jian; Wan, Chunfeng; Chung, Neal Tai-Shung

    2015-01-01

    unique advantages of high water recovery rate, huge osmotic power generation, well controlled membrane fouling, and minimal environmental impacts. Experimental results show that the PRO−MD hybrid process is promising that not only can harvest osmotic

  5. A study on emergency response guideline during the loss of steam generator secondary heat sink in pressurizer water reactor

    International Nuclear Information System (INIS)

    Yoon, D. J.; Lee, J. Y.; Song, D. S.

    1999-01-01

    A loss of secondary heat sink can occur as a result of several different initiating events, which are a loss of main feedwater during power operation, a loss of off-site power, or any other scenario for which main feedwater is isolated or lost. At this point the opening and closing of the PORV or safety valves will result in a loss of RCS inventory similar in nature to a small break loss of coolant accident. If operator action is not taken, the pressurizer PORV or safety valves will continue to cycle open and closed at the valve setpoint pressure removing RCS inventory and a limited amount of core decay heat until eventually enough inventory will be lost to result in core uncovery. We conclude that a requirement to successfully initiate bleed and feed on steam generator dryout, without any significant core uncovery expected to occur, is that the PORV flow to power ratio must exceed 140 (lbm/hr)/Mwt. For all plants whose PORV capacity is less than 140 (lbm/hr)/Mwt, since symptoms of SG dryout cannot be used to initiate bleed and feed, increasing RCS pressure and temperature or pressure greater than 2335 psig cannot be used. The only alternative symptom available is SG narrow range level. Since Kori 1,2,3 and 4' PORV capacity is more than the criteria, the bleed and feed operation can be initiated at steam generator dryout

  6. technical guidelines for the design and construction of the next generation of nuclear power plants with pressurized water reactors

    International Nuclear Information System (INIS)

    2009-01-01

    These technical guidelines present the opinion of the French 'Groupe Permanent charge des Reacteurs nucleaires' (GPR) concerning the safety philosophy and approach as well as the general safety requirements to be applied for the design and construction of the next generation of nuclear power plants of the PWR (pressurized water reactor) type, assuming the construction of the first units of this generation would start at the beginning of the 21. century. These technical guidelines are based on common work of the French Institut de Protection et de Surete Nucleaire (IPSN) and of the German Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS). Moreover, these technical guidelines were extensively discussed with members of the German Reaktor Sicherheitskommission (RSK) until the end of 1998 and further with German experts. The context of these technical guidelines must be clearly understood. Faced with the current situation of nuclear energy in the world, the various nuclear steam supply system designers are developing new products, all of them claiming their intention of obtaining a higher safety level, by various ways. GPR believes that, for the operation of a new series of nuclear power plants at the beginning of the next century, the adequate way is to derive the design of these plants in an 'evolutionary' way from the design of existing plants, taking into account the operating experience and the in-depth studies conducted for such plants. Nevertheless, introduction of innovative features must also be considered in the frame of the design of the new generation of plants, especially in preventing and mitigating severe accidents. GPR underlines here that a significant improvement of the safety of the next generation of nuclear power plants at the design stage is necessary, compared to existing plants. If the search for improvement is a permanent concern in the field of safety, the necessity of a significant step at the design stage clearly derives from better

  7. The pressurized water reactor

    International Nuclear Information System (INIS)

    Gallagher, J.L.

    1987-01-01

    Pressurized water reactor technology has reached a maturity that has engendered a new surge of innovation, which in turn, has led to significant advances in the technology. These advances, characterized by bold thinking but conservative execution, are resulting in nuclear plant designs which offer significant performance and safety improvements. This paper describes the innovations which are being designed into mainstream PWR technology as well as the desings which are resulting from such innovations. (author)

  8. High Pressure Industrial Water Facility

    Science.gov (United States)

    1992-01-01

    In conjunction with Space Shuttle Main Engine testing at Stennis, the Nordberg Water Pumps at the High Pressure Industrial Water Facility provide water for cooling the flame deflectors at the test stands during test firings.

  9. Steam generator water lancing

    International Nuclear Information System (INIS)

    Kamler, F.; Schneider, W.

    1992-01-01

    The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

  10. Development of internal CRD for next generation BWR-endurance and robustness tests of ball-bearing materials in high-pressure and high-temperature water

    International Nuclear Information System (INIS)

    Shoji Goto; Shuichi Ohmori; Michitsugu Mori; Shohei Kawano; Tadashi Narabayashi; Shinichi Ishizato

    2005-01-01

    An internal CRD using a heatproof ceramics insulated coil is under development to be a competitive and higher performance as Next- Generation BWR. In the case of the 1700MWe next generation BWR, adapting the internal CRDs, the reactor pressure vessel is almost equivalent to that of 1356 MWe ABWR. The endurance and robustness tests were examined in order to confirm the durability of the bearing for the internal CRD. The durability of the ball bearing for the internal CRD was performed in the high-pressure and high-temperature reactor water of current BWR conditions. The experimental results confirmed the durability of rotational numbers for the operation length of 60 years. We added the cruds into water to confirm the robustness of the ball bearing. The test results also showed good robustness even in high-density crud conditions, compared with the current BWR. This program is conducted as one of the selected offers for the advertised technical developments of the Institute of Applied Energy founded by METI (Ministry of Economy, Trade and Industry) of Japan. (authors)

  11. Low cost sonoluminescence experiment in pressurized water

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, L; Insabella, M [LADOP, University of Mar del Plata (Argentina); Bilbao, L [INFIP, University of Buenos Aires and CONICET (Argentina)

    2012-06-19

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  12. Low cost sonoluminescence experiment in pressurized water

    International Nuclear Information System (INIS)

    Bernal, L; Insabella, M; Bilbao, L

    2012-01-01

    We present a low cost design for demostration and mesurements of light emission from a sonoluminescence experiment. Using pressurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  13. Water Pressure. Water in Africa.

    Science.gov (United States)

    Garrett, Carly Sporer

    The Water in Africa Project was realized over a 2-year period by a team of Peace Corps volunteers. As part of an expanded, detailed design, resources were collected from over 90 volunteers serving in African countries, photos and stories were prepared, and standards-based learning units were created for K-12 students. This unit, "Water…

  14. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  15. 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

  16. Free radical generation by non-equilibrium atmospheric pressure plasma in alcohol-water mixtures: an EPR-spin trapping study

    Science.gov (United States)

    Uchiyama, Hidefumi; Ishikawa, Kenji; Zhao, Qing-Li; Andocs, Gabor; Nojima, Nobuyuki; Takeda, Keigo; Krishna, Murali C.; Ishijima, Tatsuo; Matsuya, Yuji; Hori, Masaru; Noguchi, Kyo; Kondo, Takashi

    2018-03-01

    Free radical species in aqueous solution—various alcohol-water reaction mixtures—by exposure to non-equilibrium cold atmospheric pressure Ar plasma (CAP), were monitored using electron paramagnetic resonance spin-trapping techniques with 3, 5-dibromo-4-nitrosobenzene sulfonate as a water soluble nitroso spin trap. The major radical species were formed by H-abstraction from alcohol molecules due to ·OH radicals. In the ethanol-water mixture ·CH2CH2OH produced by H abstraction from CH3 group of the ethanol and ·CH3 radicals were detected. The latter was due to the decomposition of unstable CH3·CHOH to form the ·CH3 radicals and the stable formaldehyde by C-C bond fission. These intermediates are similar to those observed by reaction with ·OH radicals generation in the H2O2-UV photolysis of the reaction mixtures. The evidence of ·CH3 radical formation in the pyrolytic decomposition of the reaction mixtures by exposure to ultrasound or in methane irradiated with microwave plasma have been reported previously. However, the pyrolytic ·CH3 radicals were not found in both plasma and H2O2-UV photolysis condition. These results suggests that free radicals produced by Ar-CAP are most likely due to the reaction between abundant ·OH radicals and alcohol molecules.

  17. Water Pressure in Depth

    Science.gov (United States)

    Lynch, Mary Jean; Zenchak, John

    2011-01-01

    How can a science concept be taught in a way that generates interest, gives students the opportunity to consider other possibilities, does not lock them into one way of doing or seeing things, and gives them some ownership of their learning? These authors searched high and low for the perfect activity to illustrate a key concept for their partner…

  18. High-pressure water facility

    Science.gov (United States)

    2006-01-01

    NASA Test Operations Group employees, from left, Todd Pearson, Tim Delcuze and Rodney Wilkinson maintain a water pump in Stennis Space Center's high-pressure water facility. The three were part of a group of employees who rode out Hurricane Katrina at the facility and helped protect NASA's rocket engine test complex.

  19. Pressurized water reactor flow arrangement

    International Nuclear Information System (INIS)

    Gibbons, J.F.; Knapp, R.W.

    1980-01-01

    A flow path is provided for cooling the control rods of a pressurized water reactor. According to this scheme, a small amount of cooling water enters the control rod guide tubes from the top and passes downwards through the tubes before rejoining the main coolant flow and passing through the reactor core. (LL)

  20. High pressure experimental water loop

    International Nuclear Information System (INIS)

    Grenon, M.

    1958-01-01

    A high pressure experimental water loop has been made for studying the detection and evolution of cladding failure in a pressurized reactor. The loop has been designed for a maximum temperature of 360 deg. C, a maximum of 160 kg/cm 2 and flow rates up to 5 m 3 /h. The entire loop consists of several parts: a main circuit with a canned rotor circulation pump, steam pressurizer, heating tubes, two hydro-cyclones (one de-gasser and one decanter) and one tubular heat exchanger; a continuous purification loop, connected in parallel, comprising pressure reducing valves and resin pots which also allow studies of the stability of resins under pressure, temperature and radiation; following the gas separator is a gas loop for studying the recombination of the radiolytic gases in the steam phase. The preceding circuits, as well as others, return to a low pressure storage circuit. The cold water of the low pressure storage flask is continuously reintroduced into the high pressure main circuit by means of a return pump at a maximum head of 160 kg /cm 2 , and adjusted to the pressurizer level. This loop is also a testing bench for the tight high pressure apparatus. The circulating pump and the connecting flanges (Oak Ridge type) are water-tight. The feed pump and the pressure reducing valves are not; the un-tight ones have a system of leak recovery. To permanently check the tightness the circuit has been fitted with a leak detection system (similar to the HRT one). (author) [fr

  1. Water cooled static pressure probe

    Science.gov (United States)

    Lagen, Nicholas T. (Inventor); Eves, John W. (Inventor); Reece, Garland D. (Inventor); Geissinger, Steve L. (Inventor)

    1991-01-01

    An improved static pressure probe containing a water cooling mechanism is disclosed. This probe has a hollow interior containing a central coolant tube and multiple individual pressure measurement tubes connected to holes placed on the exterior. Coolant from the central tube symmetrically immerses the interior of the probe, allowing it to sustain high temperature (in the region of 2500 F) supersonic jet flow indefinitely, while still recording accurate pressure data. The coolant exits the probe body by way of a reservoir attached to the aft of the probe. The pressure measurement tubes are joined to a single, larger manifold in the reservoir. This manifold is attached to a pressure transducer that records the average static pressure.

  2. Low cost sonoluminescence experiment in pressurized water

    Science.gov (United States)

    Bernal, L.; Insabella, M.; Bilbao, L.

    2012-06-01

    We present a low cost design for demostration and mesurements of light emmision from a sonoluminescence experiment. Using presurized water introduced in an acrylic cylinder and one piezoelectric from an ultrasonic cleaner, we are able to generate cavitacion zones with emission of light. The use of argon to pressurize the water improves the emission an the light can be seen at naked eye in a softlit ambient.

  3. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  4. Osmosis-Based Pressure Generation: Dynamics and Application

    Science.gov (United States)

    Li, Suyi; Billeh, Yazan N.; Wang, K. W.; Mayer, Michael

    2014-01-01

    This paper describes osmotically-driven pressure generation in a membrane-bound compartment while taking into account volume expansion, solute dilution, surface area to volume ratio, membrane hydraulic permeability, and changes in osmotic gradient, bulk modulus, and degree of membrane fouling. The emphasis lies on the dynamics of pressure generation; these dynamics have not previously been described in detail. Experimental results are compared to and supported by numerical simulations, which we make accessible as an open source tool. This approach reveals unintuitive results about the quantitative dependence of the speed of pressure generation on the relevant and interdependent parameters that will be encountered in most osmotically-driven pressure generators. For instance, restricting the volume expansion of a compartment allows it to generate its first 5 kPa of pressure seven times faster than without a restraint. In addition, this dynamics study shows that plants are near-ideal osmotic pressure generators, as they are composed of many small compartments with large surface area to volume ratios and strong cell wall reinforcements. Finally, we demonstrate two applications of an osmosis-based pressure generator: actuation of a soft robot and continuous volume delivery over long periods of time. Both applications do not need an external power source but rather take advantage of the energy released upon watering the pressure generators. PMID:24614529

  5. Osmosis-based pressure generation: dynamics and application.

    Science.gov (United States)

    Bruhn, Brandon R; Schroeder, Thomas B H; Li, Suyi; Billeh, Yazan N; Wang, K W; Mayer, Michael

    2014-01-01

    This paper describes osmotically-driven pressure generation in a membrane-bound compartment while taking into account volume expansion, solute dilution, surface area to volume ratio, membrane hydraulic permeability, and changes in osmotic gradient, bulk modulus, and degree of membrane fouling. The emphasis lies on the dynamics of pressure generation; these dynamics have not previously been described in detail. Experimental results are compared to and supported by numerical simulations, which we make accessible as an open source tool. This approach reveals unintuitive results about the quantitative dependence of the speed of pressure generation on the relevant and interdependent parameters that will be encountered in most osmotically-driven pressure generators. For instance, restricting the volume expansion of a compartment allows it to generate its first 5 kPa of pressure seven times faster than without a restraint. In addition, this dynamics study shows that plants are near-ideal osmotic pressure generators, as they are composed of many small compartments with large surface area to volume ratios and strong cell wall reinforcements. Finally, we demonstrate two applications of an osmosis-based pressure generator: actuation of a soft robot and continuous volume delivery over long periods of time. Both applications do not need an external power source but rather take advantage of the energy released upon watering the pressure generators.

  6. Osmosis-based pressure generation: dynamics and application.

    Directory of Open Access Journals (Sweden)

    Brandon R Bruhn

    Full Text Available This paper describes osmotically-driven pressure generation in a membrane-bound compartment while taking into account volume expansion, solute dilution, surface area to volume ratio, membrane hydraulic permeability, and changes in osmotic gradient, bulk modulus, and degree of membrane fouling. The emphasis lies on the dynamics of pressure generation; these dynamics have not previously been described in detail. Experimental results are compared to and supported by numerical simulations, which we make accessible as an open source tool. This approach reveals unintuitive results about the quantitative dependence of the speed of pressure generation on the relevant and interdependent parameters that will be encountered in most osmotically-driven pressure generators. For instance, restricting the volume expansion of a compartment allows it to generate its first 5 kPa of pressure seven times faster than without a restraint. In addition, this dynamics study shows that plants are near-ideal osmotic pressure generators, as they are composed of many small compartments with large surface area to volume ratios and strong cell wall reinforcements. Finally, we demonstrate two applications of an osmosis-based pressure generator: actuation of a soft robot and continuous volume delivery over long periods of time. Both applications do not need an external power source but rather take advantage of the energy released upon watering the pressure generators.

  7. 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

  8. Pressurizer and steam-generator behavior under PWR transient conditions

    International Nuclear Information System (INIS)

    Wahba, A.B.; Berta, V.T.; Pointner, W.

    1983-01-01

    Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

  9. Fundamentals of pressurized water reactors

    International Nuclear Information System (INIS)

    Murray, L.

    1982-01-01

    In many countries, the pressurized water reactor (PWR) is the most widely used, even though it requires enrichment of the uranium to about 3% in U-235 and the moderator-coolant must be maintained at a high pressure, about 2200 pounds per square inch. Our objective in this series of seven lectures is to describe the design and operating characteristics of the PWR system, discuss the reactor physics methods used to evaluate performance, examine the way fuel is consumed and produced, study the instrumentation system, review the physics measurements made during initial startup of the reactor, and outline the administrative aspects of starting up a reactor and operating it safely and effectively

  10. Pressurized water reactor inspection procedures

    International Nuclear Information System (INIS)

    Heinrich, D.; Mueller, G.; Otte, H.J.; Roth, W.

    1998-01-01

    Inspections of the reactor pressure vessels of pressurized water reactors (PWR) so far used to be carried out with different central mast manipulators. For technical reasons, parallel inspections of two manipulators alongside work on the refueling cavity, so as to reduce the time spent on the critical path in a revision outage, are not possible. Efforts made to minimize the inspection time required with one manipulator have been successful, but their effects are limited. Major reductions in inspection time can be achieved only if inspections are run with two manipulators in parallel. The decentralized manipulator built by GEC Alsthom Energie and so far emmployed in boiling water reactors in the USA, Spain, Switzerland and Japan allows two systems to be used in parallel, thus reducing the time required for standard inspection of a pressure vessel from some six days to three days. These savings of approximately three days are made possible without any compromises in terms of positioning by rail-bound systems. During inspection, the reactor refueling cavity is available for other revision work without any restrictions. The manipulator can be used equally well for inspecting standard PWR, PWR with a thermal shield, for inspecting the land between in-core instrumentation nozzles, BWR with and without jet pumps (complementary inspection), and for inspecting core support shrouds. (orig.) [de

  11. Water box for steam generator

    International Nuclear Information System (INIS)

    Lecomte, Robert; Viaud, Michel.

    1975-01-01

    This invention relates to a water box for connecting an assembly composed of a vertical steam generator and a vertical pump to the vessel of the nuclear reactor, the assembly forming the primary cooling system of a pressurised water reactor. This invention makes it easy to dismantle the pump on the water box without significant loss of water in the primary cooling system of the reactor and particularly without it being necessary to drain the water contained in the steam generator beforehand. It makes it possible to shorten the time required for dismantling the primary pump in order to service or repair it and makes dismantling safer in that the dismantling does not involve draining the steam generator and therefore the critical storage of a large amount of cooling water that has been in contact with the fuel assemblies of the nuclear reactor core [fr

  12. Simulation of low pressure water hammer

    Science.gov (United States)

    Himr, D.; Habán, V.

    2010-08-01

    Numerical solution of water hammer is presented in this paper. The contribution is focused on water hammer in the area of low pressure, which is completely different than high pressure case. Little volume of air and influence of the pipe are assumed in water, which cause sound speed change due to pressure alterations. Computation is compared with experimental measurement.

  13. The European pressurized water reactor

    International Nuclear Information System (INIS)

    Leny, J.C.

    1993-01-01

    The present state of development of the European Pressurized Water Reactor (EPR) is outlined. During the so-called harmonization phase, the French and German utilities drew up their common requirements and evaluated the reactor concept developed until then with respect to these requirements. A main result of the harmonization phase was the issue, in September 1993, of the 'EPR Conceptual Safety Feature Review File' to be jointly assessed by the safety authorities in France and Germany. The safety objectives to be met by the EPR are specified in the second part of the paper, and some details of the primary and secondary side safety systems are given. (orig.) [de

  14. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1985-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (orig./PW)

  15. Pressurized water reactor with reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1980-01-01

    The pressure vessel has a cylindrical jacket with a domed floor. A guide is arranged on the domed floor to even out the flow in the core. It consists of a cylindrical jacket, whose lower end has slots and fins. These fins are welded to the domed floor. (DG) [de

  16. Osmotic generation of 'anomalous' fluid pressures in geological environments

    Science.gov (United States)

    Neuzii, C.E.

    2000-01-01

    Osmotic pressures are generated by differences in chemical potential of a solution across a membrane. But whether osmosis can have a significant effect on the pressure of fluids in geological environments has been controversial, because the membrane properties of geological media are poorly understood. 'Anomalous' pressures - large departures from hydrostatic pressure that are not explicable in terms of topographic or fluid-density effects are widely found in geological settings, and are commonly considered to result from processes that alter the pore or fluid volume, which in turn implies crustal changes happening at a rate too slow to observe directly. Yet if osmosis can explain some anomalies, there is no need to invoke such dynamic geological processes in those cases. Here I report results of a nine- year in situ measurement of fluid pressures and solute concentrations in shale that are consistent with the generation of large (up to 20 MPa) osmotic-pressure anomalies which could persist for tens of millions of years. Osmotic pressures of this magnitude and duration can explain many of the pressure anomalies observed in geological settings. The require, however, small shale porosity and large contrasts in the amount of dissolved solids in the pore waters - criteria that may help to distinguish between osmotic and crystal-dynamic origins of anomalous pressures.

  17. Water hammer characteristics of integral pressurized water reactor primary loop

    International Nuclear Information System (INIS)

    Zuo, Qiaolin; Qiu, Suizheng; Lu, Wei; Tian, Wenxi; Su, Guanghui; Xiao, Zejun

    2013-01-01

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions

  18. Water hammer characteristics of integral pressurized water reactor primary loop

    Energy Technology Data Exchange (ETDEWEB)

    Zuo, Qiaolin [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Qiu, Suizheng, E-mail: szqiu@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Lu, Wei; Tian, Wenxi; Su, Guanghui [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an, Shanxi 710049 (China); Xiao, Zejun [Nuclear Power Institute of China, Chengdu, Sichuan 610041 (China)

    2013-08-15

    Highlights: • Water hammer models developed for IPWR primary loop using MOC. • Good agreement between the developed code and the experiment. • The good agreement between WAHAP and Flowmaster can validate the equations in WAHAP. • The primary loop of IPWR suffers from slight water hammer impact. -- Abstract: The present work discussed the single-phase water hammer phenomenon, which was caused by the four-pump-alternate startup in an integral pressurized water reactor (IPWR). A new code named water hammer program (WAHAP) was developed independently based on the method of characteristic to simulate hydraulic transients in the primary system of IPWR and its components such as reactor core, once-through steam generators (OTSG), the main coolant pumps and so on. Experimental validation for the correctness of the equations and models in WAHAP was carried out and the models fit the experimental data well. Some important variables were monitored including transient volume flow rates, opening angle of valve disc and pressure drop in valves. The water hammer commercial software Flowmaster V7 was also employed to compare with WAHAP and the good agreement can validate the equations in WAHAP. The transient results indicated that the primary loop of IPWR suffers from slight water hammer impact under pump switching conditions.

  19. Next generation light water reactors

    International Nuclear Information System (INIS)

    Omoto, Akira

    1992-01-01

    In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

  20. High temperature pressure water's blowdown into water. Experimental results

    International Nuclear Information System (INIS)

    Ishida, Toshihisa; Kusunoki, Tsuyoshi; Kasahara, Yoshiyuki; Iida, Hiromasa

    1994-01-01

    The purpose of the present experimental study is to clarify the phenomena in blowdown of high temperature and pressure water in pressure vessel into the containment water for evaluation of design of an advanced marine reactor(MRX). The water blown into the containment water flushed and formed steam jet plume. The steam jet condensed in the water, but some stream penetrated to gas phase of containment and contributed to increase of containment pressure. (author)

  1. Managing water pressure for water savings in developing countries

    African Journals Online (AJOL)

    2014-03-03

    Mar 3, 2014 ... effort into providing customers with a reliable level of service, often via poor water ... budgets. There are many factors contributing to water losses in water .... given relationship does not reflect the impact of pressure on.

  2. 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

  3. Weld region corrosion during chemical cleaning of PWR [pressurized-water reactor] steam generators: Volume 2, Tests and analyses: Final report

    International Nuclear Information System (INIS)

    Barna, J.L.; Bozeka, S.A.; Jevec, J.M.

    1987-07-01

    The potential for preferential corrosion of steam generator weld regions during chemical cleaning using the generic SGOG solvents was investigated. The investigations included development and use of a corrosion assessment test facility which measured corrosion currents in a realistic model of the steam generator geometry in the vicinity of a specific weld during a simulated chemical dissolution of sludge consisting of essentially pure magnetite. A corrosion monitoring technique was developed and qualified. In this technique free corrosion rates measured by linear polarization techniques are added to corrosion rates calculated from galvanic current measured using a zero resistance ammeter to give an estimate of total corrosion rate for a galvanically corroding material. An analytic modeling technique was developed and proved useful in determining the size requirements for the weld region mockup used in the corrosion assessment test facility. The technique predicted galvanic corrosion rates consistent with that observed in a corrosion assessement test when polarization data used as model input were obtained on-line during the test. The test results obtained during this investigation indicated that chemical cleaning using the SGOG magnetite dissolution solvent can be performed with a small amount of corrosion of secondary side internals and pressure boundary welds. The maximum weld region corrosion measured during a typical chemical cleaning cycle to remove essentially pure magnetite sludge was about 8 mils. However, additional site specific weld region corrosion assessment testing and qualification will be required prior to chemical cleaning steam generators at a specific plant. Recommendations for site specific qualification of chemical cleaning processes and for use of process monitors and on-line corrosion instrumentation are included in this report

  4. Pressure pressure-balanced pH sensing system for high temperature and high pressure water

    International Nuclear Information System (INIS)

    Tachibana, Koji

    1995-01-01

    As for the pH measurement system for high temperature, high pressure water, there have been the circumstances that first the reference electrodes for monitoring corrosion potential were developed, and subsequently, it was developed for the purpose of maintaining the soundness of metallic materials in high temperature, high pressure water in nuclear power generation. In the process of developing the reference electrodes for high temperature water, it was clarified that the occurrence of stress corrosion cracking in BWRs is closely related to the corrosion potential determined by dissolved oxygen concentration. As the types of pH electrodes, there are metal-hydrogen electrodes, glass electrodes, ZrO 2 diaphragm electrodes and TiO 2 semiconductor electrodes. The principle of pH measurement using ZrO 2 diaphragms is explained. The pH measuring system is composed of YSZ element, pressure-balanced type external reference electrode, pressure balancer and compressed air vessel. The stability and pH response of YSZ elements are reported. (K.I.)

  5. High-pressure-induced water penetration into 3-isopropylmalate dehydrogenase

    International Nuclear Information System (INIS)

    Nagae, Takayuki; Kawamura, Takashi; Chavas, Leonard M. G.; Niwa, Ken; Hasegawa, Masashi; Kato, Chiaki; Watanabe, Nobuhisa

    2012-01-01

    Structures of 3-isopropylmalate dehydrogenase were determined at pressures ranging from 0.1 to 650 MPa. Comparison of these structures gives a detailed picture of the swelling of a cavity at the dimer interface and the generation of a new cleft on the molecular surface, which are accompanied by water penetration. Hydrostatic pressure induces structural changes in proteins, including denaturation, the mechanism of which has been attributed to water penetration into the protein interior. In this study, structures of 3-isopropylmalate dehydrogenase (IPMDH) from Shewanella oneidensis MR-1 were determined at about 2 Å resolution under pressures ranging from 0.1 to 650 MPa using a diamond anvil cell (DAC). Although most of the protein cavities are monotonically compressed as the pressure increases, the volume of one particular cavity at the dimer interface increases at pressures over 340 MPa. In parallel with this volume increase, water penetration into the cavity could be observed at pressures over 410 MPa. In addition, the generation of a new cleft on the molecular surface accompanied by water penetration could also be observed at pressures over 580 MPa. These water-penetration phenomena are considered to be initial steps in the pressure-denaturation process of IPMDH

  6. Condensation induced water hammer in steam generators

    International Nuclear Information System (INIS)

    Jones, O.C. Jr.; Saha, P.; Wu, B.J.C.; Ginsberg, T.

    1979-06-01

    The case of condensation induced water hammer in nuclear steam generators is summarized, including both feed ring-type and economizer-type geometries. A slug impact model is described and used to demonstrate the parametric dependence of the impact pressures on heat transfer rates, initial pressures, and relative initial slug and void lengths. The results of the parametric study are related also to the economizer geometry and a suggested alternative model is presented. The importance of concerns regarding attenuation of shocks in two-phase media is delineated, and a simple experiment is described which was used to determine negligible attenuation within the accuracy of the experiment for void fractions up to over 30% in bubbly and slug flows

  7. Water-Based Pressure-Sensitive Paints

    Science.gov (United States)

    Jordan, Jeffrey D.; Watkins, A. Neal; Oglesby, Donald M.; Ingram, JoAnne L.

    2006-01-01

    Water-based pressure-sensitive paints (PSPs) have been invented as alternatives to conventional organic-solvent-based pressure-sensitive paints, which are used primarily for indicating distributions of air pressure on wind-tunnel models. Typically, PSPs are sprayed onto aerodynamic models after they have been mounted in wind tunnels. When conventional organic-solvent-based PSPs are used, this practice creates a problem of removing toxic fumes from inside the wind tunnels. The use of water-based PSPs eliminates this problem. The waterbased PSPs offer high performance as pressure indicators, plus all the advantages of common water-based paints (low toxicity, low concentrations of volatile organic compounds, and easy cleanup by use of water).

  8. Water Delivery--It's All about Pressure

    Science.gov (United States)

    Roman, Harry T.

    2005-01-01

    There is a great deal of wisdom in the old saying "water seeks its level." In fact, the concept has bearing on a very practical side of human life as well, since the public water delivery system is based on it. In this article, the author discusses the concept behind water pressure and describes how the water systems work based on this concept.…

  9. Pressurized water reactor fuel rod design methodology

    International Nuclear Information System (INIS)

    Silva, A.T.; Esteves, A.M.

    1988-08-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  10. Peer Pressure: An Issue That Crosses Generations.

    Science.gov (United States)

    Kittredge, Karen; McCarthy, Alice R.

    2000-01-01

    Recent research on peer pressure shows that: parents are important to teens, today's teens face unique challenges, and teaching teens to say no does not mean losing friends. The paper presents parenting tips for countering peer pressure, noting the influence of adult peer pressure on children. A sidebar examines the right age to start talking to…

  11. The steam generating heavy water reactor

    International Nuclear Information System (INIS)

    Middleton, J.E.

    1975-01-01

    A review is presented on the evolution of the SGHWR concept by the United Kingdom Atomic Energy Authority and the production of early commercial designs, together with later development by the Design and Construction Companies. This is followed by a description of the current commercial design. Possible future developments are suggested. The many advantageous features of the concept are mentioned with a view to supporting optimism for the future of the system. Headings include the following: safety criteria and risk assessment; emergency core cooling system design and development; protective systems; reactor coolant system; reactivity control; off-load refuelling; pressure containment; 'fence' header coolant circuit design; feed water injection; continuous spray cooling; low pressure cooling systems for residual heat removal during refuelling; high pressure cooling system for guaranteed feed water supply; auxiliary systems; structural materials; calandria and neutron shields; fuel element development; alternative loop circuit design; future developments (use of hydraulic diodes to provide a substantial reverse flow resistance by the generation of a vortex; multi-drum and multi-pump schemes; refuelling alternatives; coolant circuit inversion; use of superheat channels). (U.K.)

  12. Water-Based Pressure Sensitive Paint

    Science.gov (United States)

    Oglesby, Donald M.; Ingram, JoAnne L.; Jordan, Jeffrey D.; Watkins, A. Neal; Leighty, Bradley D.

    2004-01-01

    Preparation and performance of a water-based pressure sensitive paint (PSP) is described. A water emulsion of an oxygen permeable polymer and a platinum porphyrin type luminescent compound were dispersed in a water matrix to produce a PSP that performs well without the use of volatile, toxic solvents. The primary advantages of this PSP are reduced contamination of wind tunnels in which it is used, lower health risk to its users, and easier cleanup and disposal. This also represents a cost reduction by eliminating the need for elaborate ventilation and user protection during application. The water-based PSP described has all the characteristics associated with water-based paints (low toxicity, very low volatile organic chemicals, and easy water cleanup) but also has high performance as a global pressure sensor for PSP measurements in wind tunnels. The use of a water-based PSP virtually eliminates the toxic fumes associated with the application of PSPs to a model in wind tunnels.

  13. Performance Limiting Effects in Power Generation from Salinity Gradients by Pressure Retarded Osmosis

    KAUST Repository

    Yip, Ngai Yin; Elimelech, Menachem

    2011-01-01

    Pressure retarded osmosis has the potential to utilize the free energy of mixing when fresh river water flows into the sea for clean and renewable power generation. Here, we present a systematic investigation of the performance limiting phenomena

  14. Cold plasma brush generated at atmospheric pressure

    International Nuclear Information System (INIS)

    Duan Yixiang; Huang, C.; Yu, Q. S.

    2007-01-01

    A cold plasma brush is generated at atmospheric pressure with low power consumption in the level of several watts (as low as 4 W) up to tens of watts (up to 45 W). The plasma can be ignited and sustained in both continuous and pulsed modes with different plasma gases such as argon or helium, but argon was selected as a primary gas for use in this work. The brush-shaped plasma is formed and extended outside of the discharge chamber with typical dimension of 10-15 mm in width and less than 1.0 mm in thickness, which are adjustable by changing the discharge chamber design and operating conditions. The brush-shaped plasma provides some unique features and distinct nonequilibrium plasma characteristics. Temperature measurements using a thermocouple thermometer showed that the gas phase temperatures of the plasma brush are close to room temperature (as low as 42 deg. C) when running with a relatively high gas flow rate of about 3500 ml/min. For an argon plasma brush, the operating voltage from less than 500 V to about 2500 V was tested, with an argon gas flow rate varied from less than 1000 to 3500 ml/min. The cold plasma brush can most efficiently use the discharge power as well as the plasma gas for material and surface treatment. The very low power consumption of such an atmospheric argon plasma brush provides many unique advantages in practical applications including battery-powered operation and use in large-scale applications. Several polymer film samples were tested for surface treatment with the newly developed device, and successful changes of the wettability property from hydrophobic to hydrophilic were achieved within a few seconds

  15. GENERATION OF HIGH SHOCK PRESSURES BY LASER PULSES

    OpenAIRE

    Romain , J.

    1984-01-01

    Aspects of laser generated high shock pressures and results obtained over the last years are reviewed. Shock pressures up to 5 TPa inferred from shock velocity measurements are reported. Effects of laser wavelength, intensity and 2-D plasma expansion on the generated shock pressure are discussed. The hydrodynamic efficiency determined from various data including new results at 0,26 µm wavelength outlines the advantage of short wavelengths for producing very high pressures. The possibility of ...

  16. PWR [pressurized water reactor] pressurizer transient response: Final report

    International Nuclear Information System (INIS)

    Murphy, S.I.

    1987-08-01

    To predict PWR pressurizer transients, Ahl proposed a three region model with a universal coefficient to represent condensation on the water surface. Specifically, this work checks the need for three regions and the modeling of the interfacial condensation coefficient. A computer model has been formulated using the basic mass and energy conservation laws. A two region vapor and liquid model was first used to predict transients run on a one-eleventh scale Freon pressurizer. These predictions verified the need for a second liquid region. As a result, a three region model was developed and used to predict full-scale pressurizer transients at TMI-2, Shippingport, and Stade. Full-scale pressurizer predictions verified the three region model and pointed out the shortcomings of Ahl's universal condensation coefficient. In addition, experiments were run using water at low pressure to study interface condensation. These experiments showed interface condensation to be significant only when spray flow is turned on; this result was incorporated in the final three region model

  17. Reactor vessel pressure transient protection for pressurized water reactors

    International Nuclear Information System (INIS)

    Zech, G.

    1978-09-01

    During the past few years the NRC has been studying the issue of protection of the reactor pressure vessels at Pressurized Water Reactors (PWRs) from transients when the vessels are at a relatively low temperature. This effort was prompted by concerns related to the safety margins available to vessel damage as a result of such events. Nuclear Reactor Regulation Category A Technical Activity No. A-26 was established to set forth the NRC plan for resolution of the generic aspects of this safety issue. The purpose of the report is to document the completion of this generic technical activity

  18. Water-Pressure Distribution on Seaplane Float

    Science.gov (United States)

    Thompson, F L

    1929-01-01

    The investigation presented in this report was conducted for the purpose of determining the distribution and magnitude of water pressures likely to be experienced on seaplane hulls in service. It consisted of the development and construction of apparatus for recording water pressures lasting one one-hundredth second or longer and of flight tests to determine the water pressures on a UO-1 seaplane float under various conditions of taxiing, taking off, and landing. The apparatus developed was found to operate with satisfactory accuracy and is suitable for flight tests on other seaplanes. The tests on the UO-1 showed that maximum pressures of about 6.5 pounds per square inch occur at the step for the full width of the float bottom. Proceeding forward from the step the maximum pressures decrease in magnitude uniformly toward the bow, and the region of highest pressures narrows toward the keel. Immediately abaft the step the maximum pressures are very small, but increase in magnitude toward the stern and there once reached a value of about 5 pounds per square inch. (author)

  19. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    Energy Technology Data Exchange (ETDEWEB)

    Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  20. Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

    International Nuclear Information System (INIS)

    Monteleone, S.

    1997-02-01

    This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  1. China's coal-fired power plants impose pressure on water resources

    NARCIS (Netherlands)

    Zhang, Xinxin; Liu, Junguo; Tang, Yu; Zhao, Xu; Yang, Hong; Gerbens-Leenes, P.W.; Vliet, van Michelle T.H.; Yan, Jinyue

    2017-01-01

    Coal is the dominant fuel for electricity generation around the world. This type of electricity generation uses large amounts of water, increasing pressure on water resources. This calls for an in-depth investigation in the water-energy nexus of coal-fired electricity generation. In China,

  2. Pressure dependence of dynamical heterogeneity in water

    International Nuclear Information System (INIS)

    Teboul, Victor

    2008-01-01

    Using molecular dynamics simulations we investigate the effect of pressure on the dynamical heterogeneity in water. We show that the effect of a pressure variation in water is qualitatively different from the effect of a temperature variation on the dynamical heterogeneity in the liquid. We observe a strong decrease of the aggregation of molecules of low mobility together with a decrease of the characteristic time associated with this aggregation. However, the aggregation of the most mobile molecules and the characteristic time of this aggregation are only slightly affected. In accordance with this result, the non-Gaussian parameter shows an important decrease with pressure while the characteristic time t* of the non-Gaussian parameter is only slightly affected. These results highlight then the importance of pressure variation investigations in low temperature liquids on approach to the glass transition

  3. Reuse of waste water from high pressure water jet decontamination for reactor decommissioning scrap metal

    International Nuclear Information System (INIS)

    Deng Junxian; Li Xin; Hou Huijuan

    2011-01-01

    For recycle and reuse of reactor decommissioning scrap metal by high pressure water jet decontamination, large quantity of radioactive waste water will be generated. To save the cost of radioactive waste water treatment and to reduce the cost of the scrap decontamination, this part of radioactive waste water should be reused. Most of the radioactivities in the decontamination waste water come from the solid particle in the water. Thus to reuse the waste water, the solid particle in the waster should be removed. Different possible treatment technologies have been investigated. By cost benefit analysis the centrifugal separation technology is selected. (authors)

  4. 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

  5. High pressure water jet cutting and stripping

    Science.gov (United States)

    Hoppe, David T.; Babai, Majid K.

    1991-01-01

    High pressure water cutting techniques have a wide range of applications to the American space effort. Hydroblasting techniques are commonly used during the refurbishment of the reusable solid rocket motors. The process can be controlled to strip a thermal protective ablator without incurring any damage to the painted surface underneath by using a variation of possible parameters. Hydroblasting is a technique which is easily automated. Automation removes personnel from the hostile environment of the high pressure water. Computer controlled robots can perform the same task in a fraction of the time that would be required by manual operation.

  6. High pressure water jet mining machine

    Science.gov (United States)

    Barker, Clark R.

    1981-05-05

    A high pressure water jet mining machine for the longwall mining of coal is described. The machine is generally in the shape of a plowshare and is advanced in the direction in which the coal is cut. The machine has mounted thereon a plurality of nozzle modules each containing a high pressure water jet nozzle disposed to oscillate in a particular plane. The nozzle modules are oriented to cut in vertical and horizontal planes on the leading edge of the machine and the coal so cut is cleaved off by the wedge-shaped body.

  7. Generation of high shock pressures by laser pulses

    Energy Technology Data Exchange (ETDEWEB)

    Romain, J.P. (GRECO ILM, Laboratoire d' Energetique et Detonique, E.N.S.M.A., 86 - Poitiers (France))

    1984-11-01

    Aspects of laser generated high shock pressures and results obtained over the last years are reviewed. Shock pressures up to 5 TPa inferred from shock velocity measurements are reported. Effects of laser wavelength, intensity and 2-D plasma expansion on the generated shock pressure are discussed. The hydrodynamic efficiency determined from various data including new results at 0,26 ..mu..m wavelength outlines the advantage of short wavelengths for producing very high pressures. The possibility of achieving shock pressures in the 10 TPa range with the use of the impedance match technique is examined.

  8. Generation of high shock pressures by laser pulses

    International Nuclear Information System (INIS)

    Romain, J.P.

    1984-01-01

    Aspects of laser generated high shock pressures and results obtained over the last years are reviewed. Shock pressures up to 5 TPa inferred from shock velocity measurements are reported. Effects of laser wavelength, intensity and 2-D plasma expansion on the generated shock pressure are discussed. The hydrodynamic efficiency determined from various data including new results at 0,26 μm wavelength outlines the advantage of short wavelengths for producing very high pressures. The possibility of achieving shock pressures in the 10 TPa range with the use of the impedance match technique is examined

  9. Cavitation nuclei in water exposed to transient pressures

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2015-01-01

    A model of skin-stabilized interfacial cavitation nuclei and their response to tensile and compressive stressing is presented. The model is evaluated in relation to experimental tensile strength results for water at rest at the bottom of an open water-filled container at atmospheric pressure...... and room temperature. These results are obtained by recording the initial growth of cavities generated by a short tensile pulse applied to the bottom of the container. It is found that the cavitation nuclei shift their tensile strength depending on their pressure history. Static pressurization...... for an extended period of time prior to testing is known to increase the tensile strength of water, but little information is available on how it is affected by compression pulses of short duration. This is addressed by imposing compression pulses of approximately 1 ms duration and a peak intensity of a few bar...

  10. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  11. Continuous positive airway pressure therapy: new generations.

    LENUS (Irish Health Repository)

    Garvey, John F

    2010-02-01

    Continuous positive airway pressure (CPAP) is the treatment of choice for obstructive sleep apnoea syndrome (OSAS). However, CPAP is not tolerated by all patients with OSAS and alternative modes of pressure delivery have been developed to overcome pressure intolerance, thereby improving patient comfort and adherence. Auto-adjustable positive airway pressure (APAP) devices may be utilised for the long-term management of OSAS and may also assist in the initial diagnosis of OSAS and titration of conventional CPAP therapy. Newer modalities such as C-Flex and A-Flex also show promise as treatment options in the future. However, the evidence supporting the use of these alternative modalities remains scant, in particular with regard to long-term cardiovascular outcomes. In addition, not all APAP devices use the same technological algorithms and data supporting individual APAP devices cannot be extrapolated to support all. Further studies are required to validate the roles of APAP, C-Flex and A-Flex. In the interim, standard CPAP therapy should continue as the mainstay of OSAS management.

  12. Continuous positive airway pressure therapy: new generations.

    LENUS (Irish Health Repository)

    Garvey, John F

    2012-02-01

    Continuous positive airway pressure (CPAP) is the treatment of choice for obstructive sleep apnoea syndrome (OSAS). However, CPAP is not tolerated by all patients with OSAS and alternative modes of pressure delivery have been developed to overcome pressure intolerance, thereby improving patient comfort and adherence. Auto-adjustable positive airway pressure (APAP) devices may be utilised for the long-term management of OSAS and may also assist in the initial diagnosis of OSAS and titration of conventional CPAP therapy. Newer modalities such as C-Flex and A-Flex also show promise as treatment options in the future. However, the evidence supporting the use of these alternative modalities remains scant, in particular with regard to long-term cardiovascular outcomes. In addition, not all APAP devices use the same technological algorithms and data supporting individual APAP devices cannot be extrapolated to support all. Further studies are required to validate the roles of APAP, C-Flex and A-Flex. In the interim, standard CPAP therapy should continue as the mainstay of OSAS management.

  13. Evaluation of a continuous-positive pressure generating system

    Directory of Open Access Journals (Sweden)

    Herrera N

    2016-03-01

    Full Text Available Nestor Herrera,1,2 Roberto Regnícoli,1,2 Mariel Murad1,2 1Neonatology Unit, Italian Hospital Garibaldi, Rosario, Argentina; 2Experimental Medicine and Surgery Unit, Italian University Institute of Rosario, Argentina Abstract: The use of systems that apply continuous-positive airway pressure by means of noninvasive methods is widespread in the neonatal care practice and has been associated with a decrease in the use of invasive mechanical ventilation, less administration of exogenous surfactant, and bronchopulmonary dysplasia. Few experimental studies on the functioning of the neonatology systems that generate continuous-positive airway pressure have been reported. A flow resistor system associated with an underwater seal resistor in a lung test model was described, and it was compared with an underwater seal threshold resistor system. Important differences in the pressures generated in the different systems studied were verified. The generation of pressure was associated with the immersion depth and the diameter of the bubble tubing. The flow resistor associated with an underwater seal, with small bubble tubing, showed no important differences in the evaluated pressures, exerting a stabilizing effect on the generated pressures. The importance of measuring the pressure generated by the different systems studied was verified, due to the differences between the working pressures set and the pressures measured. Keywords: continuous-positive pressure, flow and threshold resistor, BCPAP

  14. Functional systems of a pressurized water reactor

    International Nuclear Information System (INIS)

    Heinzel, V.

    1982-01-01

    The main topics, discussed in the present paper, are: - Principle design of the reactor coolant system - reactor pressure vessel with internals - containment design - residual heat removal and emergency cooling systems - nuclear component cooling systems - emergency feed water systems - plant electric power supply system. (orig./RW)

  15. Water cycles in closed ecological systems: effects of atmospheric pressure.

    Science.gov (United States)

    Rygalov, Vadim Y; Fowler, Philip A; Metz, Joannah M; Wheeler, Raymond M; Bucklin, Ray A

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from ~1 to 10 L m-2 d-1 (~1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  16. Water cycles in closed ecological systems: effects of atmospheric pressure

    Science.gov (United States)

    Rygalov, Vadim Y.; Fowler, Philip A.; Metz, Joannah M.; Wheeler, Raymond M.; Bucklin, Ray A.; Sager, J. C. (Principal Investigator)

    2002-01-01

    In bioregenerative life support systems that use plants to generate food and oxygen, the largest mass flux between the plants and their surrounding environment will be water. This water cycle is a consequence of the continuous change of state (evaporation-condensation) from liquid to gas through the process of transpiration and the need to transfer heat (cool) and dehumidify the plant growth chamber. Evapotranspiration rates for full plant canopies can range from 1 to 10 L m-2 d-1 (1 to 10 mm m-2 d-1), with the rates depending primarily on the vapor pressure deficit (VPD) between the leaves and the air inside the plant growth chamber. VPD in turn is dependent on the air temperature, leaf temperature, and current value of relative humidity (RH). Concepts for developing closed plant growth systems, such as greenhouses for Mars, have been discussed for many years and the feasibility of such systems will depend on the overall system costs and reliability. One approach for reducing system costs would be to reduce the operating pressure within the greenhouse to reduce structural mass and gas leakage. But managing plant growth environments at low pressures (e.g., controlling humidity and heat exchange) may be difficult, and the effects of low-pressure environments on plant growth and system water cycling need further study. We present experimental evidence to show that water saturation pressures in air under isothermal conditions are only slightly affected by total pressure, but the overall water flux from evaporating surfaces can increase as pressure decreases. Mathematical models describing these observations are presented, along with discussion of the importance for considering "water cycles" in closed bioregenerative life support systems.

  17. Airfoil Trailing Edge Noise Generation and Its Surface Pressure Fluctuation

    DEFF Research Database (Denmark)

    Zhu, Wei Jun; Shen, Wen Zhong

    2015-01-01

    In the present work, Large Eddy Simulation (LES) of turbulent flows over a NACA 0015 airfoil is performed. The purpose of such numerical study is to relate the aerodynamic surface pressure with the noise generation. The results from LES are validated against detailed surface pressure measurements...... where the time history pressure data are recorded by the surface pressure microphones. After the flow-field is stabilized, the generated noise from the airfoil Trailing Edge (TE) is predicted using the acoustic analogy solver, where the results from LES are the input. It is found that there is a strong...

  18. Conclusions from the sodium-water reaction experiments performed with a straight tube bundle model for a steam generator with respect to the calculation method of the accident design pressure

    International Nuclear Information System (INIS)

    Ludwig, P.W.P.; Ruijterman, C.

    1975-01-01

    This paper will give some conclusions, drawn from the big-leakages sodium-water reaction experiments, on the calculation methods to be used in determining the sodium blow down, the water supply and the bubble pressure. The necessity of taking into account the compressibility of sodium is demonstrated. (author)

  19. Conclusions from the sodium-water reaction experiments performed with a straight tube bundle model for a steam generator with respect to the calculation method of the accident design pressure

    Energy Technology Data Exchange (ETDEWEB)

    Ludwig, P W.P.; Ruijterman, C

    1975-07-01

    This paper will give some conclusions, drawn from the big-leakages sodium-water reaction experiments, on the calculation methods to be used in determining the sodium blow down, the water supply and the bubble pressure. The necessity of taking into account the compressibility of sodium is demonstrated. (author)

  20. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C.

    1997-01-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  1. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  2. Steam Generator Owners Group PWR secondary water chemistry guidelines

    International Nuclear Information System (INIS)

    Welty, C.S. Jr.; Green, S.J.

    1985-01-01

    In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

  3. High-Pressure Oxygen Generation for Outpost EVA Study

    Science.gov (United States)

    Jeng, Frank F.; Conger, Bruce; Ewert, Michael K.; Anderson, Molly S.

    2009-01-01

    The amount of oxygen consumption for crew extravehicular activity (EVA) in future lunar exploration missions will be significant. Eight technologies to provide high pressure EVA O2 were investigated. They are: high pressure O2 storage, liquid oxygen (LOX) storage followed by vaporization, scavenging LOX from Lander followed by vaporization, LOX delivery followed by sorption compression, water electrolysis followed by compression, stand-alone high pressure water electrolyzer, Environmental Control and Life Support System (ECLSS) and Power Elements sharing a high pressure water electrolyzer, and ECLSS and In-Situ Resource Utilization (ISRU) Elements sharing a high pressure electrolyzer. A trade analysis was conducted comparing launch mass and equivalent system mass (ESM) of the eight technologies in open and closed ECLSS architectures. Technologies considered appropriate for the two architectures were selected and suggested for development.

  4. Water waves generated by underwater explosion

    CERN Document Server

    Mehaute, Bernard Le

    1996-01-01

    This is the first book on explosion-generated water waves. It presents the theoretical foundations and experimental results of the generation and propagation of impulsively generated waves resulting from underwater explosions. Many of the theories and concepts presented herein are applicable to other types of water waves, in particular, tsunamis and waves generated by the fall of a meteorite. Linear and nonlinear theories, as well as experimental calibrations, are presented for cases of deep and shallow water explosions. Propagation of transient waves on dissipative, nonuniform bathymetries to

  5. Tritium issues in commercial pressurized water reactors

    International Nuclear Information System (INIS)

    Jones, G.

    2008-01-01

    Tritium has become an important radionuclide in commercial Pressurized Water Reactors because of its mobility and tendency to concentrate in plant systems as tritiated water during the recycling of reactor coolant. Small quantities of tritium are released in routine regulated effluents as liquid water and as water vapor. Tritium has become a focus of attention at commercial nuclear power plants in recent years due to inadvertent, low-level, chronic releases arising from routine maintenance operations and from component failures. Tritium has been observed in groundwater in the vicinity of stations. The nuclear industry has undertaken strong proactive corrective measures to prevent recurrence, and continues to eliminate emission sources through its singular focus on public safety and environmental stewardship. This paper will discuss: production mechanisms for tritium, transport mechanisms from the reactor through plant, systems to the environment, examples of routine effluent releases, offsite doses, basic groundwater transport and geological issues, and recent nuclear industry environmental and legal ramifications. (authors)

  6. Pumps for German pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dernedde, R.

    1984-01-01

    The article describes the development of a selection of pumps which are used in the primary coolant system and the high-pressure safety injection system and feed water system during the past 2 decades. The modifications were caused by the step-wise increasing power output of the plants from 300 MW up to 1300 MW. Additional important influences were given be the increased requirements for quality assurance and final-documentation. The good operating results of the delivered pumps proved that the reliability is independent of the volume of the software-package. The outlook expects that consolidation will be followed by additional steps for the order processing of components for the convoy pumps. KW: main coolant pump; primary system; boiler feed pump; reactor pump; secondary system; barrel insert pump; pressure water reactor; convoy pump; state of the art.

  7. Chloride Ingress into Concrete under Water Pressure

    DEFF Research Database (Denmark)

    Lund, Mia Schou; Sander, Lotte Braad; Grelk, Bent

    2011-01-01

    The chloride ingress into concrete under water pressures of 100 kPa and 800 kPa have been investigated by experiments. The specimens were exposed to a 10% NaCl solution and water mixture. For the concrete having w/c = 0.35 the experimental results show the chloride diffusion coefficient at 800 k......Pa (~8 atm.) is 12 times greater than at 100 kPa (~1 atm.). For w/c = 0.45 and w/c = 0.55 the chloride diffusion coefficients are 7 and 3 times greater. This means that a change in pressure highly influences the chloride ingress into the concrete and thereby the life length models for concrete structures....

  8. Integral Pressurized Water Reactor Simulator Manual

    International Nuclear Information System (INIS)

    2017-01-01

    This publication provides detailed explanations of the theoretical concepts that the simulator users have to know to gain a comprehensive understanding of the physics and technology of integral pressurized water reactors. It provides explanations of each of the simulator screens and various controls that a user can monitor and modify. A complete description of all the simulator features is also provided. A detailed set of exercises is provided in the Exercise Handbook accompanying this publication.

  9. Modeling Diesel engine combustion using pressure dependent Flamelet Generated Manifolds

    NARCIS (Netherlands)

    Bekdemir, C.; Somers, L.M.T.; Goey, de L.P.H.

    2011-01-01

    Flamelet Generated Manifolds (FGMs) are constructed and applied to simulations of a conventional compression ignition engine cycle. To study the influence of pressure and temperature variations on the ignition process after the compression stroke, FGMs with several pressure levels are created. These

  10. Membrane-based processes for sustainable power generation using water

    KAUST Repository

    Logan, Bruce E.; Elimelech, Menachem

    2012-01-01

    Water has always been crucial to combustion and hydroelectric processes, but it could become the source of power in membrane-based systems that capture energy from natural and waste waters. Two processes are emerging as sustainable methods for capturing energy from sea water: pressure-retarded osmosis and reverse electrodialysis. These processes can also capture energy from waste heat by generating artificial salinity gradients using synthetic solutions, such as thermolytic salts. A further source of energy comes from organic matter in waste waters, which can be harnessed using microbial fuel-cell technology, allowing both wastewater treatment and power production. © 2012 Macmillan Publishers Limited. All rights reserved.

  11. Membrane-based processes for sustainable power generation using water

    KAUST Repository

    Logan, Bruce E.

    2012-08-15

    Water has always been crucial to combustion and hydroelectric processes, but it could become the source of power in membrane-based systems that capture energy from natural and waste waters. Two processes are emerging as sustainable methods for capturing energy from sea water: pressure-retarded osmosis and reverse electrodialysis. These processes can also capture energy from waste heat by generating artificial salinity gradients using synthetic solutions, such as thermolytic salts. A further source of energy comes from organic matter in waste waters, which can be harnessed using microbial fuel-cell technology, allowing both wastewater treatment and power production. © 2012 Macmillan Publishers Limited. All rights reserved.

  12. The Westinghouse Advanced Passive Pressurized Water Reactor, AP1000

    International Nuclear Information System (INIS)

    Schene, R.

    2009-01-01

    Featuring proven technology and innovative passive safety systems, the Westinghouse AP1000 pressurized water reactor can achieve competitive generation costs in the current electricity market without emitting harmful greenhouse gases and further harming the environment. Westinghouse Electric Company, the pioneer in nuclear energy once again sets a new industry standard with the AP1000. The AP1000 is a two-loop pressurized water reactor that uses simplified, innovative and effective approach to safety. With a gross power rating of 3415 megawatt thermal and a nominal net electrical output of 1117 megawatt electric, the AP1000 is ideal for new base load generation. The AP1000 is the safest and most economical nuclear power plant available in the worldwide commercial marketplace, and is the only Generation III+ reactor to receive a design certification from the U.S. Nuclear Regulatory Commission (NRC). Based on nearly 20 years of research and development, the AP1000 builds and improves upon the established technology of major components used in current Westinghouse designed plants. These components, including steam generators, digital instrumentation and controls, fuel, pressurizers, and reactor vessels, are currently in use around the world and have years of proven, reliable operating experience. Historically, Westinghouse plant designs and technology have forged the cutting edge technology of nuclear plant around the world. Today, nearly 50 percent of the world's 440 nuclear plants are based on Westinghouse technology. Westinghouse continues to be the nuclear industry's global leader. (author)

  13. A single-stage high pressure steam injector for next generation reactors: test results and analysis

    International Nuclear Information System (INIS)

    Cattadori, G.; Galbiati, L.; Mazzocchi, L.; Vanini, P.

    1995-01-01

    Steam injectors can be used in advanced light water reactors (ALWRs) for high pressure makeup water supply; this solution seems to be very attractive because of the ''passive'' features of steam injectors, that would take advantage of the available energy from primary steam without the introduction of any rotating machinery. The reference application considered in this work is a high pressure safety injection system for a BWR; a water flow rate of about 60 kg/s to be delivered against primary pressures covering a quite wide range up to 9 MPa is required. Nevertheless, steam driven water injectors with similar characteristics could be used to satisfy the high pressure core coolant makeup requirements of next generation PWRs. With regard to BWR application, an instrumented steam injector prototype with a flow rate scaling factor of about 1:6 has been built and tested. The tested steam injector operates at a constant inlet water pressure (about 0.2 MPa) and inlet water temperature ranging from 15 to 37 o C, with steam pressure ranging from 2.5 to 8.7 MPa, always fulfilling the discharge pressure target (10% higher than steam pressure). To achieve these results an original double-overflow flow rate-control/startup system has been developed. (Author)

  14. Cooling of pressurized water nuclear reactor vessels

    International Nuclear Information System (INIS)

    Curet, H.D.

    1978-01-01

    The improvement of pressurized water nuclear reactor vessels comprising flow dividers providing separate and distinct passages for the flow of core coolant water from each coolant water inlet, the flow dividers being vertically disposed in the annular flow areas provided by the walls of the vessel, the thermal shield (if present), and the core barrel is described. In the event of rupture of one of the coolant water inlet lines, water, especially emergency core coolant water, in the intact lines is thus prevented from by-passing the core by circumferential flow around the outermost surface of the core barrel and is instead directed so as to flow vertically downward through the annulus area between the vessel wall and the core barrel in a more normal manner to increase the probability of cooling of the core by the available cooling water in the lower plenum, thus preventing or delaying thermal damage to the core, and providing time for other appropriate remedial or damage preventing action by the operator

  15. Water-stabilized plasma generators

    Czech Academy of Sciences Publication Activity Database

    Hrabovský, Milan

    1998-01-01

    Roč. 70, č. 6 (1998), s. 1157-1162 ISSN 0033-4545 R&D Projects: GA ČR GA102/95/0592; GA ČR GV106/96/K245 Institutional research plan: CEZ:AV0Z2043910 Keywords : thermal plasma, plasma torch, water-stabilized plasma Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 1.677, year: 1998

  16. Recent developments in high pressure water technology

    International Nuclear Information System (INIS)

    Johnson, N.A.; Johnson, T.

    1992-01-01

    High Pressure Water Jetting has advanced rapidly in the last decade to a point where the field is splitting into specialised areas. This has left the end user or client in the dark as to whether water jetting will work and if so what equipment is best suited to their particular application. The aim of this paper is to give an overview of:-1. The way water is delivered to the surface and the parameters which control the concentration of energy available on impact. 2. The factors governing application driven selection of equipment. 3. The effects to technical advances in pumps and delivery systems on equipment selection with reference to their to their application to concrete removal and nuclear decontamination. (Author)

  17. Terapascal static pressure generation with ultrahigh yield strength nanodiamond.

    Science.gov (United States)

    Dubrovinskaia, Natalia; Dubrovinsky, Leonid; Solopova, Natalia A; Abakumov, Artem; Turner, Stuart; Hanfland, Michael; Bykova, Elena; Bykov, Maxim; Prescher, Clemens; Prakapenka, Vitali B; Petitgirard, Sylvain; Chuvashova, Irina; Gasharova, Biliana; Mathis, Yves-Laurent; Ershov, Petr; Snigireva, Irina; Snigirev, Anatoly

    2016-07-01

    Studies of materials' properties at high and ultrahigh pressures lead to discoveries of unique physical and chemical phenomena and a deeper understanding of matter. In high-pressure research, an achievable static pressure limit is imposed by the strength of available strong materials and design of high-pressure devices. Using a high-pressure and high-temperature technique, we synthesized optically transparent microballs of bulk nanocrystalline diamond, which were found to have an exceptional yield strength (~460 GPa at a confining pressure of ~70 GPa) due to the unique microstructure of bulk nanocrystalline diamond. We used the nanodiamond balls in a double-stage diamond anvil cell high-pressure device that allowed us to generate static pressures beyond 1 TPa, as demonstrated by synchrotron x-ray diffraction. Outstanding mechanical properties (strain-dependent elasticity, very high hardness, and unprecedented yield strength) make the nanodiamond balls a unique device for ultrahigh static pressure generation. Structurally isotropic, homogeneous, and made of a low-Z material, they are promising in the field of x-ray optical applications.

  18. High pressure gas driven liquid metal MHD homopolar generator

    International Nuclear Information System (INIS)

    Itoh, Yasuyuki

    1988-01-01

    A liquid metal MHD homopolar generator is proposed to be used as a high repetition rate pulsed power supply. In the generator, the thermal energy stored in a high pressure gas (He) reservoir is rapidly converted into kinetic energy of a rotating liquid metal (NaK) cylinder which is contracted by a gas driven annular free piston. The rotational kinetic energy is converted into electrical energy by making use of the homopolar generator principle. The conversion efficiency is calculated to be 47% in generating electrical energy of 20 kJ/pulse (1.7 MW peak power) at a repetition rate of 7 Hz. From the viewpoint of energy storage, the high pressure gas reservoir with a charging pressure of 15 MPa is considered to ''electrically'' store the energy at a density of 10 MJ/m 3 . (author)

  19. STUDY ON OPERATING CHARACTERISTECS OF WATER HAMMER GENERATING DEVICE FOR TREATMENT OF MICROORGANISM

    Science.gov (United States)

    Hamada, Tatsuhisa; Endo, Shigekatsu; Oda, Akira; Shimizu, Yasushi

    The phenomenon that has been actualized due to the water quality deterioration because of the inflow of drainage and the industrial wastewater includes the phenomenon that is called water-bloom generated in the freshwater environment made a eutrophic. This is becoming a serious problem to secure the water.Mixing with the drinking water has already been confirmed, and the generation of water-bloom is becoming a big social problem, and fundamental measures have not been established yet.On the other hand, authors are proving the pressure to be a fast the pressure speed of the impact pressure by the water hammer and effective in the destruction of the blue-green algae.In this studies, the hydraulics flow characteristics of an effective water hammer pressure generator to the shredding of the water-bloom cell were examined.As a result, there was a boundary in the region where the water hammer was generated by length and the water supply head of conduit, and the water hammer pressure was able to be understood to be influenced according to the angle of the valve that generated the water hammer in addition in the water hammer generator.The demonstration in the locale was confirmed based on these and the scale etc. of an effective device to the doing water-bloom processing were able to be confirmed by continuous running.

  20. Kinetics and dynamics of nanosecond streamer discharge in atmospheric-pressure gas bubble suspended in distilled water under saturated vapor pressure conditions

    KAUST Repository

    Sharma, Ashish; Levko, Dmitry; Raja, Laxminarayan L; Cha, Min

    2016-01-01

    We perform computational studies of nanosecond streamer discharges generated in helium bubbles immersed in distilled water under atmospheric pressure conditions. The model takes into account the presence of water vapor in the gas bubble

  1. Automated corrosion fatigue crack growth testing in pressurized water environments

    International Nuclear Information System (INIS)

    Ceschini, L.J.; Liaw, P.K.; Rudd, G.E.; Logsdon, W.A.

    1984-01-01

    This paper describes in detail a novel approach to construct a test facility for developing corrosion fatigue crack growth rate (FCGR) properties in aggressive environments. The environment studied is that of a pressurized water reactor (PWR) at 288 0 C (550 0 F) and 13.8 MPa (200 psig). To expedite data generation, each chamber was designed to accommodate two test specimens. A common water recirculation and pressurization system was employed to service two test chambers. Thus, four fatigue crack propagation rate tests could be conducted simultaneously in the pressurized water environment. The data analysis was automated to minimize the typically high labor costs associated with corrosion fatigue crack propagation testing. Verification FCGR tests conducted on an ASTM A469 rotor steel in a room temperature air environment as well as actual PWR environment FCGR tests performed on an ASTM A533 Grade B Class 2 pressure vessel steel demonstrated that the dual specimen test facility is an excellent system for developing the FCGR properties of materials in adverse environments

  2. Pulsed high-pressure (PHP) drain-down of steam generating system

    International Nuclear Information System (INIS)

    Petrusek, R.A.

    1991-01-01

    This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one vertical-type steam generator in which the upper inverted U-shaped ends of the tubes are closed and the lower ends thereof are open, the steam generator having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being part of a nuclear-powered steam generating system wherein the reactor-coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator, and the reactor being in communication with pressurizer means and comprising the steps of introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tubesheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator while permitting the water to flow out from the open ends of the U-tubes, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step

  3. Startup and commissioning of pressurized water reactors

    International Nuclear Information System (INIS)

    Albert, L.J.; Gilbert, C.F.

    1983-05-01

    A critical phase of plant development is the test, startup, and commissioning period. The effort expended prior to commissioning has a definite effect on the reliability and continuing availability of the plant during its life. This paper describes a test, startup, and commissioning program for a pressurized water reactor (PWR) plant. This program commences with the completion of construction and continues through the turnover of equipment/systems to the owner's startup/ commissioning group. The paper addresses the organization of the test/startup group, planning and scheduling, test procedures and initial testing, staffing and certification of the test group, training of operators, and turnover to the owner

  4. Thermohydraulic analysis of pressurized water reactors

    International Nuclear Information System (INIS)

    Veloso, M.A.

    1980-01-01

    The computer program PANTERA is applied in the thermo-hydraulic analysis of Pressurized Water Reactor Cores (PWR). It is a version of COBRA-IIIC in which a new thermal conduction model for fuel rods was introduced. The results calculated by this program are compared with experimental data obtained from bundles of fuel rods, simulating reactor conditions. The validity of the new thermal model is checked too. The PANTERA code, through a simplified procedure of calculation, is used in the thermo-hydraulic analysis of Indian Point, Unit 2, reactor core, in stationary conditions. The results are discussed and compared with design data. (Autor) [pt

  5. Steam plant for pressurized water reactors

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

  6. Pulsed water jet generated by pulse multiplication

    Czech Academy of Sciences Publication Activity Database

    Dvorský, R.; Sitek, Libor; Sochor, T.

    2016-01-01

    Roč. 23, č. 4 (2016), s. 959-967 ISSN 1330-3651 R&D Projects: GA MŠk(CZ) LO1406; GA MŠk ED2.1.00/03.0082 Institutional support: RVO:68145535 Keywords : high- pressure pulses * pulse intensifier * pulsed water jet * water hammer effect Subject RIV: JQ - Machines ; Tools Impact factor: 0.723, year: 2016 http://hrcak.srce.hr/163752?lang=en

  7. MEDEA, Steady-State Pressure and Temperature Distribution in He H2O Steam Generator

    International Nuclear Information System (INIS)

    Hansen, Ulf

    1976-01-01

    1 - Nature of physical problem solved: MEDEA calculates the time-independent pressure and temperature distribution in a helium-water steam generator. The changing material properties of the fluids with pressure and temperature are treated exactly. The steam generator may consist of economizer, evaporator, superheater and reheater in variable flow patterns. In case of reheating the high-pressure turbine is taken into account. The main control circuits influencing the behaviour of the system are simulated. These are water spraying of the hot steam, load-dependent control of steam pressure at the HP-turbine inlet and valves before the LP-turbine to ensure constant pressure in the reheater section. Investigations of hydrodynamic flow stability in single tubes can be performed. 2 - Method of solution: The steam generator is calculated as a 1-dimensional model, (i.e. all parallel tubes working under equal conditions) and is divided into small heat exchanger elements with helium and water in ideal parallel or counter flow. The material and thermodynamic properties are kept constant within one element. The calculations start at the cold end of the steam generator and proceed stepwise along the water flow pattern to produce pressure and temperature distributions of helium and water. The gas outlet temperature is changed until convergence is reached with a continuous temperature profile on the gas side. MEDEA chooses the iteration scheme according to flow pattern and other special arrangements in the steam generator. The hydrodynamic stability is calculated for a single tube assuming that all tubes are exposed to the same gas temperature profile and changing the water flow in a single tube will not influence the conditions on the gas side. Varying the water flow by keeping gas temperature constant and repeating the steam generator calculations yield pressure drop and steam temperature as a function of flow rate. 3 - Restrictions on the complexity of the problem: Maximum

  8. Shockwave generation by a semiconductor bridge operation in water

    Energy Technology Data Exchange (ETDEWEB)

    Zvulun, E.; Toker, G.; Gurovich, V. Tz.; Krasik, Ya. E. [Physics Department, Technion, Haifa 32000 (Israel)

    2014-05-28

    A semiconductor bridge (SCB) is a silicon device, used in explosive systems as the electrical initiator element. In recent years, SCB plasma has been extensively studied, both electrically and using fast photography and spectroscopic imaging. However, the value of the pressure buildup at the bridge remains unknown. In this study, we operated SCB devices in water and, using shadow imaging and reference beam interferometry, obtained the velocity of the shock wave propagation and distribution of the density of water. These results, together with a self-similar hydrodynamic model, were used to calculate the pressure generated by the exploding SCB. In addition, the results obtained showed that the energy of the water flow exceeds significantly the energy deposited into the exploded SCB. The latter can be explained by the combustion of the aluminum and silicon atoms released in water, which acts as an oxidizing medium.

  9. A method for generating hydrogen from water

    International Nuclear Information System (INIS)

    Godin, Paul; Mascarello, Jean; Millet, Jacques.

    1974-01-01

    Description is given of a method and an installation for generating hydrogen from water, through an endothermic cycle of several successive chemical reactions involving intermediate substances regenerated during said cycle, said reactions occuring at different temperatures. The reaction which takes place at the highest temperature is carried out electrochemically. This can be applied to power-generating units comprising a nuclear reactor [fr

  10. Entropy Generation in Steady Laminar Boundary Layers with Pressure Gradients

    Directory of Open Access Journals (Sweden)

    Donald M. McEligot

    2014-07-01

    Full Text Available In an earlier paper in Entropy [1] we hypothesized that the entropy generation rate is the driving force for boundary layer transition from laminar to turbulent flow. Subsequently, with our colleagues we have examined the prediction of entropy generation during such transitions [2,3]. We found that reasonable predictions for engineering purposes could be obtained for flows with negligible streamwise pressure gradients by adapting the linear combination model of Emmons [4]. A question then arises—will the Emmons approach be useful for boundary layer transition with significant streamwise pressure gradients as by Nolan and Zaki [5]. In our implementation the intermittency is calculated by comparison to skin friction correlations for laminar and turbulent boundary layers and is then applied with comparable correlations for the energy dissipation coefficient (i.e., non-dimensional integral entropy generation rate. In the case of negligible pressure gradients the Blasius theory provides the necessary laminar correlations.

  11. Electron beam generation in the fore-vacuum pressure range

    CERN Document Server

    Burachevskij, Y A; Kuzemchenko, M N; Mytnikov, A V; Oks, E M

    2001-01-01

    One presents the results of investigations to generate electron beams within 0.01-0.1 Torr gas pressure range. To generate a beam one used a plasma source based on a hollow cathode discharge in combination with a plane accelerating gap. Peculiar features of electron emission and acceleration within the mentioned pressure range are associated with high probability of gas ionization in an accelerating gap and with generation of ion flow meeting electron beam. It results in reduction of discharge combustion intensification, as well as, in plasma concentration range. The developed design of an electron source enables to generate cylindrical beams with up to 1 A current and with up to 10 keV energy

  12. The temperature control and water quality regulation for steam generator secondary side hydrostatic test

    International Nuclear Information System (INIS)

    Xiao Bo; Liu Dongyong

    2014-01-01

    The secondary side hydrostatic test for the steam generator of M310 unit is to verify the pressure tightness of steam generator secondary side tube sheet and related systems. As for the importance of the steam generator, the water temperature and water quality of hydrostatic test has strict requirements. The discussion on the water temperature control and water quality regulation for the secondary loop hydrostatic test of Fuqing Unit 1 contribute greatly to the guiding work for the preparation of the steam generator pressure test for M310 unit. (authors)

  13. Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

    2015-01-01

    Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

  14. Radiation pressure induced difference-sideband generation beyond linearized description

    OpenAIRE

    Xiong, Hao; Fan, Y. W.; Yang, X.; Wu, Y.

    2016-01-01

    We investigate radiation-pressure induced generation of the frequency components at the difference-sideband in an optomechanical system, which beyond the conventional linearized description of optomechanical interactions between cavity fields and the mechanical oscillation. We analytically calculate amplitudes of these signals, and identify a simple square-root law for both the upper and lower difference-sideband generation which can describe the dependence of the intensities of these signals...

  15. Generating Electricity from Water through Carbon Nanomaterials.

    Science.gov (United States)

    Xu, Yifan; Chen, Peining; Peng, Huisheng

    2018-01-09

    Over the past ten years, electricity generation from water in carbon-based materials has aroused increasing interest. Water-induced mechanical-to-electrical conversion has been discovered in carbon nanomaterials, including carbon nanotubes and graphene, through the interaction with flowing water as well as moisture. In this Concept article, we focus on the basic principles of electric energy harvesting from flowing water through carbon nanomaterials, and summarize the material modification and structural design of these nanogenerators. The current challenges and potential applications of power conversion with carbon nanomaterials are finally highlighted. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  16. Pressurized water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23 'Boiling Water Reactor Simulator' (2003). This report consists of course material for workshops using a pressurized water reactor simulator

  17. Natural Circulation Characteristics of an Integral Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Junli Gou; Suizheng Qiu; Guanghui Su; Dounan Jia

    2006-01-01

    Natural circulation potential is of great importance to the inherent safety of a nuclear reactor. This paper presents a theoretical investigation on the natural circulation characteristics of an integrated pressurized water reactor. Through numerically solved the one-dimensional model, the steady-state single phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the once-through steam generator, the natural circulation characteristics are studied. Based on the preliminary calculation analysis, it is found that natural circulation mass flow rate is proportional to the exponential function of the power, and the value of the exponent is related to working conditions of the steam generator secondary side. The higher height difference between the core center and the steam generator center is favorable to the heat removal capacity of the natural circulation. (authors)

  18. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  19. Pressurized water reactors: the EPR project

    International Nuclear Information System (INIS)

    Py, J.P.; Yvon, M.

    2007-01-01

    EPR (originally 'European pressurized water reactor', and now 'evolutionary power reactor') is a model of reactor initially jointly developed by French and German engineers which fulfills the particular safety specifications of both countries but also the European utility requirements jointly elaborated by the main European power companies under the initiative of Electricite de France (EdF). Today, two EPR-based reactors are under development: one is under construction in Finland and the other, Flamanville 3 (France), received its creation permit decree on April 10, 2007. This article presents, first, the main objectives of the EPR, and then, describes the Flamanville 3 reactor: reactor type and general conditions, core and conditions of operation, primary and secondary circuits with their components, main auxiliary and recovery systems, man-machine interface and instrumentation and control system, confinement and serious accidents, arrangement of buildings. (J.S.)

  20. Status of advanced small pressurized water reactors

    International Nuclear Information System (INIS)

    Chen Peipei; Zhou Yun

    2012-01-01

    In order to expand the nuclear power in energy and desalination, increase competitiveness in global nuclear power market, many developed countries with strong nuclear energy technology have realized the importance of Small Modular Reactor (SMR) and initiated heavy R and D programs in SMR. The Advanced Small Pressurized Water Reactor (ASPWR) is characterized by great advantages in safety and economy and can be used in remote power grid and replace mid/small size fossil plant economically. This paper reviews the history and current status of SMR and ASPWR, and also discusses the design concept, safety features and other advantages of ASPWR. The purpose of this paper is to provide an overall review of ASPWR technology in western countries, and to promote the R and D in ASPWR in China. (authors)

  1. Operator Support System for Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Wei Renjie; Shen Shifei

    1996-01-01

    Operator Support System for Pressurized Water Reactor (OSSPWR) has been developed under the sponsorship of IAEA from August 1994. The project is being carried out by the Department of Engineering Physics, Tsinghua University, Beijing, China. The Design concepts of the operator support functions have been established. The prototype systems of OSSPWR has been developed as well. The primary goal of the project is to create an advanced operator support system by applying new technologies such as artificial intelligence (AI) techniques, advanced communication technologies, etc. Recently, the advanced man-machine interface for nuclear power plant operators has been developed. It is connected to the modern computer systems and utilizes new high performance graphic displays. (author). 6 refs, 4 figs

  2. The safety of pressurized water reactors

    International Nuclear Information System (INIS)

    Panossian, J.; Tanguy, P.

    1991-01-01

    In this paper we present a review of the status of the safety level of modern pressurized water reactors, that is to say those that meet the safety criteria accepted today by the international nuclear community. We will mainly rely on the operating experience and the Probabilistic Safety Assessments concerning French reactors. We will not back over the basic safety concepts of these reactors, which are well known. We begin with a brief review of some of the lessons learned from the two main accidents discussed in the present meeting. Three Mile Island and Chernobyl, without entering into details presented in previous papers. The presentation ends with a rather lengthy conclusion, aimed more at those not directly involved in the technical details of nuclear safety matters

  3. Pressure equalization systems in pressurized water reactor fuel rods

    International Nuclear Information System (INIS)

    Steven, J.; Wunderlich, F.

    1979-01-01

    For the development of a pressure reduction system in PWR fuel rods the capability of charcoal to adsorb Helium, Xenon and Krypton at temperatures of about 300 0 C was investigated. The influence of the adsorption on fuel rod internal pressure and in creep strain on the tube was evaluated in a design study. (orig.) [de

  4. Validation of the dynamic model for a pressurized water reactor

    International Nuclear Information System (INIS)

    Zwingelstein, Gilles.

    1979-01-01

    Dynamic model validation is a necessary procedure to assure that the developed empirical or physical models are satisfactorily representing the dynamic behavior of the actual plant during normal or abnormal transients. For small transients, physical models which represent isolated core, isolated steam generator and the overall pressurized water reactor are described. Using data collected during the step power changes that occured during the startup procedures, comparisons of experimental and actual transients are given at 30% and 100% of full power. The agreement between the transients derived from the model and those recorded on the plant indicates that the developed models are well suited for use for functional or control studies

  5. State of the art of the advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Seifritz, W.; Chawla, R.

    1987-01-01

    A review is given of the present status of the works concerned with an advanced pressurized water reactor (APWR). It includes the following items: reactor physics, thermal and hydraulic investigations and other engineering aspects as well as an analysis of electricity generation cost and long-term problems of embedding the APWR in a plutonium economy. As a summary it can be stated that there are discernible no principal obstacles of technically accomplishing an APWR, but there will be necessary considerable expenses in research and development works if it should be intended to start commercial service of an APWR up to the end of this century. (author)

  6. Analytical treatment of large leak pressure behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Hori, Masao; Miyake, Osamu

    1980-07-01

    Simplified analytical methods applicable to the estimation of initial pressure spike in case of a large leak accident in LMFBR steam generators were devised as follows; (i) Estimation of the initial water leak rate by the centered rarefaction wave method, (ii) Estimation of the initial pressure spike by the one-dimensional compressible method with either the columnar bubble growth model or the spherical bubble growth model. These methods were compared with relevant experimental data or other more elaborate analyses and validated to be usable in simple geometry and limited time span cases. Application of these methods to an actual steam generator case was explained and demonstrated. (author)

  7. Experimental study of pressure drops through LOCA-generated debris deposited on a fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jeong Kwan, E-mail: jksuh@khnp.co.kr [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Kim, Jae Won; Kwon, Sun Guk; Lee, Jae Yong [KHNP Central Research Institute, 1312-70 Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of); Cho, Hyoung Kyu; Park, Goon Cherl [Department of Nuclear Engineering, Seoul National University, Seoul 151-742 (Korea, Republic of)

    2015-08-15

    Highlights: • In-vessel downstream effect tests were performed in the presence of LOCA-generated debris. • Available driving heads under each LOCA scenario were verified using experimental data. • Fibrous debris was prepared to satisfy the length distribution obtained from the bypass test. • Limiting test conditions were identified through sensitivity studies. - Abstract: Under post loss-of-coolant accident (LOCA) conditions, it is postulated that debris can be generated and transported to the containment sump strainer. Some of the debris may pass through the strainer and could challenge the long-term core cooling capability of the plant. To address this safety issue, in-vessel downstream effect tests for the advanced power reactor (APR) 1400 were performed. Fibrous debris is the most crucial material in terms of causing pressure drops, and was prepared in this study to satisfy the fiber length distribution obtained through a strainer bypass test. Sensitivity studies on pressure drops through LOCA-generated debris deposited on a fuel assembly were performed to evaluate the effects of water chemistry and fiber length distribution. The pressure drops with debris laden pure water were substantially less than those with debris laden ordinary tap water. The experiment with fiber length distribution suggested by WCAP-16793 showed lower pressure drops than those with the APR1400 specific fiber length distribution. All the experimental results showed that the pressure drops in the mock-up fuel assembly were less than the available driving head at each LOCA scenario.

  8. Fracture risk assessment for the pressurized water reactor pressure vessel under pressurized thermal shock events

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2016-01-01

    Highlight: • The PTS loading conditions consistent with the USNRC's new PTS rule are applied as the loading condition for a Taiwan domestic PWR. • The state-of-the-art PFM technique is employed to analyze a reactor pressure vessel. • Novel flaw model and embrittlement correlation are considered in the study. • The RT-based regression formula of NUREG-1874 was also utilized to evaluate the failure risks of RPV. • For slightly embrittled RPV, the SO-1 type PTSs play more important role than other types of PTS. - Abstract: The fracture risk of the pressurized water reactor pressure vessel of a Taiwan domestic nuclear power plant has been evaluated according to the technical basis of the U.S.NRC's new pressurized thermal shock (PTS) screening criteria. The ORNL's FAVOR code and the PNNL's flaw models were employed to perform the probabilistic fracture mechanics analysis associated with plant specific parameters of the domestic reactor pressure vessel. Meanwhile, the PTS thermal hydraulic and probabilistic risk assessment data analyzed from a similar nuclear power plant in the United States for establishing the new PTS rule were applied as the loading conditions. Besides, an RT-based regression formula derived by the U.S.NRC was also utilized to verify the through-wall cracking frequencies. It is found that the through-wall cracking of the analyzed reactor pressure vessel only occurs during the PTS events resulted from the stuck-open primary safety relief valves that later reclose, but with only an insignificant failure risk. The results indicate that the Taiwan domestic PWR pressure vessel has sufficient structural margin for the PTS attack until either the current license expiration dates or during the proposed extended operation periods.

  9. Water and waste water management Generation Victoria - Latrobe Valley

    Energy Technology Data Exchange (ETDEWEB)

    Longmore, G. [Hazelwood Power Corporation, VIC (Australia); Pacific Power (International) Pty. Ltd., Sydney, NSW (Australia)

    1995-12-31

    Water is a necessary resource for coal fired power plant and waste water is generated. The efficient management of water and waste water systems becomes an important operational environmental factor. This paper describes the development and implementation of a ten year water and waste water management strategy for the Latrobe Valley Group of brown coal fired power stations in Victoria. In early 1991, a team was put together of representatives from each power site to develop the strategy entitled `SECV Latrobe Valley Water and Wastewater Management Strategy`. The strategy was developed with extensive public consultation, which was a factor in protracting the process such that the final document was not promulgated until late 1992. However, the final comprehensive document endorsed and agreed by management, has since attracted favourable comment as a model of its type. (author). 2 figs.

  10. Water and waste water management Generation Victoria - Latrobe Valley

    International Nuclear Information System (INIS)

    Longmore, G.

    1995-01-01

    Water is a necessary resource for coal fired power plant and waste water is generated. The efficient management of water and waste water systems becomes an important operational environmental factor. This paper describes the development and implementation of a ten year water and waste water management strategy for the Latrobe Valley Group of brown coal fired power stations in Victoria. In early 1991, a team was put together of representatives from each power site to develop the strategy entitled 'SECV Latrobe Valley Water and Wastewater Management Strategy'. The strategy was developed with extensive public consultation, which was a factor in protracting the process such that the final document was not promulgated until late 1992. However, the final comprehensive document endorsed and agreed by management, has since attracted favourable comment as a model of its type. (author). 2 figs

  11. Limit regulation system for pressurized water nuclear reactors

    International Nuclear Information System (INIS)

    Aleite, W.; Bock, H.W.

    1976-01-01

    Described is a limit regulation system for a pressurized water nuclear reactor in combination with a steam generating system connected to a turbine, the nuclear reactor having control rods as well as an operational regulation system and a protective system, which includes reactor power limiting means operatively associated with the control rods for positioning the same and having response values between operating ranges of the operational regulation system, on the one hand, and response values of the protective system, on the other hand, and a live steam-minimal pressure regulation system cooperating with the reactor power limiting means and operatively connected to a steam inlet valve to the turbine for controlling the same

  12. Water Pressure Distribution on a Flying Boat Hull

    Science.gov (United States)

    Thompson, F L

    1931-01-01

    This is the third in a series of investigations of the water pressures on seaplane floats and hulls, and completes the present program. It consisted of determining the water pressures and accelerations on a Curtiss H-16 flying boat during landing and taxiing maneuvers in smooth and rough water.

  13. Lower limb intracast pressures generated by different types of immobilisation casts.

    Science.gov (United States)

    Chaudhury, Salma; Hazlerigg, Alexandra; Vusirikala, Anuhya; Nguyen, Joseph; Matthews, Stuart

    2017-02-18

    To determine if complete, split casts and backslabs [plaster of Paris (POP) and fiberglass] generate different intracast pressures and pain. Increased swelling within casts was modeled by a closed water system attached to an expandable bag placed directly under different types of casts applied to a healthy lower limb. Complete fiberglass and POP casts, split casts and backslabs were applied. Twenty-five milliliter aliquots of saline were injected into the system and the generated intracast pressures were measured using a sphygmomanometer. The subject was blinded to the pressure scores to avoid bias. All casts were applied to the same right limb on the same subject to avoid the effects of variations in anatomy or physiology on intracast pressures. Pain levels were evaluated using the Visual Analogue Score after each sequential saline injection. Each type of cast was reapplied four times and the measurements were repeated on four separate occasions. Sample sizes were determined by a pre-study 90% power calculation to detect a 20% difference in intracast pressures between cast groups. A significant difference between the various types of casts was noted when the saline volume was greater than 100 mL ( P = 0.009). The greatest intracast pressure was generated by complete fiberglass casts, which were significantly higher than complete POP casts or backslabs ( P = 0.018 and P = 0.008 respectively) at intracast saline volumes of 100 mL and higher. Backslabs produced a significantly lower intracast pressure compared to complete POP only once the saline volume within casts exceeded 225 mL ( P = 0.009). Intracast pressures were significantly lower in split casts ( P = 0.003). Split POP and fiberglass casts produced the lowest intracast pressures, even compared to backslabs ( P = 0.009). Complete fiberglass casts generated the highest pain levels at manometer pressures of 75 mmHg and greater ( P = 0.001). Split fiberglass casts had significantly reduced pain levels ( P = 0

  14. Decomposition of water-insoluble organic waste by water plasma at atmospheric pressure

    International Nuclear Information System (INIS)

    Choi, S; Watanabe, T

    2012-01-01

    The water plasma was generated in atmospheric pressure with the emulsion state of 1-decanol which is a source of soil and ground water pollution. In order to investigate effects of operating conditions on the decomposition of 1-decanol, generated gas and liquid from the water plasma treatment were analysed in different arc current and 1-decanol concentration. The 1-decanol was completely decomposed generating hydrogen, carbon monoxide, carbon dioxide, methane, treated liquid and solid carbon in all experimental conditions. The feeding rate of 1- decanol emulsion was increased with increasing the arc current in virtue of enhanced input power. The generation rate of gas and the ratio of carbon dioxide to carbon monoxide were increased in the high arc current, while the generation rate of solid carbon was decreased due to enhanced oxygen radicals in the high input power. Generation rates of gas and solid carbon were increased at the same time with increasing the concentration of 1-decanol, because carbon radicals were increased without enhancement of oxygen radicals in a constant power level. In addition, the ratio of carbon dioxide to carbon monoxide was increased along with the concentration of 1-decanol due to enhanced carbon radicals in the water plasma flame.

  15. Excess pore water pressure induced in the foundation of a tailings dyke at Muskeg River Mine, Fort McMurray

    Energy Technology Data Exchange (ETDEWEB)

    Eshraghian, A.; Martens, S. [Klohn Crippen Berger Ltd., Calgary, AB (Canada)

    2010-07-01

    This paper discussed the effect of staged construction on the generation and dissipation of excess pore water pressure within the foundation clayey units of the External Tailings Facility dyke. Data were compiled from piezometers installed within the dyke foundation and used to estimate the dissipation parameters for the clayey units for a selected area of the foundation. Spatial and temporal variations in the pore water pressure generation parameters were explained. Understanding the process by which excess pore water pressure is generated and dissipates is critical to optimizing dyke design and performance. Piezometric data was shown to be useful in improving estimates of the construction-induced pore water pressure and dissipation rates within the clay layers in the foundation during dyke construction. In staged construction, a controlled rate of load application is used to increase foundation stability. Excess pore water pressure dissipates after each application, so the most critical stability condition happens after each load. Slow loading allows dissipation, whereas previous load pressure remains during fast loading. The dyke design must account for the rate of loading and the rate of pore pressure dissipation. Controlling the rate of loading and the rate of stress-induced excess pore water pressure generation is important to dyke stability during construction. Effective stress-strength parameters for the foundation require predictions of the pore water pressure induced during staged construction. It was found that both direct and indirect loading generates excess pore water pressure in the foundation clays. 2 refs., 2 tabs., 11 figs.

  16. Generation of runaway electron beams in high-pressure nitrogen

    Science.gov (United States)

    Tarasenko, V. F.; Burachenko, A. G.; Baksht, E. Kh

    2017-07-01

    In this paper the results of experimental studies of the amplitude-temporal characteristics of a runaway electron beam, as well as breakdown voltage in nitrogen are presented. The voltage pulses with the amplitude in incident wave ≈120 kV and the rise time of ≈0.3 ns was used. The supershort avalanche electron beam (SAEB) was detected by a collector behind the flat anode. The amplitude-time characteristics of the voltage and SAEB current were studied with subnanosecond time resolution. The maximum pressure at which a SAEB is detectable by collector was ∼1 MPa. This pressure increases with decreasing the voltage rise time. The waveforms of the discharge and runaway electron beam currents was synchronized with the voltage pulses. The mechanism of the runaway electron generation in atmospheric-pressure gases is analyzed on the basis of the obtained experimental data.

  17. Radiation pressure induced difference-sideband generation beyond linearized description

    Energy Technology Data Exchange (ETDEWEB)

    Xiong, Hao, E-mail: haoxiong1217@gmail.com; Fan, Yu-Wan; Yang, Xiaoxue; Wu, Ying, E-mail: yingwu2@126.com [School of Physics, Huazhong University of Science and Technology, Wuhan 430074 (China)

    2016-08-08

    We investigate radiation-pressure induced generation of the frequency components at the difference-sideband in an optomechanical system, which beyond the conventional linearized description of optomechanical interactions between cavity fields and the mechanical oscillation. We analytically calculate amplitudes of these signals, and identify a simple square-root law for both the upper and lower difference-sideband generation which can describe the dependence of the intensities of these signals on the pump power. Further calculation shows that difference-sideband generation can be greatly enhanced via achieving the matching conditions. The effect of difference-sideband generation, which may have potential application for manipulation of light, is especially suited for on-chip optomechanical devices, where nonlinear optomechanical interaction in the weak coupling regime is within current experimental reach.

  18. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  19. Determining Atmospheric Pressure Using a Water Barometer

    Science.gov (United States)

    Lohrengel, C. Frederick, II; Larson, Paul R.

    2012-01-01

    The atmosphere is an envelope of compressible gases that surrounds Earth. Because of its compressibility and nonuniform heating by the Sun, it is in constant motion. The atmosphere exerts pressure on Earth's surface, but that pressure is in constant flux. This experiment allows students to directly measure atmospheric pressure by measuring the…

  20. Computerized cost model for pressurized water reactors

    International Nuclear Information System (INIS)

    Meneely, T.K.; Tabata, Hiroaki; Labourey, P.

    1999-01-01

    A computerized cost model has been developed in order to allow utility users to improve their familiarity with pressurized water reactor overnight capital costs and the various factors which influence them. This model organizes its cost data in the standard format of the Energy Economic Data Base (EEDB), and encapsulates simplified relationships between physical plant design information and capital cost information in a computer code. Model calculations are initiated from a base case, which was established using traditional cost calculation techniques. The user enters a set of plant design parameters, selected to allow consideration of plant models throughout the typical three- and four-loop PWR power range, and for plant sites in Japan, Europe, and the United States. Calculation of the new capital cost is then performed in a very brief time. The presentation of the program's output allows comparison of various cases with each other or with separately calculated baseline data. The user can start at a high level summary, and by selecting values of interest on a display grid show progressively more and more detailed information, including links to background information such as individual cost driver accounts and physical plant variables for each case. Graphical presentation of the comparison summaries is provided, and the numerical results may be exported to a spreadsheet for further processing. (author)

  1. Chemistry management of generator stator water system

    International Nuclear Information System (INIS)

    Sankar, N.; Santhanam, V.S.; Ayyar, S.R.; Umapathi, P.; Jeena, P.; Hari Krishna, K.; Rajendran, D.

    2015-01-01

    Chemistry management of water cooled turbine generators with hollow copper conductors is very essential to avoid possible re-deposition of released copper oxides on stator windings, which otherwise may cause flow restrictions by partial plugging of copper hollow conductors and impair cooling. The phenomenon which is of more concern is not strictly of corrosion failure, but the consequences caused by the re-deposition of copper oxides that were formed by reaction of copper with oxygen. There were also some Operating experiences (OE) related to Copper oxide fouling in the system resulting shut down/off-line of plants. In Madras Atomic Power Station (MAPS), the turbine generator stator windings are of Copper material and cooled by demineralized water passing through the hollow conductors. The heated water from the stator is cooled by process water. A part of the stator water is continuously passed through a mixed bed polisher to remove any soluble ionic contaminants to maintain the purity of system water and also maintain copper content as low as possible to avoid possible re-deposition of released copper oxides on stator windings. The chemistry regime employed is neutral water with dissolved oxygen content between 1000-2000 ppb. Chemistry management of Stator water system was reviewed to know its effectiveness. Detailed chemical analyses of the spent resins from the polishing unit were carried out in various campaigns which indicated only part exhaustion of the polishing unit resins and reasonably low levels of copper entrapment in the resins, thus highlighting the effectiveness of the in-practice chemistry regime. (author)

  2. Thermodynamic and energy efficiency analysis of power generation from natural salinity gradients by pressure retarded osmosis.

    Science.gov (United States)

    Yip, Ngai Yin; Elimelech, Menachem

    2012-05-01

    The Gibbs free energy of mixing dissipated when fresh river water flows into the sea can be harnessed for sustainable power generation. Pressure retarded osmosis (PRO) is one of the methods proposed to generate power from natural salinity gradients. In this study, we carry out a thermodynamic and energy efficiency analysis of PRO work extraction. First, we present a reversible thermodynamic model for PRO and verify that the theoretical maximum extractable work in a reversible PRO process is identical to the Gibbs free energy of mixing. Work extraction in an irreversible constant-pressure PRO process is then examined. We derive an expression for the maximum extractable work in a constant-pressure PRO process and show that it is less than the ideal work (i.e., Gibbs free energy of mixing) due to inefficiencies intrinsic to the process. These inherent inefficiencies are attributed to (i) frictional losses required to overcome hydraulic resistance and drive water permeation and (ii) unutilized energy due to the discontinuation of water permeation when the osmotic pressure difference becomes equal to the applied hydraulic pressure. The highest extractable work in constant-pressure PRO with a seawater draw solution and river water feed solution is 0.75 kWh/m(3) while the free energy of mixing is 0.81 kWh/m(3)-a thermodynamic extraction efficiency of 91.1%. Our analysis further reveals that the operational objective to achieve high power density in a practical PRO process is inconsistent with the goal of maximum energy extraction. This study demonstrates thermodynamic and energetic approaches for PRO and offers insights on actual energy accessible for utilization in PRO power generation through salinity gradients. © 2012 American Chemical Society

  3. Evaluation of pressure transducers under turbid natural waters

    Digital Repository Service at National Institute of Oceanography (India)

    Joseph, A.; Desa, E.; Desa, E.; Smith, D.; Peshwe, V.B.; VijayKumar, K.; Desa, J.A.E.

    , the use of rho sub(eff) in contrast to the bulk density, significantly improves the measurement accuracy. For celar waters, precision density measurements made on discrete water samples agreed with rho sub(eff) values derived from pressure measurements...

  4. Electrokinetic Power Generation from Liquid Water Microjets

    Energy Technology Data Exchange (ETDEWEB)

    Duffin, Andrew M.; Saykally, Richard J.

    2008-02-15

    Although electrokinetic effects are not new, only recently have they been investigated for possible use in energy conversion devices. We have recently reported the electrokinetic generation of molecular hydrogen from rapidly flowing liquid water microjets [Duffin et al. JPCC 2007, 111, 12031]. Here, we describe the use of liquid water microjets for direct conversion of electrokinetic energy to electrical power. Previous studies of electrokinetic power production have reported low efficiencies ({approx}3%), limited by back conduction of ions at the surface and in the bulk liquid. Liquid microjets eliminate energy dissipation due to back conduction and, measuring only at the jet target, yield conversion efficiencies exceeding 10%.

  5. Pressurized water reactor with a reactor pressure vessel

    International Nuclear Information System (INIS)

    Werres, L.

    1979-01-01

    The core barrel is suspended from a flange by means of a grid. The coolant enters the barrel from below through the grid. In order to get a uniform flow over the reactor core there is provided for a guiding device below the grid. It consists of a cylindrical shell with borings uniformly distributed around the shell as well as fins on the inner surface of the shell and slots at the bottom facing the pressure vessel. (GL) [de

  6. Reactor core and passive safety systems descriptions of a next generation pressure tube reactor - mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Gaudet, M.; Rhodes, D.; Hamilton, H.; Pencer, J. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    Canada has been developing a channel-type supercritical water-cooled nuclear reactor concept, often called the Canadian SCWR. The objective of this reactor concept is to meet the technology goals of the Generation IV International Forum (GIF) for the next generation nuclear reactor development, which include enhanced safety features (inherent safe operation and deploying passive safety features), improved resource utilization, sustainable fuel cycle, and greater proliferation resistance than Generation III nuclear reactors. The Canadian SCWR core concept consists of a high-pressure inlet plenum, a separate low-pressure heavy water moderator contained in a calandria vessel, and 336 pressure tubes surrounded by the moderator. The reactor uses supercritical water as a coolant, and a direct steam power cycle to generate electricity. The reactor concept incorporates advanced safety features such as passive core cooling, long-term decay heat rejection to the environment and fuel melt prevention via passive moderator cooling. These features significantly reduce core damage frequency relative to existing nuclear reactors. This paper presents a description of the design concepts for the Canadian SCWR core, reactor building layout and the plant layout. Passive safety concepts are also described that address containment and core cooling following a loss-of coolant accident, as well as long term reactor heat removal at station blackout conditions. (author)

  7. Characteristics of fluctuating pressure generated in BWR main steam lines

    International Nuclear Information System (INIS)

    Takahashi, Shiro; Okuyama, Keita; Tamura, Akinori

    2009-01-01

    The BWR-3 steam dryer in the Quad Cities Unit 2 Nuclear Power Plant was damaged by high cycle fatigue due to acoustic-induced vibration. The dryer failure was as attributed to flow-induced acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSLs). The acoustic resonance was considered to be generated by interaction between the sound field and an unstable shear layer across the closed side branches with SRV stub pipes. We have started a research program on BWR dryers to develop their loading evaluation methods. Moreover, it has been necessary to evaluate the dryer integrity of BWR-5 plants which are the main type of BWR in Japan. In the present study, we used 1/10-scale BWR tests and analyses to investigate the flow-induced acoustic resonance and acoustic characteristics in MSLs. The test apparatus consisted of a steam dryer, a steam dome and 4 MSLs with 20 SRV stub pipes. A finite element method (FEM) was applied for the calculation of three-dimensional wave equations in acoustic analysis. We demonstrated that remarkable fluctuating pressures occurred in high and low frequency regions. High frequency fluctuating pressures was generated by the flow-induced acoustic resonance in the SRV stub pipes. Low frequency fluctuating pressure was generated in an MSL with the dead leg. The frequency of the latter almost coincided with the natural frequency of the MSL with the dead leg. The amplitude of the fluctuating pressures in the multiple stub pipes became more intense because of interaction between them compared with that in the single stub pipe. Acoustic analysis results showed that the multiple stub pipes caused several natural frequencies in the vicinity of the natural frequency of the single stub pipe and several modes of the standing wave in the MSLs. (author)

  8. Evaluation of Pressure Generated by Resistors From Different Positive Expiratory Pressure Devices.

    Science.gov (United States)

    Fagevik Olsén, Monika; Carlsson, Maria; Olsén, Erik; Westerdahl, Elisabeth

    2015-10-01

    Breathing exercises with positive expiratory pressure (PEP) are used to improve pulmonary function and airway clearance. Different PEP devices are available, but there have been no studies that describe the pressure generated by different resistors. The purpose of this study was to compare pressures generated from the proprietary resistor components of 4 commercial flow-dependent PEP valves with all other parameters kept constant. Resistors from 4 flow-regulated PEP devices (Pep/Rmt system, Wellspect HealthCare; Pipe P breathing exerciser, Koo Medical Equipment; Mini-PEP, Philips Respironics [including resistors by Rüsch]; and 15-mm endo-adapter, VBM Medizintechnik) were tested randomly by a blinded tester at constant flows of 10 and 18 L/min from an external gas system. All resistors were tested 3 times. Resistors with a similar diameter produced statistically significant different pressures at the same flow. The differences were smaller when the flow was 10 L/min compared with 18 L/min. The differences were also smaller when the diameter of the resistor was increased. The pressures produced by the 4 resistors of the same size were all significantly different when measuring 1.5- and 2.0-mm resistors at a flow of 10 L/min and 2.0-mm resistors at a flow of 18 L/min (P < .001). There were no significant differences between any of the resistors when testing sizes of 4.5 and 5.0 mm at either flow. The Mini-PEP and adapter resistors gave the highest pressures. Pressures generated by the different proprietary resistor components of 4 commercial PEP devices were not comparable, even though the diameter of the resistors is reported to be the same. The pressures generated were significantly different, particularly when using small-diameter resistors at a high flow. Therefore, the resistors may not be interchangeable. This is important information for clinicians, particularly when considering PEP for patients who do not tolerate higher pressures. Copyright © 2015 by

  9. Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

    International Nuclear Information System (INIS)

    Greneche, D.

    2014-01-01

    This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

  10. Adaptive Reference Control for Pressure Management in Water Networks

    DEFF Research Database (Denmark)

    Kallesøe, Carsten; Jensen, Tom Nørgaard; Wisniewski, Rafal

    2015-01-01

    Water scarcity is an increasing problem worldwide and at the same time a huge amount of water is lost through leakages in the distribution network. It is well known that improved pressure control can lower the leakage problems. In this work water networks with a single pressure actuator and several....... Subsequently, these relations are exploited in an adaptive reference control scheme for the actuator pressure that ensures constant pressure at the critical points. Numerical experiments underpin the results. © Copyright IEEE - All rights reserved....

  11. Capillarity Induced Negative Pressure of Water Plugs in Nanochannels

    NARCIS (Netherlands)

    Tas, Niels Roelof; Mela, P.; Kramer, Tobias; Berenschot, Johan W.; van den Berg, Albert

    2003-01-01

    We have found evidence that water plugs in hydrophilic nanochannels can be at significant negative pressure due to tensile capillary forces. The negative pressure of water plugs in nanochannels induces bending of the thin channel capping layer, which results in a visible curvature of the liquid

  12. Photocatalysis in Generation of Hydrogen from Water

    KAUST Repository

    Takanabe, Kazuhiro

    2015-04-18

    Solar energy can be converted by utilizing the thermal or photoelectric effects of photons. Concentrated solar power systems utilize thermal energy from the sun by either making steam and then generating power or shifting the chemical equilibrium of a reaction (e.g., water splitting or CO2 reduction) that occurs at extremely high temperatures. The photocatalytic system contains powder photocatalysts. Each photocatalyst particle should collect sufficient photons from the solar flux to cause the required multielectron reactions to occur. The band gap and band edge positions of semiconductors are the most critical parameters for assessing the suitability of photocatalysts for overall water splitting. The most important requirement when selecting photocatalyst materials is the band positions relative to hydrogen and oxygen evolution potentials. For most photocatalysts, surface modification by cocatalysts was found to be essential to achieve overall water splitting.

  13. Method for steam generator water level measurement

    International Nuclear Information System (INIS)

    Srinivasan, J.S.

    1991-01-01

    This paper describes a nuclear power plant, a method of controlling the steam generator water level, wherein the steam generator has an upper level tap corresponding to an upper level, a lower level, a riser positioned between the lower and upper taps, and level sensor means for indicating water level between a first range limit and a second range limit, the sensor means being connected to at least the lower tap. It comprises: calculating a measure of velocity head at about the lower level tap; calculating a measure of full water level as the upper level less the measure of velocity head; calibrating the level sensor means to provide an output at the first limit corresponding to an input thereto representative of the measure of full level; calculating a high level setpoint equal to the level of the riser less a bias amount which is a function of the position of the riser relative to the span between the taps; and controlling the water level when the sensor means indicates that the high level setpoint has been reached

  14. EPR (European Pressurized water Reactor) The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-01-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21. century, which puts the emphasis on sustainable development. The EPR is the only 3. generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR was developed by Framatome and Siemens, whose nuclear activities were combined in January 2001 to form Framatome ANP, a subsidiary of AREVA and Siemens. EDF and the major German electricity companies played an active part in the project. The safety authorities of the two countries joined forces to bring their respective safety standards into line and draw up joint design rules for the new reactor. The project had three objectives: meet the requirements of European utilities, comply with the safety standards laid down by the French safety authority for future pressurized water reactors, in concert with its German counterpart, and make nuclear energy even more competitive than energy generated using fossil fuels. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. This document presents the main characteristics of the EPR, and in particular the additional measures to prevent the occurrence of events likely to damage the core, the leak-tight containment, the measures to reduce the exposure of operating and maintenance personnel, the solutions for an even greater protection of the environment. The foreseen development of the EPR in France and abroad (Finland, China, the United States) is summarized

  15. Development of the Next Generation Type Water Recovery System

    Science.gov (United States)

    Oguchi, Mitsuo; Tachihara, Satoru; Maeda, Yoshiaki; Ueoka, Terumi; Soejima, Fujito; Teranishi, Hiromitsu

    According to NASA, an astronaut living on the International Space Station (ISS) requires approximately 7 kg of water per day. This includes 2 kg of drinking water as well as sanitary fresh water for hand washing, gargling, etc. This water is carried to the space station from the earth, so when more people are staying on the space station, or staying for a longer period of time, the cost of transporting water increases. Accordingly, water is a valuable commodity, and restrictions are applied to such activities as brushing teeth, washing hair, and washing clothes. The life of an astronaut in space is not necessarily a healthy one. JAXA has experience in the research of water recovery systems. Today, utilizing knowledge learned through experiences living on the space station and space shuttles, and taking advantage of the development of new materials for device construction, it is possible to construct a new water recovery system. Therefore, JAXA and New Medican Tech Corporation (NMT) have created a system for collaborative development. Based on the technologies of both companies, we are proceeding to develop the next generation of water recovery devices in order to contribute to safe, comfortable, and healthy daily life for astronauts in space. The goal of this development is to achieve a water purification system based on reverse osmosis (RO) membranes that can perform the following functions. • Preprocessing that removes ammonia and breaks down organic matter contained in urine. • Post-processing that adds minerals and sterilizes the water. • Online TOC measurement for monitoring water quality. • Functions for measuring harmful substances. The RO membrane is an ultra-low-pressure type membrane with a 0.0001 micron (0.1 nanometer) pore size and an operating pressure of 0.4 to 0.6 MPa. During processing with the RO membrane, nearly all of the minerals contained in the cleaned water are removed, resulting in water that is near the quality of deionized water

  16. Evolution of Framatome pressurized water reactor systems

    International Nuclear Information System (INIS)

    Leroy, C.; Bitsch, D.; Millot, J.P.

    1985-10-01

    FRAMATOME's PWR experience covers a total of 63 units, 36 of which are operating by end of 1984. More than 10 units were operated in load follow mode. Progress features, resulting from the feedback of construction and operating experience, and from the returns of a vast research and development program, were incorporated in their design through subsequent series of standard units. The last four loop standard, the N4 model, integrates in a rational way all those progress features, together with a significant design effort. The core design is based on the new Advanced Fuel Assemblies. The reactor control implements the ''Reactor Maximum Flexibility Package'' (R-MAX) which provides a high level of automatic reactor control. The steam generator incorporates an axial-mixed flow economizer design. The triangular-pitch tube bundle, together with modular steam/water separators and a rearrangement of the dryers resulted in a compact design. The reactor coolant pump benefits of higher performances over that of previous models due to an optimal hydraulic design, and of mechanical features which increase margins and facilitate the maintenance work. Following the N4 project, design work on advanced concepts is pursued by FRAMATOME. A main way of research is focused on the optimal use of fissile materials. These concepts are based on tight pitch fuel arrays, associated with a mechanical spectral shift device

  17. 46 CFR 54.01-10 - Steam-generating pressure vessels (modifies U-1(g)).

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Steam-generating pressure vessels (modifies U-1(g)). 54... ENGINEERING PRESSURE VESSELS General Requirements § 54.01-10 Steam-generating pressure vessels (modifies U-1(g)). (a) Pressure vessels in which steam is generated are classed as “Unfired Steam Boilers” except as...

  18. Performance Limiting Effects in Power Generation from Salinity Gradients by Pressure Retarded Osmosis

    KAUST Repository

    Yip, Ngai Yin

    2011-12-01

    Pressure retarded osmosis has the potential to utilize the free energy of mixing when fresh river water flows into the sea for clean and renewable power generation. Here, we present a systematic investigation of the performance limiting phenomena in pressure retarded osmosis-external concentration polarization, internal concentration polarization, and reverse draw salt flux-and offer insights on the design criteria of a high performance pressure retarded osmosis power generation system. Thin-film composite polyamide membranes were chemically modified to produce a range of membrane transport properties, and the water and salt permeabilities were characterized to determine the underlying permeability-selectivity trade-off relationship. We show that power density is constrained by the trade-off between permeability and selectivity of the membrane active layer. This behavior is attributed to the opposing influence of the beneficial effect of membrane water permeability and the detrimental impact of reverse salt flux coupled with internal concentration polarization. Our analysis reveals the intricate influence of active and support layer properties on power density and demonstrates that membrane performance is maximized by tailoring the water and salt permeabilities to the structural parameters. An analytical parameter that quantifies the relative influence of each performance limiting phenomena is employed to identify the dominant effect restricting productivity. External concentration polarization is shown to be the main factor limiting performance at high power densities. Enhancement of the hydrodynamic flow conditions in the membrane feed channel reduces external concentration polarization and thus, yields improved power density. However, doing so will also incur additional operating costs due to the accompanying hydraulic pressure loss. This study demonstrates that by thoughtful selection of the membrane properties and hydrodynamic conditions, the detrimental

  19. Dual pressurized light water reactor producing 2000 M We

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-10-15

    The dual unit optimizer 2000 M We (Duo2000) is proposed as a new design concept for large nuclear power plant. Duo is being designed to meet economic and safety challenges facing the 21 century green and sustainable energy industry. Duo2000 has two nuclear steam supply systems (NSSS) of the unit nuclear optimizer (Uno) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. Uno is anchored to the optimized power reactor 1000 M We (OPR1000) of the Korea Hydro and Nuclear Power Co., Ltd. The concept of Duo can be extended to any number of PWRs or pressurized heavy water reactors (PHWR s), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the small and medium sized reactors (SMRs) be built as units, the concept of Duo2000 will apply to SMRs as well. With its in-vessel retention as severe accident management strategy, Duo can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for generation III + nuclear systems. The strengths of Duo2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting NSSS. The technology can further be extended to coupling modular reactors as dual, triple, or quadruple units to increase their economics, thus accelerating the commercialization as well as the customization of SMRs. (Author)

  20. Generation of subnanosecond electron beams in air at atmospheric pressure

    Science.gov (United States)

    Kostyrya, I. D.; Tarasenko, V. F.; Baksht, E. Kh.; Burachenko, A. G.; Lomaev, M. I.; Rybka, D. V.

    2009-11-01

    Optimum conditions for the generation of runaway electron beams with maximum current amplitudes and densities in nanosecond pulsed discharges in air at atmospheric pressure are determined. A supershort avalanche electron beam (SAEB) with a current amplitude of ˜30 A, a current density of ˜20 A/cm2, and a pulse full width at half maximum (FWHM) of ˜100 ps has been observed behind the output foil of an air-filled diode. It is shown that the position of the SAEB current maximum relative to the voltage pulse front exhibits a time shift that varies when the small-size collector is moved over the foil surface.

  1. Graphite-water steam-generating reactor in the USSR

    Energy Technology Data Exchange (ETDEWEB)

    Dollezhal, N A [AN SSSR, Moscow

    1981-10-01

    One of the types of power reactor used in the USSR is the graphite-water steam-generating reactor RBMK. This produces saturated steam at a pressure of 7MPa. Reactors giving 1GWe each have been installed at the Leningrad, Kursk, Chernobyl and other power stations. Further stations using reactors of this type are being built. A description is given of the fuel element design, and of the layout of the plant. The main characteristics of RBMK reactors using fuel of rated and higher enrichment are listed.

  2. In Situ Raman Study of Liquid Water at High Pressure.

    Science.gov (United States)

    Romanenko, Alexandr V; Rashchenko, Sergey V; Goryainov, Sergey V; Likhacheva, Anna Yu; Korsakov, Andrey V

    2018-06-01

    A pressure shift of Raman band of liquid water (H 2 O) may be an important tool for measuring residual pressures in mineral inclusions, in situ barometry in high-pressure cells, and as an indicator of pressure-induced structural transitions in H 2 O. However, there was no consensus as to how the broad and asymmetric water Raman band should be quantitatively described, which has led to fundamental inconsistencies between reported data. In order to overcome this issue, we measured Raman spectra of H 2 O in situ up to 1.2 GPa using a diamond anvil cell, and use them to test different approaches proposed for the description of the water Raman band. We found that the most physically meaningful description of water Raman band is the decomposition into a linear background and three Gaussian components, associated with differently H-bonded H 2 O molecules. Two of these components demonstrate a pronounced anomaly in pressure shift near 0.4 GPa, supporting ideas of structural transition in H 2 O at this pressure. The most convenient approach for pressure calibration is the use of "a linear background + one Gaussian" decomposition (the pressure can be measured using the formula P (GPa) = -0.0317(3)·Δν G (cm -1 ), where Δν G represents the difference between the position of water Raman band, fitted as a single Gaussian, in measured spectrum and spectrum at ambient pressure).

  3. Design and analysis of pressurized water reactor systems

    International Nuclear Information System (INIS)

    Juhn, P.E.; Kim, Y.H.

    1979-01-01

    To help develop nuclear engineering technologies in local industry sectors, technical and economical data on pressurized water reactor systems and components have been collected, systematically analyzed and computerized to a certain degree. Codes and standards necessary for engineering design of PWR systems have been surveyed and clarified in terms of NSSS, turbine-generator system and BOP, then again rearranged with respect to quality classes and seismic classes. Some design manuals, criteria and guidelines regarding design, construction, test and operation of PWR plants have also been surveyed and collected. Benchmark cost calculation for the construction of a 900 MWe PWR plant, according to the standard format, was carried out, and computer model on construction costs was improved and updated by considering the local supply of labor and materials. And for the indigeneous development of PWR equipment and materials, such data as delivery schedule and manufacturers of 52 systems and 36,000 components have also been reviewed herein. (author)

  4. SIMULATION OF NEGATIVE PRESSURE WAVE PROPAGATION IN WATER PIPE NETWORK

    Directory of Open Access Journals (Sweden)

    Tang Van Lam

    2017-11-01

    Full Text Available Subject: factors such as pipe wall roughness, mechanical properties of pipe materials, physical properties of water affect the pressure surge in the water supply pipes. These factors make it difficult to analyze the transient problem of pressure evolution using simple programming language, especially in the studies that consider only the magnitude of the positive pressure surge with the negative pressure phase being neglected. Research objectives: determine the magnitude of the negative pressure in the pipes on the experimental model. The propagation distance of the negative pressure wave will be simulated by the valve closure scenarios with the help of the HAMMER software and it is compared with an experimental model to verify the quality the results. Materials and methods: academic version of the Bentley HAMMER software is used to simulate the pressure surge wave propagation due to closure of the valve in water supply pipe network. The method of characteristics is used to solve the governing equations of transient process of pressure change in the pipeline. This method is implemented in the HAMMER software to calculate the pressure surge value in the pipes. Results: the method has been applied for water pipe networks of experimental model, the results show the affected area of negative pressure wave from valve closure and thereby we assess the largest negative pressure that may appear in water supply pipes. Conclusions: the experiment simulates the water pipe network with a consumption node for various valve closure scenarios to determine possibility of appearance of maximum negative pressure value in the pipes. Determination of these values in real-life network is relatively costly and time-consuming but nevertheless necessary for identification of the risk of pipe failure, and therefore, this paper proposes using the simulation model by the HAMMER software. Initial calibration of the model combined with the software simulation results and

  5. Electrokinetic Hydrogen Generation from Liquid WaterMicrojets

    Energy Technology Data Exchange (ETDEWEB)

    Duffin, Andrew M.; Saykally, Richard J.

    2007-05-31

    We describe a method for generating molecular hydrogen directly from the charge separation effected via rapid flow of liquid water through a metal orifice, wherein the input energy is the hydrostatic pressure times the volume flow rate. Both electrokinetic currents and hydrogen production rates are shown to follow simple equations derived from the overlap of the fluid velocity gradient and the anisotropic charge distribution resulting from selective adsorption of hydroxide ions to the nozzle surface. Pressure-driven fluid flow shears away the charge balancing hydronium ions from the diffuse double layer and carries them out of the aperture. Downstream neutralization of the excess protons at a grounded target electrode produces gaseous hydrogen molecules. The hydrogen production efficiency is currently very low (ca. 10-6) for a single cylindrical jet, but can be improved with design changes.

  6. Self-Propagating Frontal Polymerization in Water at Ambient Pressure

    Science.gov (United States)

    Olten, Nesrin; Kraigsley, Alison; Ronney, Paul D.

    2003-01-01

    boiling point solutions are needed because in order to produce a propagating front, a high front temperature is needed to produce sufficiently rapid decomposition of the free radical initiator and subsequent free radical polymerization and heat release at a rate faster than heat losses remove thermal energy from the system. (While the conduction heat loss rate increases linearly with temperature, the free radical initiator decomposition is a high activation energy process whose rate increases much more rapidly than linearly with temperature, thus as the temperature decreases, the ratio of heat loss to heat generation increases, eventually leading to extinction of the front if the temperature is too low.) In order to obtain atmospheric pressure frontal polymerization in water, it is necessary to identify a monomer/initiator combination that is water soluble and will not extinguish even when the peak temperature (T*) is less than 100 C. In this work acrylic acid (AA) was chosen as the monomer because is it one of the most reactive monomers and can polymerize readily at low temperatures even without initiators. Ammonium persulfate (AP) was chosen as the initiator because it decomposes readily at low temperatures, produces relatively few bubbles and is commercially available. The propagation rates and extinction conditions of the fronts are studied for a range of AA and AP concentrations. Small amounts of fumed silica powder (Cab-o-sil, Cabot Corporation) were added to the solutions to inhibit buoyancy induced convection in the solutions; future studies will investigate the effects of buoyant convection within the solutions.

  7. Modeling and simulation of pressurized water reactor power plant

    International Nuclear Information System (INIS)

    Wang, S.J.

    1983-01-01

    Two kinds of balance of plant (BOP) models of a pressurized water reactor (PWR) system are developed in this work - the detailed BOP model and the simple BOP model. The detailed model is used to simulate the normal operational performance of a whole BOP system. The simple model is used to combine with the NSSS model for a whole plant simulation. The trends of the steady state values of the detailed model are correct and the dynamic responses are reasonable. The simple BOP model approach starts the modelling work from the overall point of view. The response of the normalized turbine power and the feedwater inlet temperature to the steam generator of the simple model are compared with those of the detailed model. Both the steady state values and the dynamic responses are close to those of the detailed model. The simple BOP model is found adequate to represent the main performance of the BOP system. The simple balance of plant model was coupled with a NSSS model for a whole plant simulation. The NSSS model consists of the reactor core model, the steam generator model, and the coolant temperature control system. A closed loop whole plant simulation for an electric load perturbation was performed. The results are plausible. The coupling effect between the NSSS system and the BOP system was analyzed. The feedback of the BOP system has little effect on the steam generator performance, while the performance of the BOP system is strongly affected by the steam flow rate from the NSSS

  8. Design and safety of the Sizewell pressurized water reactor

    International Nuclear Information System (INIS)

    Marshall, W.

    1983-01-01

    The Central Electricity Generating Board propose to build a pressurized water reactor at Sizewell in Suffolk. The PWR Task Force was set up in June 1981 to provide a communications centre for developing firm design proposals for this reactor. These were to follow the Standardized Nuclear Unit Power Plant System designed by Bechtel for the Westinghouse nuclear steam supply system for reactors built in the United States. Changes were required to the design to accommodate, for example, the use of two turbine generators and to satisfy British safety requirements. Differences exist between the British and American licensing procedures. In the UK the statutory responsibility for the safety of a nuclear power station rests unambiguously with the Generating Boards. In the U.S.A. the Nuclear Regulatory Commission issues detailed written instructions, which must be followed precisely. Much of the debate on the safety of nuclear power focuses on the risks of big nuclear accidents. It is necessary to explain to the public what, in a balanced perspective, the risks of accidents actually are. The long-term consequences can be presented in terms of reduction in life expectancy, increased chance of cancer or the equivalent pattern of compulsory cigarette smoking. (author)

  9. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  10. Solubility and physical properties of sugars in pressurized water

    International Nuclear Information System (INIS)

    Saldaña, Marleny D.A.; Alvarez, Víctor H.; Haldar, Anupam

    2012-01-01

    Highlights: ► Sugar solubility in pressurized water and density at high pressures were measured. ► Glucose solubility was higher than that of lactose as predicted by their σ-profiles. ► Sugar aqueous solubility decreased with an increase in pressure from 15 to 120 bar. ► Aqueous glucose molecular packing shows high sensitivity to pressure. ► The COSMO-SAC model qualitatively predicted the sugar solubility data. - Abstract: In this study, the solubility, density, and refractive index of glucose and lactose in water as a function of temperature were measured. For solubility of sugars in pressurized water, experimental data were obtained at pressures of (15 to 120) bar and temperatures of (373 to 433) K using a dynamic flow high pressure system. Density data for aqueous sugar solutions were obtained at pressures of (1 to 300) bar and temperatures of (298 to 343) K. The refractive index of aqueous sugar solutions was obtained at 293 K and atmospheric pressure. Activity coefficient models, Van Laar and the Conductor-like Screening Model-Segment Activity Coefficient (COSMO-SAC), were used to fit and predict the experimental solubility data, respectively. The results obtained showed that the solubility of both sugars in pressurized water increase with an increase in temperature. However, with the increase of pressure from 15 bar to 120 bar, the solubility of both sugars in pressurized water decreased. The Van Laar model fit the experimental aqueous solubility data with deviations lower than 13 and 53% for glucose and lactose, respectively. The COSMO-SAC model predicted qualitatively the aqueous solubility of these sugars.

  11. Importance of pressure reducing valves (PRVs) in water supply networks.

    Science.gov (United States)

    Signoreti, R. O. S.; Camargo, R. Z.; Canno, L. M.; Pires, M. S. G.; Ribeiro, L. C. L. J.

    2016-08-01

    Challenged with the high rate of leakage from water supply systems, these managers are committed to identify control mechanisms. In order to standardize and control the pressure Pressure Reducing Valves (VRP) are installed in the supply network, shown to be more effective and provide a faster return for the actual loss control measures. It is known that the control pressure is while controlling the occurrence of leakage. Usually the network is sectored in areas defined by pressure levels according to its topography, once inserted the VRP in the same system will limit the downstream pressure. This work aims to show the importance of VRP as loss reduction for tool.

  12. The testing of a steam-water separating device used for vertical steam generators

    International Nuclear Information System (INIS)

    Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

    1989-01-01

    The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

  13. Assessment of water pipes durability under pressure surge

    Science.gov (United States)

    Pham Ha, Hai; Minh, Lanh Pham Thi; Tang Van, Lam; Bulgakov, Boris; Bazhenova, Soafia

    2017-10-01

    Surge phenomenon occurs on the pipeline by the closing valve or pump suddenly lost power. Due to the complexity of the water hammer simulation, previous researches have only considered water hammer on the single pipe or calculation of some positions on water pipe network, it have not been analysis for all of pipe on the water distribution systems. Simulation of water hammer due to closing valve on water distribution system and the influence level of pressure surge is evaluated at the defects on pipe. Water hammer on water supply pipe network are simulated by Water HAMMER software academic version and the capacity of defects are calculated by SINTAP. SINTAP developed from Brite-Euram projects in Brussels-Belgium with the aim to develop a process for assessing the integrity of the structure for the European industry. Based on the principle of mechanical fault, indicating the size of defects in materials affect the load capacity of the product in the course of work, the process has proposed setting up the diagram to fatigue assessment defect (FAD). The methods are applied for water pipe networks of Lien Chieu district, Da Nang city, Viet Nam, the results show the affected area of wave pressure by closing the valve and thereby assess the greatest pressure surge effect to corroded pipe. The SINTAP standard and finite element mesh analysis at the defect during the occurrence of pressure surge which will accurately assess the bearing capacity of the old pipes. This is one of the bases to predict the leakage locations on the water distribution systems. Amount of water hammer when identified on the water supply networks are decreasing due to local losses at the nodes as well as the friction with pipe wall, so this paper adequately simulate water hammer phenomena applying for actual water distribution systems. The research verified that pipe wall with defect is damaged under the pressure surge value.

  14. Impact of pressure losses in small-sized parabolic-trough collectors for direct steam generation

    International Nuclear Information System (INIS)

    Lobón, David H.; Valenzuela, Loreto

    2013-01-01

    Using PTC (parabolic-trough solar collectors) for industrial thermal processes in the temperature range up to 300 °C is not new, but in recent years there is a boosted interest in this type of concentrating solar technology. One of the problems that arise when designing PTC solar fields is how to deal with the pressure losses which are critical when producing saturated steam directly in the collectors. Depending on the characteristics of the collector, mainly on the receiver diameter, and on the nominal process conditions defined, a solar field configuration can be feasible or not. This paper presents a sensitivity analysis done using a software tool developed to study the thermo-hydraulic behaviour of PTC systems using water-steam as heat transfer fluid. In the case study presented, a small-sized PTC designed for industrial process heat applications is considered, which has a focal length of 0.2 m, an aperture area of 2 m 2 , and its receiver pipe has an inner diameter of 15 mm. Varied process conditions are inlet water pressure, temperature, and mass flow rate, solar irradiance and incidence angle of solar radiation. Results show that working pressure definition is particularly critical to make feasible or not the direct steam generation in solar collectors. - Highlights: • DSG (Direct steam generation) in small-sized parabolic-trough collectors. • Thermo-hydraulic sensitivity analysis. • Influence of working pressure and receiver geometry in DSG process

  15. Measurement of micro Bubbles generated by a pressurized dissolution method

    Energy Technology Data Exchange (ETDEWEB)

    Hosokawa, S; Tanaka, K; Tomiyama, A [Kobe University, 1-1 Rokkodai, Nada, Kobe 657-8501 (Japan); Maeda, Y; Yamaguchi, S; Ito, Y, E-mail: hosokawa@mech.kobe-u.ac.j [Panasonic Electric Works Co., Ltd., 1048 Kadoma, Osaka 571-8686 (Japan)

    2009-02-01

    Diameters of micro-bubbles are apt to range from about one mm to several-hundred mm, and therefore, it is difficult to measure a correct diameter distribution using a single measurement method. In this study, diameters of bubbles generated by a pressurized dissolution method are measured by using phase Doppler anemometry (PDA) and an image processing method, which is based on the Sobel filter and Hough transform. The diameter distribution and the Sauter mean diameter of micro bubbles are evaluated based on the diameters measured by both methods. Experiments are conducted for several mass flow rates of dissolved gas and of air bubbles entrained in the upstream of the decompression nozzle to examine effects of the entrained bubbles on bubble diameter. As a result, the following conclusions are obtained: (1) Diameter distribution of micro bubbles can be accurately measured for a wide range of diameter by using PDA and the image processing method. (2) The mean diameter of micro-bubbles generated by gasification of dissolved gas is smaller than that generated by breakup of air bubbles entrained in the upstream of the decompression nozzle. (3) The mean bubble diameter increases with the entrainment of air bubbles in the upstream of the decompression nozzle at a constant mass flow rate of dissolved gas.

  16. Measurement of micro Bubbles generated by a pressurized dissolution method

    International Nuclear Information System (INIS)

    Hosokawa, S; Tanaka, K; Tomiyama, A; Maeda, Y; Yamaguchi, S; Ito, Y

    2009-01-01

    Diameters of micro-bubbles are apt to range from about one mm to several-hundred mm, and therefore, it is difficult to measure a correct diameter distribution using a single measurement method. In this study, diameters of bubbles generated by a pressurized dissolution method are measured by using phase Doppler anemometry (PDA) and an image processing method, which is based on the Sobel filter and Hough transform. The diameter distribution and the Sauter mean diameter of micro bubbles are evaluated based on the diameters measured by both methods. Experiments are conducted for several mass flow rates of dissolved gas and of air bubbles entrained in the upstream of the decompression nozzle to examine effects of the entrained bubbles on bubble diameter. As a result, the following conclusions are obtained: (1) Diameter distribution of micro bubbles can be accurately measured for a wide range of diameter by using PDA and the image processing method. (2) The mean diameter of micro-bubbles generated by gasification of dissolved gas is smaller than that generated by breakup of air bubbles entrained in the upstream of the decompression nozzle. (3) The mean bubble diameter increases with the entrainment of air bubbles in the upstream of the decompression nozzle at a constant mass flow rate of dissolved gas.

  17. Automatic generation of water distribution systems based on GIS data.

    Science.gov (United States)

    Sitzenfrei, Robert; Möderl, Michael; Rauch, Wolfgang

    2013-09-01

    In the field of water distribution system (WDS) analysis, case study research is needed for testing or benchmarking optimisation strategies and newly developed software. However, data availability for the investigation of real cases is limited due to time and cost needed for data collection and model setup. We present a new algorithm that addresses this problem by generating WDSs from GIS using population density, housing density and elevation as input data. We show that the resulting WDSs are comparable to actual systems in terms of network properties and hydraulic performance. For example, comparing the pressure heads for an actual and a generated WDS results in pressure head differences of ±4 m or less for 75% of the supply area. Although elements like valves and pumps are not included, the new methodology can provide water distribution systems of varying levels of complexity (e.g., network layouts, connectivity, etc.) to allow testing design/optimisation algorithms on a large number of networks. The new approach can be used to estimate the construction costs of planned WDSs aimed at addressing population growth or at comparisons of different expansion strategies in growth corridors.

  18. Neutron fluence determination for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Gold, R.

    1994-01-01

    A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

  19. Performance Evaluation of Pressure Transducers for Water Impacts

    Science.gov (United States)

    Vassilakos, Gregory J.; Stegall, David E.; Treadway, Sean

    2012-01-01

    The Orion Multi-Purpose Crew Vehicle is being designed for water landings. In order to benchmark the ability of engineering tools to predict water landing loads, test programs are underway for scale model and full-scale water impacts. These test programs are predicated on the reliable measurement of impact pressure histories. Tests have been performed with a variety of pressure transducers from various manufacturers. Both piezoelectric and piezoresistive devices have been tested. Effects such as thermal shock, pinching of the transducer head, and flushness of the transducer mounting have been studied. Data acquisition issues such as sampling rate and anti-aliasing filtering also have been studied. The response of pressure transducers have been compared side-by-side on an impulse test rig and on a 20-inch diameter hemisphere dropped into a pool of water. The results have identified a range of viable configurations for pressure measurement dependent on the objectives of the test program.

  20. Bridge pressure flow scour for clear water conditions

    Science.gov (United States)

    2009-10-01

    The equilibrium scour at a bridge caused by pressure flow with critical approach velocity in clear-water simulation conditions was studied both analytically and experimentally. The flume experiments revealed that (1) the measured equilibrium scour pr...

  1. Design of an additional heat sink based on natural circulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Frischengruber, Kurt; Solanilla, Roberto; Fernandez, Ricardo; Blumenkrantz, Arnaldo; Castano, Jorge

    1989-01-01

    Residual heat removal through the steam generators in Nuclear Power Plant with pressurized water reactors (PWR) or pressurized heavy water reactors (PHWR in pressured vessel or pressured tube types) requires the maintenance of the steam generator inventory and the availability of and appropriate heat sink, which are based on the operability of the steam generators feedwater system. This paper describes the conceptual design of an assured heat removal system which includes only passive elements and is based on natural circulation. The system can supplement the original systems of the plant. The new system includes a condenser/boiler heat exchanger to condense the steam produced in the steam generator, transferring the heat to the water of an open pool at atmospheric pressure. The condensed steam flows back to the steam generators by natural circulation effects. The performance of an Atucha type PHWR nuclear power station with and without the proposed system is calculated in an emergency power case for the first 5000 seconds after the incident. The analysis shows that the proposed system offers the possibility to cool-down the plant to a low energy state during several hours and avoids the repeated actuation of the primary and secondary system safety valves. (Author) [es

  2. Device for automatically operating cooling mode of water in a pressure suppression chamber

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1975-01-01

    Object: To provide a system for removing residual heat in a reactor safety system, which can automatically cool water in a pressure suppression chamber when a load on a generator is cut off, so as not to scram the reactor. Structure: When a load cut-off signal is generated by means of rapid closure of a turbine regulating valve or due to the load unbalance relay of generator output, or the like, a sea water pump is started to fully open an outlet valve for the sea water pump, a heat exchanging inlet valve and a minimum crow valve and to fully close a heat exchanging bypass valve. In this manner, cooling water for the heat exchanger is secured to start the pump in the system for removing residual heat, and when the pump discharge pressure is in normal condition, the inlet valve in pressure suppression chamber and the spray valve in the pressure suppression chamber are fully opened to automatically cool water in the pressure suppression chamber. (Hanada, M.)

  3. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  4. Electrochemical noise measurements under pressurized water reactor conditions

    International Nuclear Information System (INIS)

    Van Nieuwenhove, R.

    2000-01-01

    Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

  5. Design of a pressurized water loop heated by electric resistances

    International Nuclear Information System (INIS)

    Ribeiro, S.V.G.

    1981-01-01

    A pressurized water loop design is presented. Its operating pressure is 420 psi and we seek to simulate qualitatively some thermo-hydraulic phenomena of PWR reactors. The primary circuit simulator consists basically of two elements: 1)the test section housing 16 electric resistences dissipating a total power of 100 Kw; 2)the loop built of SCH40S 304L steel piping, consisting of the pump, a heat exchanger and the pressurizer. (Author) [pt

  6. Methodology of fuel rod design for pressurized light water reactors

    International Nuclear Information System (INIS)

    Teixeira e Silva, A.; Esteves, A.M.

    1988-01-01

    The fuel performance program FRAPCON-1 and the structural finite element program SAP-IV are applied in a pressurized water reactor fuel rod design methodology. The applied calculation procedure allows to dimension the fuel rod components and characterize its internal pressure. (author) [pt

  7. On OH production in water containing atmospheric pressure plasmas

    NARCIS (Netherlands)

    Bruggeman, P.J.; Schram, D.C.

    2010-01-01

    In this paper radical production in atmospheric pressure water containing plasmas is discussed. As OH is often an important radical in these discharges the paper focuses on OH production. Besides nanosecond pulsed coronas and diffusive glow discharges, several other atmospheric pressure plasmas

  8. Complex cooling water systems optimization with pressure drop consideration

    CSIR Research Space (South Africa)

    Gololo, KV

    2012-12-01

    Full Text Available Pressure drop consideration has shown to be an essential requirement for the synthesis of a cooling water network where reuse/recycle philosophy is employed. This is due to an increased network pressure drop associated with additional reuse...

  9. Where Did the Water Go?: Boyle's Law and Pressurized Diaphragm Water Tanks

    Science.gov (United States)

    Brimhall, James; Naga, Sundar

    2007-01-01

    Many homes use pressurized diaphragm tanks for storage of water pumped from an underground well. These tanks are very carefully constructed to have separate internal chambers for the storage of water and for the air that provides the pressure. One might expect that the amount of water available for use from, for example, a 50-gallon tank would be…

  10. Testing of acoustic emission method during pressure tests of WWER-440 steam generators and pressurizers

    International Nuclear Information System (INIS)

    Wuerfl, K.; Crha, J.

    1987-01-01

    The results are discussed of measuring acoustic emission in output pressure testing of steam generators and pressurizers for WWER-440 reactors. The objective of the measurements was to test the reproducibility of measurements and to find the criterion which would be used in assessing the condition of the components during manufacture and in operation. The acoustic emission was measured using a single-channel Dunegan/Endevco apparatus and a 16-channel LOCAMAT system. The results showed that after the first assembly, during a repeat dismantle of the lids and during seal replacement, processes due to seal contacts and bolt and washer deformations were the main source of acoustic emission. A procedure was defined of how to exclude new acoustic emission sources in such cases. The acoustic emission method can be used for the diagnostics of plastic deformation processes or of crack production and propagation in components during service. (Z.M.)

  11. Development of second-generation pressurized fluidized bed combustion process

    Energy Technology Data Exchange (ETDEWEB)

    Wolowodiuk, W.; Robertson, A. [Foster Wheeler Development Corp., Livingston, NJ (United States); Bonk, D. [Dept. of Energy, Morgantown, WV (United States)

    1995-12-01

    Under the sponsorship of the United States Department of Energy, Foster Wheeler Development Corporation, and its team members, Westinghouse, Gilbert/Commonwealth, and the Institute of Gas Technology are developing second-generation pressurized fluidized bed combustion technology capable of achieving net plant efficiency in excess of 45 percent based on the higher heating value of the coal. A three-phase program entails design and costing of a 500 MWe power plant and identification of developments needed to commercialize this technology (Phase 1), testing of individual components (Phase 2), and finally testing these components in an integrated mode (Phase 3). This paper briefly describes the results of the first two phases as well as the progress on the third phase. Since other projects which use the same technology are in construction or in negotiation stages-namely, the Power System Development Facility and the Four Rivers Energy Modernization Projects-brief descriptions of these are also included.

  12. Combined Cycle Power Generation Employing Pressure Gain Combustion

    Energy Technology Data Exchange (ETDEWEB)

    Holley, Adam [United Technologies Corporation, East Hartford, CT (United States). Research Center

    2017-05-15

    The Phase I program assessed the potential benefit of applying pressure gain combustion (PGC) technology to a natural gas combined cycle power plant. A conceptual design of the PGC integrated gas turbine was generated which was simulated in a detailed system modeling tool. The PGC integrated system was 1.93% more efficient, produced 3.09% more power, and reduced COE by 0.58%. Since the PGC system used had the same fuel flow rate as the baseline system, it also reduced CO2 emissions by 3.09%. The PGC system did produce more NOx than standard systems, but even with the performanceand cost penalties associated with the cleanup system it is better in every measure. This technology benefits all of DOE’s stated program goals to improve plant efficiency, reduce CO2 production, and reduce COE.

  13. High-pressure water electrolysis: Electrochemical mitigation of product gas crossover

    International Nuclear Information System (INIS)

    Schalenbach, Maximilian; Stolten, Detlef

    2015-01-01

    Highlights: • New technique to reduce gas crossover during water electrolysis • Increase of the efficiency of pressurized water electrolysis • Prevention of safety hazards due to explosive gas mixtures caused by crossover • Experimental realization for a polymer electrolyte membrane electrolyzer • Discussion of electrochemical crossover mitigation for alkaline water electrolysis - Abstract: Hydrogen produced by water electrolysis can be used as an energy carrier storing electricity generated from renewables. During water electrolysis hydrogen can be evolved under pressure at isothermal conditions, enabling highly efficient compression. However, the permeation of hydrogen through the electrolyte increases with operating pressure and leads to efficiency loss and safety hazards. In this study, we report on an innovative concept, where the hydrogen crossover is electrochemically mitigated by an additional electrode between the anode and the cathode of the electrolysis cell. Experimentally, the technique was applied to a proton exchange membrane water electrolyzer operated at a hydrogen pressure that was fifty times larger than the oxygen pressure. Therewith, the hydrogen crossover was reduced and the current efficiency during partial load operation was increased. The concept is also discussed for water electrolysis that is operated at balanced pressures, where the crossover of hydrogen and oxygen is mitigated using two additional electrodes

  14. Temperature/pressure and water vapor sounding with microwave spectroscopy

    Science.gov (United States)

    Muhleman, D. O.; Janssen, M. A.; Clancy, R. T.; Gulkis, S.; Mccleese, D. J.; Zurek, R.; Haberle, R. M.; Frerking, M.

    1992-01-01

    Two intense microwave spectra lines exist in the martian atmosphere that allow unique sounding capabilities: water vapor at 183 GHz and the (2-1) rotational line of CO at 230 GHz. Microwave spectra line sounding is a well-developed technique for the Earth's atmosphere for sounding from above from spacecraft and airplanes, and from below from fixed surface sites. Two simple instruments for temperature sounding on Mars (the CO line) and water vapor measurements are described. The surface sounder proposed for the MESUR sites is designed to study the boundary layer water vapor distribution and the temperature/pressure profiles with vertical resolution of 0.25 km up to 1 km with reduced resolution above approaching a scale height. The water channel will be sensitive to a few tenths of a micrometer of water and the temperature profile will be retrieved to an accuracy between 1 and 2 K. The latter is routinely done on the Earth using oxygen lines near 60 GHz. The measurements are done with a single-channel heterodyne receiver looking into a 10-cm mirror that is canned through a range of elevation angles plus a target load. The frequency of the receiver is sweep across the water and CO lines generating the two spectra at about 1-hr intervals throughout the mission. The mass and power for the proposed instrument are 2 kg and 5-8 W continuously. The measurements are completely immune to the atmospheric dust and ice particle loads. It was felt that these measurements are the ultimate ones to properly study the martian boundary layer from the surface to a few kilometers. Sounding from above requires an orbiting spacecraft with multichannel microwave spectrometers such as the instrument proposed for MO by a subset of the authors, a putative MESUR orbiter, and a proposed Discovery mission called MOES. Such an instrument can be built with less than 10 kg and use less than 15 W. The obvious advantage of this approach is that the entire atmosphere can be sounded for temperature and

  15. Experimental Study and Engineering Practice of Pressured Water Coupling Blasting

    Directory of Open Access Journals (Sweden)

    J. X. Yang

    2017-01-01

    Full Text Available Overburden strata movement in large space stope is the major reason that induces the appearance of strong mining pressure. Presplitting blasting for hard coal rocks is crucial for the prevention and control of strong pressure in stope. In this study, pressured water coupling blasting technique was proposed. The process and effect of blasting were analyzed by orthogonal test and field practice. Results showed that the presence of pressure-bearing water and explosive cartridges in the drill are the main influence factors of the blasting effect of cement test block. The high load-transmitting performance of pore water and energy accumulation in explosive cartridges were analyzed. Noxious substances produced during the blasting process were properly controlled because of the moistening, cooling, and diluting effect of pore water. Not only the goal of safe and static rock fragmentation by high-explosive detonation but also a combination of superdynamic blast loading and static loading effect of the pressured water was achieved. Then the practice of blasting control of hard coal rocks in Datong coal mine was analyzed to determine reasonable parameters of pressured water coupling blasting. A good presplitting blasting control effect was achieved for the hard coal rocks.

  16. Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

    2009-01-01

    An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

  17. Pressure control for minimizing leakage in water distribution systems

    OpenAIRE

    Nourhan Samir; Rawya Kansoh; Walid Elbarki; Amr Fleifle

    2017-01-01

    In the last decades water resources availability has been a major issue on the international agenda. In a situation of worsening scarcity of water resources and the rapidly increasing of water demands, the state of water losses management is part of manâs survival on earth. Leakage in water supply networks makes up a significant amount, sometimes more than 70% of the total water losses. The best practices suggest that pressure management is one of the most effective way to reduce the amount o...

  18. Introduction of advanced pressurized water reactors in France

    International Nuclear Information System (INIS)

    Millot, J.P.; Nigon, M.; Vitton, M.

    1988-01-01

    Designed >30 yr ago, pressurized water reactors (PWRs) have evolved well to match the current safety, operating, and economic requirements. The first advanced PWR generation, the N4 reactor, is under construction with 1992 as a target date for commercial operation. The N4 may be considered to be a technological outcome of PWR evolution, providing advances in the fields of safety, man/machine interfaces, and load flexibility. As a step beyond N4, a second advanced PWR generation is presently under definition with, as a main objective, a greater ability to cope with the possible deterioration of the natural uranium market. In 1986, Electricite de France (EdF) launched investigations into the possible characteristics of this advanced PWR, called REP-2000 (PWR-2000: the reactor for the next century). Framatome joined EdF in 1987 but had been working on a new tight-lattice reactor. Main options are due by 1988; preliminary studies will begin and, by 1990, detailed design will proceed with the intent of firm commitments for the first unit by 1995. Commissioning is planned in the early years of the next century. This reactor type should be either an improved version of the N4 reactor or a spectral shift convertible reactor (RCVS). Through research and development efforts, Framatome, Commissariat a l'Energie Atomique (CEA), and EdF are investigating the physics of fuel rod tight lattices including neutronics, thermohydraulics, fuel behavior, and reactor mechanics

  19. Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1994-11-01

    The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

  20. The phase diagram of water at negative pressures: virtual ices.

    Science.gov (United States)

    Conde, M M; Vega, C; Tribello, G A; Slater, B

    2009-07-21

    The phase diagram of water at negative pressures as obtained from computer simulations for two models of water, TIP4P/2005 and TIP5P is presented. Several solid structures with lower densities than ice Ih, so-called virtual ices, were considered as possible candidates to occupy the negative pressure region of the phase diagram of water. In particular the empty hydrate structures sI, sII, and sH and another, recently proposed, low-density ice structure. The relative stabilities of these structures at 0 K was determined using empirical water potentials and density functional theory calculations. By performing free energy calculations and Gibbs-Duhem integration the phase diagram of TIP4P/2005 was determined at negative pressures. The empty hydrates sII and sH appear to be the stable solid phases of water at negative pressures. The phase boundary between ice Ih and sII clathrate occurs at moderate negative pressures, while at large negative pressures sH becomes the most stable phase. This behavior is in reasonable agreement with what is observed in density functional theory calculations.

  1. Development of a low frost-point generator operating at sub-atmospheric pressure

    Science.gov (United States)

    Cuccaro, R.; Rosso, L.; Smorgon, D.; Beltramino, G.; Tabandeh, S.; Fernicola, V.

    2018-05-01

    A low frost-point generator (INRIM 03) operating at sub-atmospheric pressure has been designed and constructed at the Istituto Nazionale di Ricerca Metrologica (INRIM) as part of a calibration facility for upper-air sounding instruments. This new humidity generator covers the frost-point temperature range between  ‑99 °C and  ‑20 °C and works at any controlled pressure between 200 hPa and 1100 hPa, achieving a complete saturation of the carrier gas (nitrogen) in a single passage through a stainless steel isothermal saturator. The generated humid gas contains a water vapour amount fraction between 14  ×  10‑9 mol mol‑1 and 5  ×  10‑3 mol mol‑1. In this work the design of the generator is reported together with characterisation and performance evaluation tests. A preliminary validation of the INRIM 03 against one of the INRIM humidity standards in the common region is also included. Based on experimental test results, an initial uncertainty evaluation of the generated frost-point temperature, T fp, and water vapour amount fraction, x w, in the limited range down to  ‑75 °C at atmospheric pressure is reported. For the frost-point temperature, the uncertainty budget yields a total expanded uncertainty (k  =  2) of less than 0.028 °C, while for the mole fraction the budget yields a total expanded uncertainty of less than 10‑6 mol mol‑1.

  2. A dual pressurized water reactor producing 2000 MWe

    International Nuclear Information System (INIS)

    Kang, K. M.; Suh, K. Y.

    2010-01-01

    The Dual Unit Optimizer 2000 MWe (DUO2000) is proposed as a new design concept for large nuclear power plant. DUO is being designed to meet economic and safety challenges facing the 21. century green and sustainable energy industry. DUO2000 has two nuclear steam supply systems (NSSSs) of the Unit Nuclear Optimizer (UNO) pressurized water reactor (PWR) in a single containment so as to double the capacity of the plant. UNO is anchored to the Optimized Power Reactor 1000 MWe (OPR1000). The concept of DUO can be extended to any number of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactor (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. In particular, since it is required that the Small and Medium sized Reactors (SMRs) be built as units, the concept of DUO2000 will apply to SMRs as well. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to end, but also pave ways to most promising large power capacity dispensing with huge redesigning cost for Generation III+ nuclear systems. Also, the strengths of DUO2000 include reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS. Two prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The Coolant Unit Branching Apparatus (CUBA) is proposed

  3. Pressure: the politechnics of water supply in Mumbai.

    Science.gov (United States)

    Anand, Nikhil

    2011-01-01

    In Mumbai, most all residents are delivered their daily supply of water for a few hours every day, on a water supply schedule. Subject to a more precarious supply than the city's upper-class residents, the city's settlers have to consistently demand that their water come on “time” and with “pressure.” Taking pressure seriously as both a social and natural force, in this article I focus on the ways in which settlers mobilize the pressures of politics, pumps, and pipes to get water. I show how these practices not only allow settlers to live in the city, but also produce what I call hydraulic citizenship—a form of belonging to the city made by effective political and technical connections to the city's infrastructure. Yet, not all settlers are able to get water from the city water department. The outcomes of settlers' efforts to access water depend on a complex matrix of socionatural relations that settlers make with city engineers and their hydraulic infrastructure. I show how these arrangements describe and produce the cultural politics of water in Mumbai. By focusing on the ways in which residents in a predominantly Muslim settlement draw water despite the state's neglect, I conclude by pointing to the indeterminacy of water, and the ways in which its seepage and leakage make different kinds of politics and publics possible in the city.

  4. Integrity of high-velocity water slug generated by an impacting technique

    Science.gov (United States)

    Dehkhoda, Sevda; Bourne, Neil

    2013-06-01

    A pulsed water jet is a series of discrete water slugs travelling at high velocity. Immediately after striking a target, these slugs apply high-intensity, short-duration transient stress known as the water hammer pressure, followed by low-intensity, long-duration stationary stress at the stagnation pressure. The magnitude and duration of the water hammer and stagnation pressures are controlled by the size and quality of the water slugs. The use of water jets for rock cutting in mining operations is a centuries-old technology; however, practical methods for producing high-energy water slugs repeatedly have proven difficult. This can be partly due to the fact that the geometrical properties of a jet and so its effectiveness in creating damage is controlled and influenced by the method that is employed to generate the water slugs. This paper investigates the integrity of a single water slug produced using an impacting technique where a hammer strikes a piston, resting on top of a water-filled chamber. The coherence of the generated water pulse was of concern in this study. If repeated shock reflections within the chamber were transmitted or were carried into the internal geometry of nozzle, the emerging jet could pulsate. The impact impulse of the formed water jet was measured in a Kel-F target material using an embedded PVDF (Polyvinylidene fluoride) shock gauge. The recorded stress waveform was then used to study the quality and endurance of the water pulse stream as it travelled through air.

  5. Tensile Strength of Water Exposed to Pressure Pulses

    DEFF Research Database (Denmark)

    Andersen, Anders Peter; Mørch, Knud Aage

    2012-01-01

    at an extended water-solid interface by imposing a tensile stress pulse which easily causes cavitation. Next, a compressive pulse of duration ~1 ms and a peak intensity of a few bar is imposed prior to the tensile stress pulse. A dramatic increase of the tensile strength is observed immediately after......It is well known that pressurization for an extended period of time increases the tensile strength of water, but little information is available on the effect of pressure pulses of short duration. This is addressed in the present paper where we first measure the tensile strength of water...

  6. Is high-pressure water the cradle of life?

    International Nuclear Information System (INIS)

    Bassez, Marie-Paule

    2003-01-01

    Several theories have been proposed for the synthesis of prebiotic molecules. This letter shows that the structure of supercritical water, or high-pressure water, could trigger prebiotic synthesis and the origin of life deep in the oceans, in hydrothermal vent systems. Dimer geometries of high-pressure water may have a point of symmetry and a zero dipole moment. Consequently, simple apolar molecules found in submarine hydrothermal vent systems will dissolve in the apolar environment provided by the apolar form of the water dimer. Apolar water could be the medium which helps precursor molecules to concentrate and react more efficiently. The formation of prebiotic molecules could thus be linked to the structure of the water inside chimney nanochannels and cavities where hydrothermal piezochemistry and shock wave chemistry could occur. (letter to the editor)

  7. Contribution of water vapor pressure to pressurization of plutonium dioxide storage containers

    Science.gov (United States)

    Veirs, D. Kirk; Morris, John S.; Spearing, Dane R.

    2000-07-01

    Pressurization of long-term storage containers filled with materials meeting the US DOE storage standard is of concern.1,2 For example, temperatures within storage containers packaged according to the standard and contained in 9975 shipping packages that are stored in full view of the sun can reach internal temperatures of 250 °C.3 Twenty five grams of water (0.5 wt.%) at 250 °C in the storage container with no other material present would result in a pressure of 412 psia, which is limited by the amount of water. The pressure due to the water can be substantially reduced due to interactions with the stored material. Studies of the adsorption of water by PuO2 and surface interactions of water with PuO2 show that adsorption of 0.5 wt.% of water is feasible under many conditions and probable under high humidity conditions.4,5,6 However, no data are available on the vapor pressure of water over plutonium dioxide containing materials that have been exposed to water.

  8. Response of steam-water mixtures to pressure transients

    International Nuclear Information System (INIS)

    Hull, L.M.

    1985-01-01

    During the transition phase of a hypothetical core-disruptive accident in a liquid-metal fast breeder reactor, melting fuel-steel mixtures may begin to boil, resulting in a two-phase mixture of molten reactor fuel and steel vapor. Dispersal of this mixture by pressure transients may prevent recriticality of the fuel material. This paper describes the results of a series of experiments that investigated the response of two-phase mixtures to pressure transients. Simulant fluids (steam/water) were used in a transparent 10.2-cm-dia, 63.5-cm-long acrylic tube. The pressure transient was provided by releasing pressurized nitrogen from a supply tank. The data obtained are in the form of pressure-time records and high-speed movies. The varied parameters are initial void fraction (10% and 40%) and transient pressure magnitude (3.45 and 310 kPa)

  9. Quality analysis in pressurized water reactor fuel

    International Nuclear Information System (INIS)

    Darolles, J.F.

    1975-01-01

    An integrated system which has been set up to administrate and analyze the quality is described. This system is in actual operation. The basic principles for quality analysis system are traceability, i.e., identification, location and history of fuel components and quality evaluation during manufacturing. The quality analysis system operates in the following areas: data recording and transmission, data processing, quality file generation. The interest of such a system may be noted particularly in manufacturing, for the constitution of quality files, the design of products and the processing of data from irradiated fuel assemblies [fr

  10. The hydraulics of the pressurized water reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Barbier, D.; Caruso, A.

    1999-01-01

    The SFEN organized, the 10 june 1999 at Paris, a meeting in the domain of the PWR hydraulics and in particular the hydraulic phenomena concerning the vessel and the vapor generators. The papers presented showed the importance of the industrial stakes with their associated phenomena: cores performance and safety with the more homogenous cooling system, the rods and the control rods wear, the temperature control, the fluid-structure interactions. A great part was also devoted to the progresses in the domain of the numerical simulation and the models and algorithms qualification. (A.L.B.)

  11. Decay ratio estimation in pressurized water reactor

    International Nuclear Information System (INIS)

    Por, G.; Runkel, J.

    1990-11-01

    The well known decay ratio (DR) from stability analysis of boiling water reactors (BWR) is estimated from the impulse response function which was evaluated using a simplified univariate autoregression method. This simplified DR called modified DR (mDR) was applied on neutron noise measurements carried out during five fuel cycles of a 1300 MWe PWR. Results show that this fast evaluation method can be used for monitoring of the growing oscillation of the neutron flux during the fuel cycles which is a major concern of utilities in PWRs, thus it can be used for estimating safety margins. (author) 17 refs.; 10 figs

  12. 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.)

  13. Water Pressure Distribution on a Twin-Float Seaplane

    Science.gov (United States)

    Thompson, F L

    1930-01-01

    This is the second of a series of investigations to determine water pressure distribution on various types of seaplane floats and hulls, and was conducted on a twin-float seaplane. It consisted of measuring water pressures and accelerations on a TS-1 seaplane during numerous landing and taxiing maneuvers at various speeds and angles. The results show that water pressures as great as 10 lbs. per sq. in.may occur at the step in various maneuvers and that pressures of approximately the same magnitude occur at the stern and near the bow in hard pancake landings with the stern way down. At the other parts of the float the pressures are less and are usually zero or slightly negative for some distance abaft the step. A maximum negative pressure of 0.87 lb. Per square inch was measured immediately abaft the step. The maximum positive pressures have a duration of approximately one-twentieth to one-hundredth second at any given location and are distributed over a very limited area at any particular instant.

  14. High Pressure Electrochemical Oxygen Generation for ISS, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Giner, Inc. has developed an advanced high pressure electrochemical oxygen concentrator (EOC) that offers a simple alternative to the use of pressure swing...

  15. The effect of plutonium dioxide water surface coverage on the generation of hydrogen and oxygen

    Energy Technology Data Exchange (ETDEWEB)

    Veirs, Douglas K. [Los Alamos National Laboratory; Berg, John M. [Los Alamos National Laboratory; Crowder, Mark L. [Savannah River National Laboratory

    2012-06-20

    The conditions for the production of oxygen during radiolysis of water adsorbed onto plutonium dioxide powder are discussed. Studies in the literature investigating the radiolysis of water show that both oxygen and hydrogen can be generated from water adsorbed on high-purity plutonium dioxide powder. These studies indicate that there is a threshold in the amount of water below which oxygen is not generated. The threshold is associated with the number of monolayers of adsorbed water and is shown to occur at approximately two monolayers of molecularly adsorbed water. Material in equilibrium with 50% relative humidity (RH) will be at the threshold for oxygen generation. Using two monolayers of molecularly adsorbed water as the threshold for oxygen production, the total pressure under various conditions is calculated assuming stoichiometric production of hydrogen and oxygen. The specific surface area of the oxide has a strong effect on the final partial pressure. The specific surface areas resulting in the highest pressures within a 3013 container are evaluated. The potential for oxygen generation is mitigated by reduced relative humidity, and hence moisture adsorption, at the oxide surface which occurs if the oxide is warmer than the ambient air. The potential for oxygen generation approaches zero as the temperature difference between the ambient air and the material approaches 6 C.

  16. Climate change and the vulnerability of electricity generation to water stress in the European Union

    Science.gov (United States)

    Behrens, Paul; van Vliet, Michelle T. H.; Nanninga, Tijmen; Walsh, Brid; Rodrigues, João F. D.

    2017-08-01

    Thermoelectric generation requires large amounts of water for cooling. Recent warm periods have led to curtailments in generation, highlighting concerns about security of supply. Here we assess EU-wide climate impacts for 1,326 individual thermoelectric plants and 818 water basins in 2020 and 2030. We show that, despite policy goals and a decrease in electricity-related water withdrawal, the number of regions experiencing some reduction in power availability due to water stress rises from 47 basins to 54 basins between 2014 and 2030, with further plants planned for construction in stressed basins. We examine the reasons for these pressures by including water demand for other uses. The majority of vulnerable basins lie in the Mediterranean region, with further basins in France, Germany and Poland. We investigate four adaptations, finding that increased future seawater cooling eases some pressures. This highlights the need for an integrated, basin-level approach in energy and water policy.

  17. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

    Siler, W.C.; Marda, R.S.; Smith, W.R.

    1977-01-01

    Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

  18. Generational Differences in Resistance to Peer Pressure among Mexican-Origin Adolescents.

    Science.gov (United States)

    Umana-Taylor, Adriana J.; Bamaca-Gomez, Mayra Y.

    2003-01-01

    Examined whether Mexican origin adolescents who varied by generational status would differ in their resistance to peer pressure. After controlling for gender, resistance to peer pressure varied significantly by generational status. Adolescents with no familial births in the United States were significantly more resistant to peer pressure than…

  19. Prevention and mitigation of steam-generator water-hammer events in PWR plants

    International Nuclear Information System (INIS)

    Han, J.T.; Anderson, N.

    1982-11-01

    Water hammer in nuclear power plants is an unresolved safety issue under study at the NRC (USI A-1). One of the identified safety concerns is steam generator water hammer (SGWH) in pressurized-water reactor (PWR) plants. This report presents a summary of: (1) the causes of SGWH; (2) various fixes employed to prevent or mitigate SGWH; and (3) the nature and status of modifications that have been made at each operating PWR plant. The NRC staff considers that the issue of SGWH in top feedring designs has been technically resolved. This report does not address technical findings relevant to water hammer in preheat type steam generators. 10 figures, 2 tables

  20. Prevention and mitigation of steam generator water hammer events in PWRs

    International Nuclear Information System (INIS)

    Han, J.T.; Anderson, N.

    1983-01-01

    Water hammer in nuclear power plants is an unresolved safety issue under study by the Nuclear Regulatory Commission (NRC). This article summarizes (1) the causes of steam generator water hammer (SGWH) events in pressurized-water reactors (PWRs), (2) various methods used to prevent or mitigate SGWH events, and (3) modifications that have been made at each operating PWR. The NRC staff considers the issue of SGWH in top feedring designs to be technically resolved. This article does not address technical findings relevant to water hammer in preheat-type steam generators

  1. Evaluating the Laplace pressure of water nanodroplets from simulations

    Science.gov (United States)

    Malek, Shahrazad M. A.; Sciortino, Francesco; Poole, Peter H.; Saika-Voivod, Ivan

    2018-04-01

    We calculate the components of the microscopic pressure tensor as a function of radial distance r from the centre of a spherical water droplet, modelled using the TIP4P/2005 potential. To do so, we modify a coarse-graining method for calculating the microscopic pressure (Ikeshoji et al 2003 Mol. Simul. 29 101) in order to apply it to a rigid molecular model of water. As test cases, we study nanodroplets ranging in size from 776 to 2880 molecules at 220 K. Beneath a surface region comprising approximately two molecular layers, the pressure tensor becomes approximately isotropic and constant with r. We find that the dependence of the pressure on droplet radius is that expected from the Young-Laplace equation, despite the small size of the droplets.

  2. Direct Drive Generator for Renewable Power Conversion from Water Currents

    International Nuclear Information System (INIS)

    Segergren, Erik

    2005-01-01

    In this thesis permanent magnet direct drive generator for power conversion from water currents is studied. Water currents as a power source involves a number of constrains as well as possibilities, especially when direct drive and permanent magnets are considered. The high power fluxes and low current velocities of a water current, in combination with its natural variations, will affect the way the generator is operated and, flowingly, the appearance of the generator. The work in this thesis can, thus, be categorized into two general topics, generator technology and optimization. Under the first topic, fundamental generator technology is used to increase the efficiency of a water current generator. Under the latter topic, water current generators are optimized to a specific environment. The conclusion drawn from this work is that it is possible to design very low speed direct drive generators with good electromagnetic properties and wide efficiency peak

  3. Noise investigations in boiling water and pressurized water reactors

    International Nuclear Information System (INIS)

    Kosaly, G.

    1979-08-01

    The paper begins with the general discussion of the neutron noise measured by in-core and ex-core detectors. The theory of BWR noise is surveyed in detail. After the theoretical survey the physical background of the measuring methods applied in the fluctuation analysis of BWRs is discussed. The section on PWR noise discusses the information content of the neutron, pressure and displacement signals about the vibration of the vessel and its internals. (author)

  4. Pneumatic pressure wave generator provides economical, simple testing of pressure transducers

    Science.gov (United States)

    Gaal, A. E.; Weldon, T. P.

    1967-01-01

    Testing device utilizes the change in pressure about a bias or reference pressure level produced by displacement of a center-driven piston in a closed cylinder. Closely controlled pneumatic pressure waves allow testing under dynamic conditions.

  5. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  6. Biogas generation and water management; Biogaserzeugung und Wasserwirtschaft

    Energy Technology Data Exchange (ETDEWEB)

    Fohrmann, Reinhard [IWW Rheinisch-Westfaelisches Institut fuer Wasser Beratungs- und Entwicklungsgesellschaft mbH, Muelheim an der Ruhr (Germany)

    2012-12-15

    Is it possible to attenuate the consequences of the increasing development of biogas generation for the water production by means of a meshing of energy industry and water management? On the occasion of the 25th Muelheim water technology symposium, experts discuss on opportunities and risks of the biogas generation.

  7. Design of virtual SCADA simulation system for pressurized water reactor

    International Nuclear Information System (INIS)

    Wijaksono, Umar; Abdullah, Ade Gafar; Hakim, Dadang Lukman

    2016-01-01

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor

  8. Design of virtual SCADA simulation system for pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wijaksono, Umar, E-mail: umar.wijaksono@student.upi.edu; Abdullah, Ade Gafar; Hakim, Dadang Lukman [Electrical Power System Research Group, Department of Electrical Engineering Education, Jl. Dr. Setiabudi No. 207 Bandung, Indonesia 40154 (Indonesia)

    2016-02-08

    The Virtual SCADA system is a software-based Human-Machine Interface that can visualize the process of a plant. This paper described the results of the virtual SCADA system design that aims to recognize the principle of the Nuclear Power Plant type Pressurized Water Reactor. This simulation uses technical data of the Nuclear Power Plant Unit Olkiluoto 3 in Finland. This device was developed using Wonderware Intouch, which is equipped with manual books for each component, animation links, alarm systems, real time and historical trending, and security system. The results showed that in general this device can demonstrate clearly the principles of energy flow and energy conversion processes in Pressurized Water Reactors. This virtual SCADA simulation system can be used as instructional media to recognize the principle of Pressurized Water Reactor.

  9. Microwave measurements of water vapor partial pressure at high temperatures

    International Nuclear Information System (INIS)

    Latorre, V.R.

    1991-01-01

    One of the desired parameters in the Yucca Mountain Project is the capillary pressure of the rock comprising the repository. This parameter is related to the partial pressure of water vapor in the air when in equilibrium with the rock mass. Although there are a number of devices that will measure the relative humidity (directly related to the water vapor partial pressure), they generally will fail at temperatures on the order of 150C. Since thee author has observed borehole temperatures considerably in excess of this value in G-Tunnel at the Nevada Test Site (NTS), a different scheme is required to obtain the desired partial pressure data at higher temperatures. This chapter presents a microwave technique that has been developed to measure water vapor partial pressure in boreholes at temperatures up to 250C. The heart of the system is a microwave coaxial resonator whose resonant frequency is inversely proportional to the square root of the real part of the complex dielectric constant of the medium (air) filling the resonator. The real part of the dielectric constant of air is approximately equal to the square of the refractive index which, in turn, is proportional to the partial pressure of the water vapor in the air. Thus, a microwave resonant cavity can be used to measure changes in the relative humidity or partial pressure of water vapor in the air. Since this type of device is constructed of metal, it is able to withstand very high temperatures. The actual limitation is the temperature limit of the dielectric material in the cable connecting the resonator to its driving and monitoring equipment-an automatic network analyzer in our case. In the following sections, the theory of operation, design, construction, calibration and installation of the microwave diagnostics system is presented. The results and conclusions are also presented, along with suggestions for future work

  10. Method and device for feeding purified water to a pressure vessel

    International Nuclear Information System (INIS)

    Hirato, Miharu.

    1982-01-01

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

  11. A computer program for accident calculations of a standard pressurized water reactor

    International Nuclear Information System (INIS)

    Keutner, H.

    1979-01-01

    In this computer program the dynamic of a standard pressurized water reactor should be realized by both circulation loops with all important components. All important phenomena are taken into consideration, which appear for calculation of disturbances in order to state a realistic process for some minutes after a disturbance or a desired change of condition. In order to optimize the computer time simplifications are introduced in the statement of a differential-algebraic equalization system such that all important effects are taken into consideration. The model analysis starts from the heat production of the fuel rod via cladding material to the cooling medium water and considers the delay time from the core to the steam generator. Alternations of the cooling medium pressure as well as the different temperatures in the primary loop influence the pressuring system - the pressurizer - which is realized by a water and a steam zone with saturated and superheated steam respectively saturated and undercooled water with injection, heating and blow-down devices. The bilance of the steam generator to the secondary loop realizes the process engineering devices. Thereby the control regulation of the steam pressure and the reactor performance is realized. (orig.) [de

  12. Experiments on aerosol removal by high-pressure water spray

    Energy Technology Data Exchange (ETDEWEB)

    Corno, Ada del, E-mail: delcorno@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Morandi, Sonia, E-mail: morandi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Parozzi, Flavio, E-mail: parozzi@rse-web.it [RSE, Power Generation Technologies and Materials Dept, via Rubattino 54, I-20134 Milano (Italy); Araneo, Lucio, E-mail: lucio.araneo@polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy); CNR-IENI, via Cozzi 53, I-20125 Milano (Italy); Casella, Francesco, E-mail: francesco2.casella@mail.polimi.it [Politecnico di Milano, Department of Energy, via Lambruschini 4A, I-20156 Milano (Italy)

    2017-01-15

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m{sup 3}. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO{sub 2} particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m{sup 3}. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was

  13. Light-water reactor pressure vessel surveillance standards

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

  14. The inner containment of an EPR trademark pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

    2014-08-15

    On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

  15. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert

    2017-09-14

    The present invention discloses a system and method for supercritical water gasification (SCWG) of biomass materials wherein the system includes a SCWG reactor and a plurality of heat exchangers located within a shared pressurized vessel, which decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed and fabricated in manner that would support commercial scaled-up SCWG operations. By using heat exchangers coupled to the reactor in a series configuration, significant efficiencies are achieved by the present invention SCWG system over prior known SCWG systems.

  16. Experiments on aerosol removal by high-pressure water spray

    International Nuclear Information System (INIS)

    Corno, Ada del; Morandi, Sonia; Parozzi, Flavio; Araneo, Lucio; Casella, Francesco

    2017-01-01

    Highlights: • Experimental research to measure the efficiency of high-pressure sprays in capturing aerosols if applied to a filtered containment venting system in case of severe accident. • Cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration in the range 2–90 mg/m"3. • Carried out in a chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls equipped with a high pressure water spray with single nozzle. • Respect to low-pressure sprays, removal efficiency turned out significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure sprays system. - Abstract: An experimental research was managed in the framework of the PASSAM European Project to measure the efficiency of high-pressure sprays in capturing aerosols when applied to a filtered containment venting system in case of severe accident. The campaign was carried out in a purposely built facility composed by a scrubbing chamber 0.5 × 1.0 m and 1.5 m high, with transparent walls to permit the complete view of the aerosol removal process, where the aerosol was injected to form a cloud of specific particle concentration. The chamber was equipped with a high pressure water spray system with a single nozzle placed on its top. The test matrix consisted in the combination of water pressure injections, in the range 50–130 bar, on a cloud of monodispersed SiO_2 particles with sizes 0.5 or 1.0 μm and initial concentration ranging between 2 and 99 mg/m"3. The spray was kept running for 2 min and the efficiency of the removal was evaluated, along the test time, using an optical particle sizer. With respect to low-pressure sprays, the removal efficiency turned out much more significant: the half-life for 1 μm particles with a removal high-pressure spray system is orders of magnitude shorter than that with a low-pressure spray system. The highest removal rate was detected with 1

  17. A novel coordinated control for Integrated Pressurized Water Reactor

    International Nuclear Information System (INIS)

    Zhao, Yuxin; Du, Xue; Xia, Genglei; Gao, Feng

    2015-01-01

    Highlights: • Proposed IPWR coordinated control strategy to avoid flow instability of OTSG. • Tuned PID controller parameters by Fuzzy kernel wavelet neural network with kernel trick and adaptive variable step-size. • Transition process exhibit the effectiveness of the novel IPWR control system. - Abstract: Integrated Pressurized Water Reactor (IPWR) has the characteristic of strong coupling, nonlinearity and complicated dynamic performance, which requires high standards of the control strategy and controller design. Most of IPWR systems utilize control strategy of ideal steady-state and PID controller, even though this strategy causes flow instability in the once through steam generator (OTSG) in low load conditions. Besides, the simple form of PID limits the performance developing which could not appropriately satisfy the requirements for quality. Motivated by these drawbacks, this paper proposes an IPWR coordinated control strategy and adopts PID controller to control each subsystem. The control strategy considers the system as a two-level hierarchical control system, and considers coordinating controller and bottom controllers. In the period of controller design, this strategy utilizes PID controller to control each subsystem, and modifies the controller parameters in real time by Fuzzy-KWNN algorithm, which adaptively achieves the system adjustment. Finally, simulation results are presented to exhibit the effectiveness of the proposed IPWR control system

  18. Methodology for surge pressure evaluation in a water injection system

    Energy Technology Data Exchange (ETDEWEB)

    Meliande, Patricia; Nascimento, Elson A. [Universidade Federal Fluminense (UFF), Niteroi, RJ (Brazil). Dept. de Engenharia Civil; Mascarenhas, Flavio C.B. [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Lab. de Hidraulica Computacional; Dandoulakis, Joao P. [SHELL of Brazil, Rio de Janeiro, RJ (Brazil)

    2009-07-01

    Predicting transient effects, known as surge pressures, is of high importance for offshore industry. It involves detailed computer modeling that attempts to simulate the complex interaction between flow line and fluid in order to ensure efficient system integrity. Platform process operators normally raise concerns whether the water injection system is adequately designed or not to be protected against possible surge pressures during sudden valve closure. This report aims to evaluate the surge pressures in Bijupira and Salema water injection systems due to valve closure, through a computer model simulation. Comparisons among the results from empirical formulations are discussed and supplementary analysis for Salema system were performed in order to define the maximum volumetric flow rate for which the design pressure was able to withstand. Maximum surge pressure values of 287.76 bar and 318.58 bar, obtained in Salema and Bijupira respectively, using empirical formulations have surpassed the operating pressure design, while the computer model results have pointed the greatest surge pressure value of 282 bar in Salema system. (author)

  19. Standard Technical Specifications for Westinghouse pressurized water reactors

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1980-09-01

    The Standard Technical Specifications for Westinghouse Pressurized Water Reactors (W-STS) is a generic document prepared by the U.S. NRC for use in the licensing process of current Westinghouse Pressurized Water Reactors. The W-STS sets forth the Limits, Operating Conditions and other requirements applicable to nuclear reactor facility operation as set forth in by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. This document is revised periodically to reflect current licensing requirements

  20. Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Vito, D.J.

    1980-12-01

    The Standard Technical Specifications for Combustion Engineering Pressurized Water Reactors (CE-STS) is a generic document prepared by the US NRC for use in the licensing process of current Combustion Engineering Pressurized Water Reactors. The CE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

  1. The future 700 MWe pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Bhardwaj, S.A.

    2006-01-01

    The design of a 700 MWe pressurized heavy water reactor has been developed. The design is based on the twin 540 MWe reactors at Tarapur of which the first unit has been made critical in less than 5 years from construction commencement. In the 700 MWe design boiling of the coolant, to a limited extent, has been allowed near the channel exit. While making the plant layout more compact, emphasis has been on constructability. Saving in capital cost of about 15%, over the present units, is expected. The paper describes salient design features of 700 MWe pressurized heavy water reactor

  2. Application-specific integrated circuit design for a typical pressurized water reactor pressure channel trip

    International Nuclear Information System (INIS)

    Battle, R.E.; Manges, W.W.; Emery, M.S.; Vendermolen, R.I.; Bhatt, S.

    1994-01-01

    This article discusses the use of application-specific integrated circuits (ASICs) in nuclear plant safety systems. ASICs have certain advantages over software-based systems because they can be simple enough to be thoroughly tested, and they can be tailored to replace existing equipment. An architecture to replace a pressurized water reactor pressure channel trip is presented. Methods of implementing digital algorithms are also discussed

  3. Observations on steam generator water hammer

    International Nuclear Information System (INIS)

    Hildebrandt, P.C.

    1979-01-01

    The present state of understanding of water hammer phenomena apparently does not permit analytically predicting the occurrence or consequence of water hammer, with any confidence, or conclusively evaluating the benefit associated with any particular action taken to ameliorate an existing or potential water hammer problem. Features or fixes are incorporated based on overwhelming the problem, and trial and error

  4. Water pressure and ground vibrations induced by water guns at a backwater pond on the Illinois River near Morris, Illinois

    Science.gov (United States)

    Koebel, Carolyn M.; Egly, Rachel M.

    2016-09-27

    Three different geophysical sensor types were used to characterize the underwater pressure waves and ground velocities generated by the underwater firing of seismic water guns. These studies evaluated the use of water guns as a tool to alter the movement of Asian carp. Asian carp are aquatic invasive species that threaten to move into the Great Lakes Basin from the Mississippi River Basin. Previous studies have identified a threshold of approximately 5 pounds per square inch (lb/in2) for behavioral modification and for structural limitation of a water gun barrier.Two studies were completed during August 2014 and May 2015 in a backwater pond connected to the Illinois River at a sand and gravel quarry near Morris, Illinois. The August 2014 study evaluated the performance of two 80-cubic-inch (in3) water guns. Data from the 80-in3 water guns showed that the pressure field had the highest pressures and greatest extent of the 5-lb/in2 target value at a depth of 5 feet (ft). The maximum recorded pressure was 13.7 lb/in2, approximately 25 ft from the guns. The produced pressure field took the shape of a north-south-oriented elongated sphere with the 5-lb/in2 target value extending across the entire study area at a depth of 5 ft. Ground velocities were consistent over time, at 0.0067 inches per second (in/s) in the transverse direction, 0.031 in/s in the longitudinal direction, and 0.013 in/s in the vertical direction.The May 2015 study evaluated the performance of one and two 100-in3 water guns. Data from the 100-in3 water guns, fired both individually and simultaneously, showed that the pressure field had the highest pressures and greatest extent of the 5-lb/in2 target value at a depth of 5 ft. The maximum pressure was 57.4 lb/in2, recorded at the underwater blast sensor closest to the water guns (at a horizontal distance of approximately 3 ft), as two guns fired simultaneously. Pressures and extent of the 5-lb/in2 target value decrease above and below this 5-ft depth

  5. Nonlinear vibration of a hemispherical dome under external water pressure

    International Nuclear Information System (INIS)

    Ross, C T F; McLennan, A; Little, A P F

    2011-01-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  6. Nonlinear vibration of a hemispherical dome under external water pressure

    Science.gov (United States)

    Ross, C. T. F.; McLennan, A.; Little, A. P. F.

    2011-07-01

    The aim of this study was to analyse the behaviour of a hemi-spherical dome when vibrated under external water pressure, using the commercial computer package ANSYS 11.0. In order to achieve this aim, the dome was modelled and vibrated in air and then in water, before finally being vibrated under external water pressure. The results collected during each of the analyses were compared to the previous studies, and this demonstrated that ANSYS was a suitable program and produced accurate results for this type of analysis, together with excellent graphical displays. The analysis under external water pressure, clearly demonstrated that as external water pressure was increased, the resonant frequencies decreased and a type of dynamic buckling became likely; because the static buckling eigenmode was similar to the vibration eigenmode. ANSYS compared favourably with the in-house software, but had the advantage that it produced graphical displays. This also led to the identification of previously undetected meridional modes of vibration; which were not detected with the in-house software.

  7. Improved water density feedback model for pressurized water reactors

    International Nuclear Information System (INIS)

    Casadei, A.L.

    1976-01-01

    An improved water density feedback model has been developed for neutron diffusion calculations of PWR cores. This work addresses spectral effects on few-group cross sections due to water density changes, and water density predictions considering open channel and subcooled boiling effects. An homogenized spectral model was also derived using the unit assembly diffusion method for employment in a coarse mesh 3D diffusion computer program. The spectral and water density evaluation models described were incorporated in a 3D diffusion code, and neutronic calculations for a typical PWR were completed for both nominal and accident conditions. Comparison of neutronic calculations employing the open versus the closed channel model for accident conditions indicates that significant safety margin increases can be obtained if subcooled boiling and open channel effects are considered in accident calculations. This is attributed to effects on both core reactivity and power distribution, which result in increased margin to fuel degradation limits. For nominal operating conditions, negligible differences in core reactivity and power distribution exist since flow redistribution and subcooled voids are not significant at such conditions. The results serve to confirm the conservatism of currently employed closed channel feedback methods in accident analysis, and indicate that the model developed in this work can contribute to show increased safety margins for certain accidents

  8. Pressure wave propagation in the discharge piping with water pool

    International Nuclear Information System (INIS)

    Bang, Young S.; Seul, Kwang W.; Kim, In Goo

    2004-01-01

    Pressure wave propagation in the discharge piping with a sparger submerged in a water pool, following the opening of a safety relief valve, is analyzed. To predict the pressure transient behavior, a RELAP5/MOD3 code is used. The applicability of the RELAP5 code and the adequacy of the present modeling scheme are confirmed by simulating the applicable experiment on a water hammer with voiding. As a base case, the modeling scheme was used to calculate the wave propagation inside a vertical pipe with sparger holes and submerged within a water pool. In addition, the effects on wave propagation of geometric factors, such as the loss coefficient, the pipe configuration, and the subdivision of sparger pipe, are investigated. The effects of inflow conditions, such as water slug inflow and the slow opening of a safety relief valve are also examined

  9. Investigating the water consumption for electricity generation at Turkish power plants

    Science.gov (United States)

    El-Khozondar, Balkess; Aydinalp Koksal, Merih

    2017-11-01

    The water-energy intertwined relationship has recently gained more importance due to the high water consumption in the energy sector and to the limited availability of the water resources. The energy and electricity demand of Turkey is increasing rapidly in the last two decades. More thermal power plants are expected to be built in the near future to supply the rapidly increasing demand in Turkey which will put pressure on water availability. In this study, the water consumption for electricity generation at Turkish power plants is investigated. The main objectives of this study are to identify the amount of water consumed to generate 1 kWh of electricity for each generation technology currently used in Turkey and to investigate ways to reduce the water consumption at power plants expected to be built in the near future to supply the increasing demand. The various electricity generation technology mixture scenarios are analyzed to determine the future total and per generation water consumption, and water savings based on changes of cooling systems used for each technology. The Long-range Energy Alternatives Planning (LEAP) program is used to determine the minimum water consuming electricity generation technology mixtures using optimization approaches between 2017 and 2035.

  10. Analytical and computational methodology to assess the over pressures generated by a potential catastrophic failure of a cryogenic pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Zamora, I.; Fradera, J.; Jaskiewicz, F.; Lopez, D.; Hermosa, B.; Aleman, A.; Izquierdo, J.; Buskop, J.

    2014-07-01

    Idom has participated in the risk evaluation of Safety Important Class (SIC) structures due to over pressures generated by a catastrophic failure of a cryogenic pressure vessel at ITER plant site. The evaluation implements both analytical and computational methodologies achieving consistent and robust results. (Author)

  11. Analytical and computational methodology to assess the over pressures generated by a potential catastrophic failure of a cryogenic pressure vessel

    International Nuclear Information System (INIS)

    Zamora, I.; Fradera, J.; Jaskiewicz, F.; Lopez, D.; Hermosa, B.; Aleman, A.; Izquierdo, J.; Buskop, J.

    2014-01-01

    Idom has participated in the risk evaluation of Safety Important Class (SIC) structures due to over pressures generated by a catastrophic failure of a cryogenic pressure vessel at ITER plant site. The evaluation implements both analytical and computational methodologies achieving consistent and robust results. (Author)

  12. Flashing of high-pressure saturated water into the pool water

    International Nuclear Information System (INIS)

    Takamasa, Tomoji; Kondo, Koichi; Aya, Izuo.

    1997-01-01

    This paper presents an experimental study on a saturated high-pressure water discharging into a water pool. The purpose of the experiment is to clarify the phenomena that occur by a blow-down of the water from the pressure vessel into the water-filled containment in the case of a wall-crack accident or a LOCA in a passive safety reactor. The results show that a flashing oscillation (FO) occurs when the water discharges into the pool, under specified experimental conditions. The range of the flashing location oscillates between a point very close to and some distance away from the vent hole. The pressures in the vent tube and water pool constantly fluctuate due to the flashing oscillation. The pressure oscillation and alternating flashing location might be caused by the balancing action between the supply of saturated water, flashing at the control volume and steam condensation on the steam-water interface. The frequencies of FO, or frequencies of pressure oscillation and alternating flashing location, increased as water subcooling increased, and as discharging pressure and vent hole diameter decreased. A linear analysis was conducted using a spherical flashing bubble model in which the motion of bubble is controlled by steam condensation. The effects of these parameters on the period of FO in the experiments can be predicted well by the analysis. (author)

  13. TRANP - a computer code for digital simulation of steady - state and transient behavior of a pressurizer water reactor primary circuit

    International Nuclear Information System (INIS)

    Chalhoub, E.S.

    1980-09-01

    A digital computer code TRANP was developed to simulate the steady-state and transient behavior of a pressurizer water reactor primary circuit. The development of this code was based on the combining of three codes already developed for the simulation of a PWR core, a pressurizer, a steam generator and a main coolant pump, representing the primary circuit components. (Author) [pt

  14. Ultra-high pressure water jet: Baseline report

    International Nuclear Information System (INIS)

    1997-01-01

    The ultra-high pressure waterjet technology was being evaluated at Florida International University (FIU) as a baseline technology. In conjunction with FIU's evaluation of efficiency and cost, this report covers the evaluation conducted for safety and health issues. It is a commercially available technology and has been used for various projects at locations throughout the country. The ultra-high pressure waterjet technology acts as a cutting tool for the removal of surface substrates. The Husky trademark pump feeds water to a lance that directs the high pressure water at the surface to be removed. The safety and health evaluation during the testing demonstration focused on two main areas of exposure. These were dust and noise. The dust exposure was found to be minimal, which would be expected due to the wet environment inherent in the technology, but noise exposure was at a significant level. Further testing for noise is recommended because of the outdoor environment where the testing demonstration took place. In addition, other areas of concern found were arm-hand vibration, ergonomics, heat stress, tripping hazards, electrical hazards, lockout/tagout, fall hazards, slipping hazards, hazards associated with the high pressure water, and hazards associated with air pressure systems

  15. Solid polymer electrolyte water electrolysis system development. [to generate oxygen for manned space station applications

    Science.gov (United States)

    1975-01-01

    Solid polymer electrolyte technology used in a water electrolysis system (WES) to generate oxygen and hydrogen for manned space station applications was investigated. A four-man rated, low pressure breadboard water electrolysis system with the necessary instrumentation and controls was fabricated and tested. A six man rated, high pressure, high temperature, advanced preprototype WES was developed. This configuration included the design and development of an advanced water electrolysis module, capable of operation at 400 psig and 200 F, and a dynamic phase separator/pump in place of a passive phase separator design. Evaluation of this system demonstrated the goal of safe, unattended automated operation at high pressure and high temperature with an accumulated gas generation time of over 1000 hours.

  16. Iodine generator for reclaimed water purification

    Science.gov (United States)

    Wynveen, R. A.; Powell, J. D.; Schubert, F. H. (Inventor)

    1977-01-01

    The system disclosed is for controlling the iodine level in a water supply in a spacecraft. It includes an iodine accumulator which stores crystalline iodine, an electrochemical valve to control the input of iodine to the drinking water and an iodine dispenser. A pump dispenses fluid through the iodine dispenser and an iodine sensor to a potable water tank storage. The iodine sensor electronically detects the iodine level in the water, and through electronic means, produces a correction current control. The correction current control operates the electro-chemical iodine valve to release iodine from the iodine accumulator into the iodine dispenser.

  17. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  18. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  19. Sodium-water reaction studies for MONJU steam generators

    International Nuclear Information System (INIS)

    Hori, M.; Sato, M.; Nei, H.; Harasaki, T.; Hishida, M.; Saito, T.

    1975-01-01

    The R and D results of the PNC's sodium-water reaction project are reviewed. The purposes of the project with the specific object for each test rig and computer code are given. The test items which should be investigated for the safety evaluation of the MONJU steam generators are discussed, and the status of the PNC's work on each item is described. The results on the small-leak wastage measurement are shown and the improved experimental equations to predict the wastage rate from the leak rate and the sodium temperature are given. The preliminary results on the wastage of tube bundle in the intermediate leak range are shown. The depth and the area of the wastage and also the wastage rate for each tube are shown graphically. The measured peak value of the initial pressure spike for the large leak is shown. The scatter of the data and its causes are discussed. The bubble growth rate estimated from the void probe measurement is presented. The results of the simulation experiment on the pressure wave propagation to the secondary circuit are given, comparing them with the prediction by the one-dimensional computer codes SWAC-5K and SWAC-5H. (author)

  20. Sodium-water reaction studies for MONJU steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Hori, M; Sato, M; Nei, H; Harasaki, T; Hishida, M; Saito, T

    1975-07-01

    The R and D results of the PNC's sodium-water reaction project are reviewed. The purposes of the project with the specific object for each test rig and computer code are given. The test items which should be investigated for the safety evaluation of the MONJU steam generators are discussed, and the status of the PNC's work on each item is described. The results on the small-leak wastage measurement are shown and the improved experimental equations to predict the wastage rate from the leak rate and the sodium temperature are given. The preliminary results on the wastage of tube bundle in the intermediate leak range are shown. The depth and the area of the wastage and also the wastage rate for each tube are shown graphically. The measured peak value of the initial pressure spike for the large leak is shown. The scatter of the data and its causes are discussed. The bubble growth rate estimated from the void probe measurement is presented. The results of the simulation experiment on the pressure wave propagation to the secondary circuit are given, comparing them with the prediction by the one-dimensional computer codes SWAC-5K and SWAC-5H. (author)

  1. Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

    2012-02-15

    Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

  2. Variations of free gas content in water during pressure fluctuations

    International Nuclear Information System (INIS)

    Keller, A.; Zielke, W.

    1977-01-01

    In this paper an experimental programme is described in order to determine the influence of the cavitation nuclei distribution on cavitation inception. This programme has been used to measure air bubbles dimensions and number and particularly to determine the influence of quick pressure variations on the size on the number of bubbles in a pipe. An optical device counting scattered light is used as a measuring technique. Gas bubbles go through an optical control volume where they receive a high intensity light beam and scatter the light, then led to a photomultiplier; the signals are sorted and counted according to their size. If the number of nuclei, the dimensions of the control volume and the velocity of the water are known, it is possible to determine bubbles concentrations and the bulk modulus of the water. This measuring technique has been applied to a flow in a 140 mm diameter pipe with quick pressure variations from 2 bar to 0-10 bar. During the variations, the void fraction depends on the Reynolds number of the flow and on the gas content of the water. The bulk modulus has been computed with different conditions. Most results concern pressures slightly over the vapor pressure. Air content has a strong influence on cavitation and on water compressibility after a vapor cavity collapse

  3. Alkali free hydrolysis of sodium borohydride for hydrogen generation under pressure

    Energy Technology Data Exchange (ETDEWEB)

    Ferreira, M.J.F.; Pinto, A.M.F.R. [Centro de Estudos de Fenomenos de Transporte, Departamento de Engenharia Quimica, Faculdade de Engenharia da Universidade do Porto, Rua Dr. Roberto Frias s/n, 4200-465 Porto (Portugal); Gales, L. [Instituto de Biologia Molecular e Celular, Universidade do Porto, Rua do Campo Alegre 823, 4150-180 Porto and Instituto de Ciencias Biomedicas Abel Salazar, Largo Prof. Abel Salazar 2, 4099-003 Porto (Portugal); Fernandes, V.R.; Rangel, C.M. [Laboratorio Nacional de Energia e Geologia - LNEG, Fuel Cells and Hydrogen Unit Estrada do Paco do Lumiar 22, 1649-038 Lisboa (Portugal)

    2010-09-15

    The present study is related with the production of hydrogen gas (H{sub 2}), at elevated pressures and with high gravimetric storage density, to supply a PEM fuel cell on-demand. To achieve this goal, solid sodium borohydride (NaBH{sub 4}) was mixed with a proper amount of a powder reused nickel-ruthenium based catalyst (Ni-Ru based/NaBH{sub 4}: 0.2 and 0.4 g/g; {approx}150 times reused) inside the bottom of a batch reactor. Then, a stoichiometric amount of pure liquid water (H{sub 2}O/NaBH{sub 4}: 2-8 mol/mol) was added and the catalyzed NaBH{sub 4} hydrolysis evolved, in the absence of an alkali inhibitor. In this way, this research work is designated alkali free hydrolysis of NaBH{sub 4} for H{sub 2} generation. This type of hydrolysis is excellent from an environmental point of view because it does not involve strongly caustic solutions. Experiments were performed in three batch reactors with internal volumes 646, 369 and 229 cm{sup 3}, and having different bottom geometries (flat and conical shapes). The H{sub 2} generated was a function of the added water and completion was achieved with H{sub 2}O/NaBH{sub 4} = 8 mol/mol. The results show that hydrogen yields and rates increase remarkably increasing both system temperature and pressure. Reactor bottom shape influences deeply H{sub 2} generation: the conical bottom shape greatly enhances the rate and practically eliminates the reaction induction time. Our system of compressed hydrogen generation up to 1.26 MPa shows 6.3 wt% and 70 kg m{sup -3}, respectively, for gravimetric and volumetric hydrogen storage capacities (materials-only basis) and therefore is a viable hydrogen storage candidate for portable applications. (author)

  4. Absorber rod bundle actuator in a pressurized water nuclear reactor

    International Nuclear Information System (INIS)

    Martin, J.; Peletan, R.

    1984-01-01

    The invention concerns an absorber rod bundle actuator in a pressurized water reactor with spectral shift control. The device comprises two coaxial control bars. The inner bar is integral with the absorber rod bundle; it has an enlarged zone which acts as a proton under pressure difference across an annular seal which can be radially expanded, the pressure difference allowing to the absorber rod bundles actuating on the piston. When a pressure difference is applied, the seal expands radially by a sufficient amount to make sealing contact with the zone of larger diameter in the outer bar. The invention applies more particularly to reactors with spectral shift control using bundles of fertile rods [fr

  5. Development test procedure High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    Development testing will be performed on the water jet cleaning fixture to determine the most effective arrangement of water jet nozzles to remove contamination from the surfaces of canisters and other debris. The following debris may be stained with dye to simulate surface contaminates: Mark O, Mark I, and Mark II Fuel Storage Canisters (both stainless steel and aluminum), pipe of various size, (steel, stainless, carbon steel and aluminum). Carbon steel and stainless steel plate, channel, angle, I-beam and other surfaces, specifically based on the Scientific Ecology Group (SEG) inventory and observations of debris within the basin. Test procedure for developmental testing of High Pressure Water Jet System

  6. 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

  7. Structural analysis of fuel rod applied to pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Danilo P.; Pinheiro, Andre Ricardo M.; Lotto, André A., E-mail: danilo.pinheiro@marinha.mil.br [Centro Tecnológico da Marinha em São Paulo (CTMSP), São Paulo, SP (Brazil)

    2017-07-01

    The design of fuel assemblies applied to Pressurized Water Reactors (PWR) has several requirements and acceptance criteria that must be attended for licensing. In the case of PWR fuel rods, an important mechanical structural requirement is to keep the radial stability when submitted to the coolant external pressure. In the framework of the Accident Tolerant Fuel (ATF) program new materials have been studied to replace zirconium based alloys as cladding, including iron-based alloys. In this sense, efforts have been made to evaluate the behavior of these materials under PWR conditions. The present work aims to evaluate the collapse cold pressure of a stainless steel thin-walled tube similar to that used as cladding material of fuel rods by means of the comparison of numeric data, and experimental results. As a result of the simulations, it was observed that the collapse pressure has a value intermediate value between those found by regulatory requirements and analytical calculations. The experiment was carried out for the validation of the computational model using test specimens of thin-walled tubes considering empty tube. The test specimens were sealed at both ends by means of welding. They were subjected to a high pressure device until the collapse of the tubes. Preliminary results obtained from experiments with the empty test specimens indicate that the computational model can be validated for stainless steel cladding, considering the difference between collapse pressure indicated in the regulatory document and the actual limit pressure concerning to radial instability of tubes with the studied characteristics. (author)

  8. High Pressure Electrochemical Oxygen Generation for ISS, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — Giner, Inc. has developed an advanced electrochemical static vapor feed oxygen (O2) concentrator (SVFOC) that offers a simple alternative to the use of pressure...

  9. Water releasing electric generating device for nuclear power plant

    International Nuclear Information System (INIS)

    Umehara, Toshihiro; Tomohara, Yasutaka; Usui, Yoshihiko.

    1994-01-01

    Warm sea water discharged after being used for cooling in an equipment of a coastal nuclear powder plant is discharged from a water discharge port to a water discharge pit, and a conduit vessel is disposed in front of the water discharge port for receiving overflown warm sea water. The warm sea water taken to the conduit vessel is converted to a fallen flow and charged to a turbine generator under water, and electric power is generated by the water head energy of the fallen flow before it is discharged to the water discharge pit. The conduit vessel incorporates a foam preventing unit having spiral flow channels therein, so that the warm sea water taken to the conduit vessel is flown into the water discharge pit after consuming the water head energy while partially branched and flown downwardly and gives lateral component to the downwarding flowing direction. Then, warm sea water is made calm when it is flown into the water discharge pit and, accordingly, generation of bubbles on the water surface of the water discharge pit is avoided. (N.H.)

  10. Water treatment for fossil fuel power generation - technology status report

    International Nuclear Information System (INIS)

    2006-01-01

    This technology status report focuses on the use of water treatment technology in fossil fuel power plants. The use of polymeric ion exchange resins for deionization of water, the currently preferred use of ion exchange for economically treating water containing low dissolved salts, the use of low pressure high-flux membranes, membrane microfiltration, and reverse osmosis are discussed. Details are given of the benefits of the technologies, water use at power plants, the current status of water treatment technologies, and the potential for future developments, along with power plant market trends and potentials, worldwide developments, and UK capabilities in water treatment plant design and manufacturing

  11. Water pressure control device for control rod drive

    International Nuclear Information System (INIS)

    Sato, Hideyuki.

    1981-01-01

    Purpose: To minimize the fluctuations in the reactor water level upon occurrence of abnormality by inputting the level signal of the reactor to an arithmetic unit for controlling the pressure of control rod drive water to thereby enable effective reactor level control. Constitution: Signal from a flow rate transmitter is inputted into an arithmetic unit to perform constant flow rate control upon normal operation. While on the other hand, if abnormality occurs such as feedwater pump trips, the arithmetic unit is switched from the constant flow rate control to the reactor water level control. Reactor water level signal is inputted into the arithmetic unit and the control valve is most suitably controlled, whereby water is fed from CST to the reactor by way of control rod drive water system to secure the reactor water level if feedwater to the reactor is interrupted by loss of coolants on the feedwater system. Since this enables to minimize the fluctuations in the reactor water level upon abnormality, the reactor water level can be controlled most suitably by the reactor water level signal. (Moriyama, K.)

  12. Photocatalysis in Generation of Hydrogen from Water

    KAUST Repository

    Takanabe, Kazuhiro; Domen, Kazunari

    2015-01-01

    of a reaction (e.g., water splitting or CO2 reduction) that occurs at extremely high temperatures. The photocatalytic system contains powder photocatalysts. Each photocatalyst particle should collect sufficient photons from the solar flux to cause

  13. Station black out concurrent with PORV failure using a Generic Pressurized Water Reactor simulator

    International Nuclear Information System (INIS)

    Zubair, Muhammad; Ababneh, Ahmad; Ishag, Ahmed

    2017-01-01

    Highlights: •SBO accident simulation by using a GPWR simulator. •Normal SBO, and SBO with additional failure of Pilot Operated Relief Valve. •The research results will provide help in future for better understanding of accidents in APR 1400 reactors. -- Abstract: Station Black Out (SBO) is an accident situation that refers to the total loss of offsite power, along with the unavailability of onsite power, which results from the failure of all Diesel Generators (DG). Probabilistic Safety Assessment (PSA) spans a number of methods that include modeling of event-trees and simulation of accidents scenarios, aimed to quantify risk and ensure safety in nuclear power plants. PSA also deals with prediction of future accidents and calculation of failure probabilities that has been done in this study. A SBO accident was simulated using a Generic Pressurized Water Reactor (GPWR) simulator from KEYMASTER™. The accident scenario consists of two stages; the first stage belongs to normal SBO, in second stage SBO accident with additional failure of Pilot Operated Relief Valve (PORV) opens and it stuck open has been considered for the pressurizer. A comparison of the two stages was made by plotting variables on the same graph. The research has been carried out to analyze the hot and cold leg temperatures, Steam Generator (SG) pressure, SG Narrow Range (NR) level, SG water-level-percentage (PCT), Pressurizer pressure, Fuel Temperature, and containment pressure. Simulation results suggest that failure in closing PORV has negligible impact on hot and cold leg temperatures, results in an overall less pressure in SGs, but higher pressure in the pressurizer. Additionally, containment pressure did not exceed the maximum approved pressure of 8.7 kg/cm 2 , but was approaching the Advanced Pressurized Water Reactor’s (APR-1400) design pressure of 4.218 kg/cm 2 . Finally, nuclear fuel temperature exceeded Probabilistic Risk Assessment (PRA) limit of 726.7 °C for both scenarios. The

  14. On the spurious pressures generated by certain GFEM solutions of the incompressible Navier-Stokes equations

    International Nuclear Information System (INIS)

    Sani, R.L.; Gresho, P.M.; Lee, R.L.

    1979-01-01

    The spurious pressures and acceptable velocities generated when using certain combinations of velocity and pressure approximations in a Galerkin finite element discretization of the primitive variable form of the incompressible Navier-Stokes equations are analyzed both theoretically and numerically for grids composed of quadrilateral finite elements. Schemes for obtaining usable pressure fields from the spurious numerical results are presented for certain cases

  15. Pressurized air cathodes for enhanced stability and power generation by microbial fuel cells

    Science.gov (United States)

    He, Weihua; Yang, Wulin; Tian, Yushi; Zhu, Xiuping; Liu, Jia; Feng, Yujie; Logan, Bruce E.

    2016-11-01

    Large differences between the water and air pressure in microbial fuel cells (MFCs) can deform and damage cathodes. To avoid deformation, the cathode air pressure was controlled to balance pressure differences between the air and water. Raising the air pressures from 0 to 10 kPa at a set cathode potential of -0.3 V (versus Ag/AgCl) enhanced cathode performance by 17%, but pressures ≥25 kPa decreased current and resulted in air leakage into the solution. Matching the air pressure with the water pressure avoided cathode deformation and improved performance. The maximum power density increased by 15%, from 1070 ± 20 to 1230 ± 70 mW m-2, with balanced air and water pressures of 10-25 kPa. Oxygen partial pressures ≥12.5 kPa in the cathode compartment maintained the oxygen reduction rate to be within 92 ± 1% of that in ambient air. The use of pressurized air flow through the cathode compartments can enable closer spacing of the cathodes compared to passive gas transfer systems, which could make the reactor design more compact. The energy cost of pressurizing the cathodes was estimated to be smaller than the increase in power that resulted from the use of pressurized cathodes.

  16. Pressurized air cathodes for enhanced stability and power generation by microbial fuel cells

    KAUST Repository

    He, Weihua

    2016-09-30

    Large differences between the water and air pressure in microbial fuel cells (MFCs) can deform and damage cathodes. To avoid deformation, the cathode air pressure was controlled to balance pressure differences between the air and water. Raising the air pressures from 0 to 10 kPa at a set cathode potential of −0.3 V (versus Ag/AgCl) enhanced cathode performance by 17%, but pressures ≥25 kPa decreased current and resulted in air leakage into the solution. Matching the air pressure with the water pressure avoided cathode deformation and improved performance. The maximum power density increased by 15%, from 1070 ± 20 to 1230 ± 70 mW m, with balanced air and water pressures of 10–25 kPa. Oxygen partial pressures ≥12.5 kPa in the cathode compartment maintained the oxygen reduction rate to be within 92 ± 1% of that in ambient air. The use of pressurized air flow through the cathode compartments can enable closer spacing of the cathodes compared to passive gas transfer systems, which could make the reactor design more compact. The energy cost of pressurizing the cathodes was estimated to be smaller than the increase in power that resulted from the use of pressurized cathodes.

  17. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  18. Orange oil/water nanoemulsions prepared by high pressure homogenizer

    International Nuclear Information System (INIS)

    Kourniatis, Loretta R.; Spinelli, Luciana S.; Mansur, Claudia R.E.

    2010-01-01

    The objective of this work was to use the high-pressure homogenizer (HPH) to prepare stable oil/water nanoemulsions presenting narrow particle size distribution. The dispersions were prepared using nonionic surfactants based on ethoxylated ether. The size and distribution of the droplets formed, along with their stability, were determined in a Zetasizer Nano ZS particle size analyzer. The stability and the droplet size distribution in these systems do not present the significant differences with the increase of the processing pressure in the HPH). The processing time can promote the biggest dispersion in the size of particles, thus reducing its stability. (author)

  19. Ultra-high pressure water jetting for coating removal and surface preparation

    Science.gov (United States)

    Johnson, Spencer T.

    1995-01-01

    This paper shall examine the basics of water technology with particular attention paid to systems currently in use and some select new applications. By providing an overview of commercially available water jet systems in the context of recent case histories, potential users may evaluate the process for future applications. With the on going introduction of regulations prohibiting the use of chemical paint strippers, manual scrapping and dry abrasive media blasting, the need for an environmentally compliant coating removal process has been mandated. Water jet cleaning has been a traditional part of many industrial processed for year, although it has only been in the last few years that reliable pumping equipment capable of ultra-high pressure operation have become available. With the advent of water jet pumping equipment capable of sustaining pressures in excess of 36,000 psi. there has been shift away from lower pressure, high water volume systems. One of the major factors in driving industry to seek higher pressures is the ability to offer higher productivity rates while lowering the quantity of water used and subsequently reprocessed. Among benefits of the trend toward higher pressure/lower volume systems is the corresponding reduction in water jet reaction forces making hand held water jetting practical and safe. Other unique applications made possible by these new generation pumping systems include the use of alternative fluids including liquid ammonia for specialized and hazardous material removal applications. A review of the equipment used and the required modifications will be presented along with the conclusions reached reached during this test program.

  20. Drought Resilience of Water Supplies for Shale Gas Extraction and Related Power Generation in Texas

    Science.gov (United States)

    Reedy, R. C.; Scanlon, B. R.; Nicot, J. P.; Uhlman, K.

    2014-12-01

    There is considerable concern about water availability to support energy production in Texas, particularly considering that many of the shale plays are in semiarid areas of Texas and the state experienced the most extreme drought on record in 2011. The Eagle Ford shale play provides an excellent case study. Hydraulic fracturing water use for shale gas extraction in the play totaled ~ 12 billion gallons (bgal) in 2012, representing ~7 - 10% of total water use in the 16 county play area. The dominant source of water is groundwater which is not highly vulnerable to drought from a recharge perspective because water is primarily stored in the confined portion of aquifers that were recharged thousands of years ago. Water supply drought vulnerability results primarily from increased water use for irrigation. Irrigation water use in the Eagle Ford play was 30 billion gallons higher in the 2011 drought year relative to 2010. Recent trends toward increased use of brackish groundwater for shale gas extraction in the Eagle Ford also reduce pressure on fresh water resources. Evaluating the impacts of natural gas development on water resources should consider the use of natural gas in power generation, which now represents 50% of power generation in Texas. Water consumed in extracting the natural gas required for power generation is equivalent to ~7% of the water consumed in cooling these power plants in the state. However, natural gas production from shale plays can be overall beneficial in terms of water resources in the state because natural gas combined cycle power generation decreases water consumption by ~60% relative to traditional coal, nuclear, and natural gas plants that use steam turbine generation. This reduced water consumption enhances drought resilience of power generation in the state. In addition, natural gas combined cycle plants provide peaking capacity that complements increasing renewable wind generation which has no cooling water requirement. However, water

  1. Development of design technology for an advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Kim, Dong Soo; Chang, Won Pyo; Park, Koon Chul

    1991-07-01

    The objective of the project is to localize technology for the improvement of the reactor coolant system through a multidimensional thermal-hydraulic analysis for the steam generator and the pressurizer. Flow distribution analysis has been done for the YGN 3/4 steam generators when steady-state output conditions were varied in the ranges such as 100, 75, 50, and 25 using three-dimensional ATHOS 3 code. The results of the thermal-hydraulic analysis have been used for flow-induced vibration analysis for the YGN 3/4 steam generators. ATHOS 3 code has been modified for YGN 3/4 steam generator tube lane region using the cartesian geometry and the local porosity in the boundaries of the two adjacent cells. Stability ratio for the tube vibration has been calculated the modified ATHOS 3 and ANSYS code. A sensitivity study for the pressurizer volume change has been analyzed using LTC code which is for the performance analysis to predict an optimistic pressurizer volume. (Author)

  2. Fresh water generators onboard a floating platform

    International Nuclear Information System (INIS)

    Tewari, P.K.; Verma, R.K.; Misra, B.M.; Sadhulkan, H.K.

    1997-01-01

    A dependable supply of fresh water is essential for any ocean going vessel. The operating and maintenance personnel on offshore platforms and marine structures also require a constant and regular supply of fresh water to meet their essential daily needs. A seawater thermal desalination unit onboard delivers good quality fresh water from seawater. The desalination units developed by Bhabha Atomic Research Centre (BARC) suitable for ocean going vessels and offshore platforms have been discussed. Design considerations of such units with reference to floating platforms and corrosive environments have been presented. The feasibility of coupling a low temperature vacuum evaporation (LTVE) desalination plant suitable for an onboard floating platform to a PHWR nuclear power plant has also been discussed. (author). 1 ref., 3 figs, 2 tabs

  3. Swelling pressure and water absorption property of compacted granular bentonite during water absorption

    International Nuclear Information System (INIS)

    Oyamada, T.; Komine, H.; Murakami, S.; Sekiguchi, T.; Sekine, I.

    2012-01-01

    Document available in extended abstract form only. Bentonite is currently planned to be used as buffer materials in engineered barrier of radioactive waste disposal. Granular bentonites are expected as the materials used in constructions as buffer materials by in-situ compaction methods. After applying these buffer materials, it is expected that the condition of the buffer area changes in long-term by the seepage of groundwater into buffer area. Therefore, it is important to understand water movement and swelling behavior of the buffer materials for evaluating the performance of engineered barrier. In this study, we investigated water absorption property and swelling pressure of compacted granular bentonite. Specifically, the process of swelling pressure and amount of water absorption of granular bentonite-GX (Kunigel-GX, produced at the Tsukinuno mine in Japan) were observed by laboratory tests. To discuss the influence of maximum grain size of bentonite particle on swelling pressure and water absorption property, two types of samples were used. One is granular sample which is Bentonite-GX controlled under 2 mm the maximum grain size, the other is milled sample which is Bentonite-GX with the maximum grain size under 0.18 mm by milling with the agate mortar. In addition, the mechanism on the swelling pressure of compacted granular bentonite was considered and discussed. In the cases of granular sample, swelling pressure increases rapidly, then gradually continues to increase up to maximum value. In the cases of milled sample, swelling pressure also increases rapidly at first. However, then its value decreases before progressing of gradual increase continues. Especially, this trend was clearly observed at a relatively low dry density. At the peaks of these curves, the swelling pressure of granular samples is lower than that of milled samples. In addition, the increasing of swelling pressure by the time the peak observed during the process of swelling pressure from

  4. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  5. Press kit. EPR (European pressurized water reactor). The advanced nuclear reactor

    International Nuclear Information System (INIS)

    2004-10-01

    Nuclear energy, which provides a steady supply of electricity at low cost, has its rightful place in the energy mix of the 21 century, which puts the emphasis on sustainable development. In this framework, this document presents the advantages of the EPR (European Pressurized water Reactor). The EPR is the only third generation reactor under construction today. It is an evolutionary reactor that represents a new generation of pressurized water reactors with no break in the technology used for the most recent models. The EPR can guarantee a safe, inexpensive electricity supply, without adding to the greenhouse effect. It meets the requirements of the safety authorities and lives up to the expectations of electricity utilities. (A.L.B.)

  6. CYANOTOXINS: NEW GENERATION OF WATER CONTAMINANTS

    Science.gov (United States)

    Cyanobacteria, more commonly known as blue-green algae, are found worldwide in various aquatic environments as well as in water distribution systems (Atikovic 2003; Carmichael 1994; Madigan et al. 2003). Blooms of cyanobacteria have recently become spatially and temporally more p...

  7. The effect of pressurizer-water-level on the low frequency component of the pressure spectrum in a PWR

    International Nuclear Information System (INIS)

    Por, G.; Izsak, E.; Valko, J.

    1984-09-01

    The pressure fluctuations were measured in the cooling system of the Paks-1 reactor. A shift of the peak was detected in low frequency component of the pressure fluctuation spectrum which is due to the fluctuations of water level in the pressurizer. Using the model of Katona and Nagy (1983), the eigenfrequencies, the magnitude of the shift and the sensitivity to the pressurizer water level were reproduced in good accordance with the experimental data. (D.Gy.)

  8. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  9. Characterization of Pressure Transients Generated by Nanosecond Electrical Pulse (nsEP) Exposure

    Science.gov (United States)

    Roth, Caleb C.; Barnes Jr., Ronald A.; Ibey, Bennett L.; Beier, Hope T.; Christopher Mimun, L.; Maswadi, Saher M.; Shadaram, Mehdi; Glickman, Randolph D.

    2015-01-01

    The mechanism(s) responsible for the breakdown (nanoporation) of cell plasma membranes after nanosecond pulse (nsEP) exposure remains poorly understood. Current theories focus exclusively on the electrical field, citing electrostriction, water dipole alignment and/or electrodeformation as the primary mechanisms for pore formation. However, the delivery of a high-voltage nsEP to cells by tungsten electrodes creates a multitude of biophysical phenomena, including electrohydraulic cavitation, electrochemical interactions, thermoelastic expansion, and others. To date, very limited research has investigated non-electric phenomena occurring during nsEP exposures and their potential effect on cell nanoporation. Of primary interest is the production of acoustic shock waves during nsEP exposure, as it is known that acoustic shock waves can cause membrane poration (sonoporation). Based on these observations, our group characterized the acoustic pressure transients generated by nsEP and determined if such transients played any role in nanoporation. In this paper, we show that nsEP exposures, equivalent to those used in cellular studies, are capable of generating high-frequency (2.5 MHz), high-intensity (>13 kPa) pressure transients. Using confocal microscopy to measure cell uptake of YO-PRO®-1 (indicator of nanoporation of the plasma membrane) and changing the electrode geometry, we determined that acoustic waves alone are not responsible for poration of the membrane. PMID:26450165

  10. Characterization of Pressure Transients Generated by Nanosecond Electrical Pulse (nsEP) Exposure.

    Science.gov (United States)

    Roth, Caleb C; Barnes, Ronald A; Ibey, Bennett L; Beier, Hope T; Christopher Mimun, L; Maswadi, Saher M; Shadaram, Mehdi; Glickman, Randolph D

    2015-10-09

    The mechanism(s) responsible for the breakdown (nanoporation) of cell plasma membranes after nanosecond pulse (nsEP) exposure remains poorly understood. Current theories focus exclusively on the electrical field, citing electrostriction, water dipole alignment and/or electrodeformation as the primary mechanisms for pore formation. However, the delivery of a high-voltage nsEP to cells by tungsten electrodes creates a multitude of biophysical phenomena, including electrohydraulic cavitation, electrochemical interactions, thermoelastic expansion, and others. To date, very limited research has investigated non-electric phenomena occurring during nsEP exposures and their potential effect on cell nanoporation. Of primary interest is the production of acoustic shock waves during nsEP exposure, as it is known that acoustic shock waves can cause membrane poration (sonoporation). Based on these observations, our group characterized the acoustic pressure transients generated by nsEP and determined if such transients played any role in nanoporation. In this paper, we show that nsEP exposures, equivalent to those used in cellular studies, are capable of generating high-frequency (2.5 MHz), high-intensity (>13 kPa) pressure transients. Using confocal microscopy to measure cell uptake of YO-PRO®-1 (indicator of nanoporation of the plasma membrane) and changing the electrode geometry, we determined that acoustic waves alone are not responsible for poration of the membrane.

  11. RELAP5 - a new tool for pressurized water reactor safety analysis

    International Nuclear Information System (INIS)

    Perneczky, L.

    1988-11-01

    The RELAP type pressurized water reactor safety system codes are used world wide for the loss of coolant accident analyses. In this paper the RELAP5, the advanced generation of the code family is presented. The relationship to RELAP4/mod6 version is discussed. The capability of the RELAP5/mod1-EUR version for small, medium and large break LOCA is investigated based on international user experience. (author) 30 refs

  12. Loss-of-Fluid Test findings in pressurized water reactor core's thermal-hydraulic behavior

    International Nuclear Information System (INIS)

    Russell, M.

    1983-01-01

    This paper summarizes the pressurized water reactor (PWR) core's thermal-hydraulic behavior findings from experiments performed at the Loss-of-Fluid Test (LOFT) Facility at the Idaho National Engineering Laboratory. The potential impact of these findings on the safety and economics of PWR's generation of electricity is also discussed. Reviews of eight important findings in the core's physical behavior and in experimental methods are presented with supporting evidence

  13. The study and improvement of water level control of pressurizer

    International Nuclear Information System (INIS)

    Gao Peng; Zhang Qinshun

    2006-01-01

    The PI controller which is used widely in water level control of pressurizer in reactor control system usually leads dynamic overshoot and long setting time. The improvement project for intelligent fuzzy controller to take the place of PI controller is advanced. This paper researches the water level control of pressurizer in reactor control system of Daya Bay Phase I, and describes the method of intelligent fuzzy control in practice. Simulation indicates that the fuzzy control has advantages of small overshoot and short settling time. It can also improve control system's real time property and anti-interference ability. Especially for non-linear and time-varying complicated control systems, it can obtain good control results. (authors)

  14. EDF's nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1987-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction-had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's 'with book' on nuclear safety. (author)

  15. Development of alternative fuel for pressurized water reactors

    International Nuclear Information System (INIS)

    Cardoso, P.E.; Ferreira, R.A.N.; Ferraz, W.B.; Lameiras, F.S.; Santos, A.; Assis, G. de; Doerr, W.O.; Wehner, E.L.

    1984-01-01

    The utilization of alternative fuel cycles in Pressurized Water Reactors (PWR) such as Th/U and Th/Pu cycles can permit a better utilization of uranium reserves without the necessity of developing new power reactor concepts. The development of the technology of alternative fuels for PWR is one of the objectives of the 'Program on Thorium Utilization in Pressurized Water Reactors' carried out jointly by Empresas Nucleares Brasileiras S.A. (NUCLEBRAS), through its Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) and by German institutions, the Julich Nuclear Research Center (KFA), the Kraftwerk Union A.G. (KWU) and NUKEM GmbH. This paper summarizes the results so far obtained in the fuel technology. The development of a fabrication process for PWR fuel pellets from gel-microspheres is reported as well as the design, the specification, and the fabrication of prototype fuel rods for irradiation tests. (Author) [pt

  16. Sources of radioiodine at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Pelletier, C.A.; Cline, J.E.; Barefoot, E.D.; Hemphill, R.T.; Voilleque, P.G.; Emel, W.A.

    1978-11-01

    The report determines specific components and operations at operating pressurized water reactors that have a potential for being significant emission sources of radioactive iodine. The relative magnitudes of these specific sources in terms of the chemical forms of the radioiodine and the resultant annual averages from major components are established. The data are generalized for broad industry use for predictive purposes. The conclusions of this study indicate that the majority of radioiodine emanating from the primary side of pressurized water reactors comes from a few major areas; in some cases these sources are locally treatable; the interaction of radioiodine with plant interior surfaces is an important phenomenon mediating the source and affecting its release to the atmosphere; the chemical form varies depending on the circumstances of the release

  17. EDF'S nuclear safety approach for pressurized water reactors

    International Nuclear Information System (INIS)

    Tanguy, P.; Kus, J.P.

    1988-01-01

    The realization of the important French program fifty-four units equipped with pressurized water reactors in service, or under construction - had led to the progressive implementation of an original approach in the field of nuclear safety. From an initial core consisting of the deterministic approach to safety devised on the other side of the Atlantic, which has been entirely preserved and often specified, further extras have been added which overall increase the level of safety of the installations, without any particular complications. This paper aims at presenting succinctly the outcome of the deliberation, which constitutes now the approach adopted by Electricite de France for the safety of nuclear units equipped with pressurized water reactors. This approach is explained in more detail in EDF's white book on nuclear safety

  18. Proposal of electric power generation from generators to water edge in the region of Sarapiqui

    International Nuclear Information System (INIS)

    Rodriguez Fallas, Cindy Veronica

    2013-01-01

    A proposed electric power generation is developed from generators to water edge in the region of Sarapiqui. The environmental characteristics, such as the hydrological network, hydrogeology, soil type, life zones, climatology, precipitation, temperature, evapotranspiration and water supply and demand, of rivers crossed by basin in the region of Sarapiqui, are determined by bibliographic consultations to implement the proposal. The most recent production statistics of the electric subsector of Costa Rica are described to reveal the growing annual demand and need for satisfaction. The zone of Sarapiqui is diagnosed as the right place to allow the generation of electric power from generators to water edge [es

  19. Water-vapor pressure control in a volume

    Science.gov (United States)

    Scialdone, J. J.

    1978-01-01

    The variation with time of the partial pressure of water in a volume that has openings to the outside environment and includes vapor sources was evaluated as a function of the purging flow and its vapor content. Experimental tests to estimate the diffusion of ambient humidity through openings and to validate calculated results were included. The purging flows required to produce and maintain a certain humidity in shipping containers, storage rooms, and clean rooms can be estimated with the relationship developed here. These purging flows are necessary to prevent the contamination, degradation, and other effects of water vapor on the systems inside these volumes.

  20. Pressure-induced transformations in computer simulations of glassy water

    Science.gov (United States)

    Chiu, Janet; Starr, Francis W.; Giovambattista, Nicolas

    2013-11-01

    Glassy water occurs in at least two broad categories: low-density amorphous (LDA) and high-density amorphous (HDA) solid water. We perform out-of-equilibrium molecular dynamics simulations to study the transformations of glassy water using the ST2 model. Specifically, we study the known (i) compression-induced LDA-to-HDA, (ii) decompression-induced HDA-to-LDA, and (iii) compression-induced hexagonal ice-to-HDA transformations. We study each transformation for a broad range of compression/decompression temperatures, enabling us to construct a "P-T phase diagram" for glassy water. The resulting phase diagram shows the same qualitative features reported from experiments. While many simulations have probed the liquid-state phase behavior, comparatively little work has examined the transitions of glassy water. We examine how the glass transformations relate to the (first-order) liquid-liquid phase transition previously reported for this model. Specifically, our results support the hypothesis that the liquid-liquid spinodal lines, between a low-density and high-density liquid, are extensions of the LDA-HDA transformation lines in the limit of slow compression. Extending decompression runs to negative pressures, we locate the sublimation lines for both LDA and hyperquenched glassy water (HGW), and find that HGW is relatively more stable to the vapor. Additionally, we observe spontaneous crystallization of HDA at high pressure to ice VII. Experiments have also seen crystallization of HDA, but to ice XII. Finally, we contrast the structure of LDA and HDA for the ST2 model with experiments. We find that while the radial distribution functions (RDFs) of LDA are similar to those observed in experiments, considerable differences exist between the HDA RDFs of ST2 water and experiment. The differences in HDA structure, as well as the formation of ice VII (a tetrahedral crystal), are a consequence of ST2 overemphasizing the tetrahedral character of water.

  1. High Pressure Oxygen Generation for Future Exploration Missions, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — The proposed innovation is the development of a cathode feed electrolysis cell stack capable of generating 3600 psi oxygen at a relevant scale for future exploration...

  2. Clinch river breeder reactor plant steam generator water quality

    International Nuclear Information System (INIS)

    Van Hoesen, D.; Lowe, P.A.

    1975-01-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

  3. Clinch river breeder reactor plant steam generator water quality

    Energy Technology Data Exchange (ETDEWEB)

    Van Hoesen, D; Lowe, P A

    1975-07-01

    The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

  4. Assessment of a small pressurized water reactor for industrial energy

    International Nuclear Information System (INIS)

    Klepper, O.H.; Fuller, L.C.; Myers, M.L.

    1977-01-01

    An evaluation of several recent ERDA/ORNL sponsored studies on the application of a small, 365 MW(t) pressurized water reactor for industrial energy is presented. Preliminary studies have investigated technical and reliability requirements; costs for nuclear and fossil based steam were compared, including consideration of economic inflation and financing methods. For base-load industrial steam production, small reactors appear economically attractive relative to coal fired boilers that use coal priced at $30/ton

  5. Solubility of solid ferrocene in pressurized hot water

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Hohnová, Barbora; Planeta, Josef; Roth, Michal

    2010-01-01

    Roč. 55, č. 8 (2010), s. 2866-2869 ISSN 0021-9568 R&D Projects: GA ČR GA203/07/0886; GA ČR GA203/08/1465; GA ČR GA203/08/1536 Institutional research plan: CEZ:AV0Z40310501 Keywords : pressurized hot water * ferrocene * solubility Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 2.089, year: 2010

  6. An expert system for pressurized water reactor load maneuvers

    International Nuclear Information System (INIS)

    Chaung Lin; Jungping Chen; Yihjiunn Lin; Lianshin Lin

    1993-01-01

    Restartup after reactor shutdown and load-follow operations are the important tasks in operating pressurized water reactors. Generally, the most efficient method is to apply constant axial offset control (CAOC) strategy during load maneuvers. An expert system using CAOC strategy, fuzzy reasoning, a two-node core model, and operational constraints has been developed. The computation time is so short that this system, which leads to an approximate closed-loop control, could be useful for on-site calculation

  7. Solubilities of oxygenated aromatic solids in pressurized hot water

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Planeta, Josef; Roth, Michal

    2009-01-01

    Roč. 54, č. 5 (2009), s. 1457-1461 ISSN 0021-9568 R&D Projects: GA ČR GA203/07/0886; GA ČR GA203/08/1536 Institutional research plan: CEZ:AV0Z40310501 Keywords : oxygenated aromatics * solubility * pressurized hot water Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 1.695, year: 2009

  8. Decommissioning of the BR3 pressurized-water reactor

    International Nuclear Information System (INIS)

    Massaut, V.

    1996-01-01

    The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

  9. Water pressure and ground vibrations induced by water guns near Bandon Road Lock and Dam and Lemont, Illinois

    Science.gov (United States)

    Adams, Ryan F.; Koebel, Carolyn M.; Morrow, William S.

    2018-02-13

    surface. A combination of two 80-in3 water guns set at 6 and 8 ft, respectively, produced a similarly shaped 5 lb/in2 pressure field 30 ft in radius. Neither of the water gun configurations exceeded the given threshold of 5 lb/in2 above the static pressure along the walls of the canal at the 700 lb/in2 water gun input pressure. Velocity and acceleration data were collected simultaneously with the underwater pressure data to understand the response of adjacent canal walls to the water gun firings. Maximum velocity and acceleration were 0.239 in/s and 0.0188 feet per second squared (ft/s2), respectively.The Lemont study replicated and expanded upon work done in 2011. The pressure field created by the water gun was evaluated in a deeper environment (about 25 ft of water depth) than that of the Brandon Road study. To replicate the 2011 study, data were collected with the same water gun placements and input pressure, but static underwater pressure monitoring was added. Two 80-in3 water guns were suspended below a platform at depths of 4 and 14 ft bws. Pressure was lower when the gun suspended at 4 ft bws was fired as compared to firing the single gun suspended at 14 ft bws. Firing both guns simultaneously produced similar pressures to the single gun suspended at 14 ft bws. Data were collected to assess the pressure field produced by two 80-in3 water guns separated by 80 ft and suspended at a depth of 14 ft bws. The spatial extent of the 5-lb/in2 threshold varied substantially with gun input air pressure. Firing the water gun with an air pressure of 2,000 lb/in2 generated a pressure field greater than the threshold at all but one location in the measured region. Additionally, the water gun with an air pressure of 1,000 lb/in2 did not reach the threshold anywhere in the measured region. Maximum velocity and acceleration were 0.304 in/s and 0.015 ft/s2, respectively.

  10. The effects of pulse pressure from seismic water gun technology on Northern Pike

    Science.gov (United States)

    Gross, Jackson A.; Irvine, Kathryn M.; Wilmoth, Siri K.; Wagner, Tristany L.; Shields, Patrick A; Fox, Jeffrey R.

    2013-01-01

    We examined the efficacy of sound pressure pulses generated from a water gun for controlling invasive Northern Pike Esox lucius. Pulse pressures from two sizes of water guns were evaluated for their effects on individual fish placed at a predetermined random distance. Fish mortality from a 5,620.8-cm3 water gun (peak pressure source level = 252 dB referenced to 1 μP at 1 m) was assessed every 24 h for 168 h, and damage (intact, hematoma, or rupture) to the gas bladder, kidney, and liver was recorded. The experiment was replicated with a 1,966.4-cm3 water gun (peak pressure source level = 244 dB referenced to 1 μP at 1 m), but fish were euthanized immediately. The peak sound pressure level (SPLpeak), peak-to-peak sound pressure level (SPLp-p), and frequency spectrums were recorded, and the cumulative sound exposure level (SELcum) was subsequently calculated. The SPLpeak, SPLp-p, and SELcum were correlated, and values varied significantly by treatment group for both guns. Mortality increased and organ damage was greater with decreasing distance to the water gun. Mortality (31%) by 168 h was only observed for Northern Pike exhibiting the highest degree of organ damage. Mortality at 72 h and 168 h postexposure was associated with increasing SELcum above 195 dB. The minimum SELcum calculated for gas bladder rupture was 199 dB recorded at 9 m from the 5,620.8-cm3 water gun and 194 dB recorded at 6 m from the 1,966.4-cm3water gun. Among Northern Pike that were exposed to the large water gun, 100% of fish exposed at 3 and 6 m had ruptured gas bladders, and 86% exposed at 9 m had ruptured gas bladders. Among fish that were exposed to pulse pressures from the smaller water gun, 78% exhibited gas bladder rupture. Results from these initial controlled experiments underscore the potential of water guns as a tool for controlling Northern Pike.

  11. Drinking water fluoride and blood pressure? An environmental study.

    Science.gov (United States)

    Amini, Hassan; Taghavi Shahri, Seyed Mahmood; Amini, Mohamad; Ramezani Mehrian, Majid; Mokhayeri, Yaser; Yunesian, Masud

    2011-12-01

    The relationship between intakes of fluoride (F) from drinking water and blood pressure has not yet been reported. We examined the relationship of F in ground water resources (GWRs) of Iran with the blood pressure of Iranian population in an ecologic study. The mean F data of the GWRs (as a surrogate for F levels in drinking water) were derived from a previously conducted study. The hypertension prevalence and the mean of systolic and diastolic blood pressures (SBP & DBP) of Iranian population by different provinces and genders were also derived from the provincial report of non-communicable disease risk factor surveillance of Iran. Statistically significant positive correlations were found between the mean concentrations of F in the GWRs and the hypertension prevalence of males (r = 0.48, p = 0.007), females (r = 0.36, p = 0.048), and overall (r = 0.495, p = 0.005). Also, statistically significant positive correlations between the mean concentrations of F in the GWRs and the mean SBP of males (r = 0.431, p = 0.018), and a borderline correlation with females (r = 0.352, p = 0.057) were found. In conclusion, we found the increase of hypertension prevalence and the SBP mean with the increase of F level in the GWRs of Iranian population.

  12. Reactor Safety Commission Code of Practice for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    1990-01-01

    The Reactor Safety Commission of the Federal German Republic has summarized in the form of Official Guidelines the safety requirements which, in the Commission's view, have to be met in the design, construction and operation of a nuclear power station equipped with a pressurized water reactor. The Third Edition of the RSK Guidelines for pressurized water reactors dated 14.10.81. is a revised and expanded version of the Second Edition dated 24.1.79. The Reactor Safety Commission will with effect from October 1981 use these Guidelines in consultations on the siting of and safety concept for the installation approval of future pressurized water reactors and will assess these nuclear power stations during their erection in the light of these Guidelines. They have not however been immediately conceived for the adaptation of existing nuclear power stations, whether under construction or in operation. The scope of application of these Guidelines to such nuclear power stations will have to be examined for each individual case. The main aim of the Guidelines is to simplify the consultation process within the reactor Safety Commission and to provide early advice on the safety requirements considered necessary by the Commission. (author)

  13. Pressure Transient Model of Water-Hydraulic Pipelines with Cavitation

    Directory of Open Access Journals (Sweden)

    Dan Jiang

    2018-03-01

    Full Text Available Transient pressure investigation of water-hydraulic pipelines is a challenge in the fluid transmission field, since the flow continuity equation and momentum equation are partial differential, and the vaporous cavitation has high dynamics; the frictional force caused by fluid viscosity is especially uncertain. In this study, due to the different transient pressure dynamics in upstream and downstream pipelines, the finite difference method (FDM is adopted to handle pressure transients with and without cavitation, as well as steady friction and frequency-dependent unsteady friction. Different from the traditional method of characteristics (MOC, the FDM is advantageous in terms of the simple and convenient computation. Furthermore, the mechanism of cavitation growth and collapse are captured both upstream and downstream of the water-hydraulic pipeline, i.e., the cavitation start time, the end time, the duration, the maximum volume, and the corresponding time points. By referring to the experimental results of two previous works, the comparative simulation results of two computation methods are verified in experimental water-hydraulic pipelines, which indicates that the finite difference method shows better data consistency than the MOC.

  14. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    International Nuclear Information System (INIS)

    Park, Soo Yong; Ahn, Kwang Il

    2015-01-01

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO

  15. Evaluation of an accident management strategy of emergency water injection using fire engines in a typical pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Following the Fukushima accident, a special safety inspection was conducted in Korea. The inspection results show that Korean nuclear power plants have no imminent risk for expected maximum potential earthquake or coastal flooding. However long- and short-term safety improvements do need to be implemented. One of the measures to increase the mitigation capability during a prolonged station blackout (SBO) accident is installing injection flow paths to provide emergency cooling water of external sources using fire engines to the steam generators or reactor cooling systems. This paper illustrates an evaluation of the effectiveness of external cooling water injection strategies using fire trucks during a potential extended SBO accident in a 1,000 MWe pressurized water reactor. With regard to the effectiveness of external cooling water injection strategies using fire engines, the strategies are judged to be very feasible for a long-term SBO, but are not likely to be effective for a short-term SBO.

  16. Design study on steam generator integration into the VVER reactor pressure vessel

    International Nuclear Information System (INIS)

    Hort, J.; Matal, O.

    2004-01-01

    The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

  17. Molecular cobalt pentapyridine catalysts for generating hydrogen from water

    Science.gov (United States)

    Long, Jeffrey R; Chang, Christopher J; Sun, Yujie

    2013-11-05

    A composition of matter suitable for the generation of hydrogen from water is described, the positively charged cation of the composition including the moiety of the general formula. [(PY5Me.sub.2)CoL].sup.2+, where L can be H.sub.2O, OH.sup.-, a halide, alcohol, ether, amine, and the like. In embodiments of the invention, water, such as tap water or sea water can be subject to low electric potentials, with the result being, among other things, the generation of hydrogen.

  18. Possibilities of tritium removal from waste waters of pressurized water reactors and fuel reprocessing plants

    International Nuclear Information System (INIS)

    Ribnikar, S.V.; Pupezin, J.D.

    1975-01-01

    Starting from parameters known for heavy water production processes, a parallel was made with separation of tritium from water. The quantity in common is the total cascade flow. The most efficient processes appear to be hydrogen sulfide, water exchange, hydrogen- and water distillation. Prospects of application of new processes are discussed briefly. Problems concerning detritiation of pressurized water reactors and large fuel reprocessing plants are analyzed. Detritiation of the former should not present problems. With the latter, economical detritiation can be achieved only after some plant flow patterns are changed. (U.S.)

  19. Pressure atomizer having multiple orifices and turbulent generation feature

    Science.gov (United States)

    VanBrocklin, Paul G.; Geiger, Gail E.; Moran, Donald James; Fournier, Stephane

    2002-01-01

    A pressure atomizer includes a silicon plate having a top surface and a bottom surface. A portion of the top surface defines a turbulent chamber. The turbulent chamber is peripherally bounded by the top surface of the plate. The turbulent chamber is recessed a predetermined depth relative to the top surface. The silicon plate further defines at least one flow orifice. Each flow orifice extends from the bottom surface of the silicon plate to intersect with and open into the turbulent chamber. Each flow orifice is in fluid communication with the turbulent chamber.

  20. Biogas generation in anaerobic wastewater treatment under tetracycline antibiotic pressure

    Science.gov (United States)

    Lu, Meiqing; Niu, Xiaojun; Liu, Wei; Zhang, Jun; Wang, Jie; Yang, Jia; Wang, Wenqi; Yang, Zhiquan

    2016-01-01

    The effect of tetracycline (TC) antibiotic on biogas generation in anaerobic wastewater treatment was studied. A lab-scale Anaerobic Baffled Reactor (ABR) with three compartments was used. The reactor was operated with synthetic wastewater in the absence of TC and in the presence of 250 μg/L TC for 90 days, respectively. The removal rate of TC, volatile fatty acids (VFAs), biogas compositions (hydrogen (H2), methane (CH4), carbon dioxide (CO2)), and total biogas production in each compartment were monitored in the two operational conditions. Results showed that the removal rate of TC was 14.97–67.97% in the reactor. The presence of TC had a large negative effect on CH4 and CO2 generation, but appeared to have a positive effect on H2 production and VFAs accumulation. This response indicated that the methanogenesis process was sensitive to TC presence, but the acidogenesis process was insensitive. This suggested that the presence of TC had less influence on the degradation of organic matter but had a strong influence on biogas generation. Additionally, the decrease of CH4 and CO2 generation and the increase of H2 and VFAs accumulation suggest a promising strategy to help alleviate global warming and improve resource recovery in an environmentally friendly approach. PMID:27341657

  1. Improvement in fuel utilization in pressurized heavy water reactors due to increased heavy water purity

    International Nuclear Information System (INIS)

    Balakrishnan, M.R.

    1991-01-01

    This paper reports that in a pressurized heavy water reactor (PHWR), the reactivity of the reactor and, consequently, the discharge burnup of the fuel depend on the isotopic purity of the heavy water used in the reactor. The optimal purity of heavy water used in PHWRs, in turn, depends on the cost of fabricated uranium fuel and on the incremental cost incurred in improving the heavy water purity. The physics and economics aspects of the desirability of increasing the heavy water purity in PHWRs in India were first examined in 1978. With the cost data available at that time, it was found that improving the heavy water purity from 99.80% to 99.95% was economically attractive. The same problem is reinvestigated with current cost data. Even now, there is sufficient incentive to improve the isotopic purity of heavy water used in PHWRs. Admittedly, the economic advantage that can be derived depends on the cost of the fabricated fuel. Nevertheless, irrespective of the economics, there is also a fairly substantial saving in natural uranium. That the increase in the heavy water purity is to be maintained only in the low-pressure moderator system, and not in the high-pressure coolant system, makes the option of achieving higher fuel burnup with higher heavy water purity feasible

  2. Graph based hydraulic modelling of pressure in water distribution networks

    CSIR Research Space (South Africa)

    Mpiana, LA

    2017-09-01

    Full Text Available given period of time and can be formulated as Where 𝐵 is the amount of incoming water at the node 𝑖 and 𝑡 is the delivery time period of time required. The amount of incoming water required at the node for a period of time is compared... to the water demand 𝐷𝑖 in order to decide on the operation of the pressure reduction valves. 𝐵𝑖𝑗 is the coefficient to evaluation in the required amount 𝐵 over a specific period of time. For a node 𝑖, 𝐵𝑖𝑗 is equal to 0...

  3. Nanoporous Membrane for Medical Grade Water Generation, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — For NASA exploration missions to the Moon and Mars, medical grade water generation is a necessity. Adsorption filter technology has shown some promise, but requires...

  4. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  5. Recent Advances in Osmotic Energy Generation via Pressure-Retarded Osmosis (PRO: A Review

    Directory of Open Access Journals (Sweden)

    Jihye Kim

    2015-10-01

    Full Text Available Global energy consumption has been highly dependent on fossil fuels which cause severe climate change and, therefore, the exploration of new technologies to produce effective renewable energy plays an important role in the world. Pressure-retarded osmosis (PRO is one of the promising candidates to reduce the reliance on fossil fuels by harnessing energy from the salinity gradient between seawater and fresh water. In PRO, water is transported though a semi-permeable membrane from a low-concentrated feed solution to a high-concentrated draw solution. The increased volumetric water flow then runs a hydro-turbine to generate power. PRO technology has rapidly improved in recent years; however, the commercial-scale PRO plant is yet to be developed. In this context, recent developments on the PRO process are reviewed in terms of mathematical models, membrane modules, process designs, numerical works, and fouling and cleaning. In addition, the research requirements to accelerate PRO commercialization are discussed. It is expected that this article can help comprehensively understand the PRO process and thereby provide essential information to activate further research and development.

  6. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  7. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Baek, W. P.; Song, C. H.; Kim, Y. S. and others

    2005-02-15

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform various integral effect tests for design, operation, and safety regulation of pressurized water reactors. During the first phase of this project (1997.8{approx}2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished: a full-height, 1/300-volume-scaled full pressure facility for APR1400, an evolutionary pressurized water reactor that was developed by Korean industry. Main objectives of the present phase (2002.4{approx}2005.2), was to optimize the facility design and to construct the experimental facility. We have performed following researches: 1) Optimization of the basic design of the thermal-hydraulic integral effect test facility for PWRs - ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) - Reduced height design for APR1400 (+ specific design features of KSNP safety injection systems) - Thermal-hydraulic scaling based on three-level scaling methodology by Ishii et al. 2) Construction of the ATLAS facility - Detailed design of the test facility - Manufacturing and procurement of components - Installation of the facility 3) Development of supporting technology for integral effect tests - Development and application of advanced instrumentation technology - Preliminary analysis of test scenarios - Development of experimental procedures - Establishment and implementation of QA system/procedure.

  8. Wave Pressures on Seawave Slot-Cone Generator

    DEFF Research Database (Denmark)

    Vicinanza, Diego; Kofoed, Jens Peter; Frigaard, Peter

    2006-01-01

    This paper presents results on loading action on an innovative caisson breakwater for electricity production. The work reported here contributes to the European Union Sixth Framework programme priority 6.1 (Sustainable Energy System), contract 019831, titled "Full- scale demonstration of robust...... and high efficiency energy converter" (WAVESSG). Information on wave loading acting on Wave Energy Convert (WEC) Seawave Slot-Cone GEnerator (SSG) exposed to extreme wave conditions are reported....

  9. Consumptive Water Use from Electricity Generation in the Southwest under Alternative Climate, Technology, and Policy Futures.

    Science.gov (United States)

    Talati, Shuchi; Zhai, Haibo; Kyle, G Page; Morgan, M Granger; Patel, Pralit; Liu, Lu

    2016-11-15

    This research assesses climate, technological, and policy impacts on consumptive water use from electricity generation in the Southwest over a planning horizon of nearly a century. We employed an integrated modeling framework taking into account feedbacks between climate change, air temperature and humidity, and consequent power plant water requirements. These direct impacts of climate change on water consumption by 2095 differ with technology improvements, cooling systems, and policy constraints, ranging from a 3-7% increase over scenarios that do not incorporate ambient air impacts. Upon additional factors being changed that alter electricity generation, water consumption increases by up to 8% over the reference scenario by 2095. With high penetration of wet recirculating cooling, consumptive water required for low-carbon electricity generation via fossil fuels will likely exacerbate regional water pressure as droughts become more common and population increases. Adaptation strategies to lower water use include the use of advanced cooling technologies and greater dependence on solar and wind. Water consumption may be reduced by 50% in 2095 from the reference, requiring an increase in dry cooling shares to 35-40%. Alternatively, the same reduction could be achieved through photovoltaic and wind power generation constituting 60% of the grid, consistent with an increase of over 250% in technology learning rates.

  10. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    Conley, D.A.; Edson, J.L.; Fineman, C.F.

    1995-03-01

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  11. Can the water content of highly compacted bentonite be increased by applying a high water pressure?

    International Nuclear Information System (INIS)

    Pusch, R.; Kasbohm, J.

    2001-10-01

    A great many laboratory investigations have shown that the water uptake in highly compacted MX-80 clay takes place by diffusion at low external pressure. It means that wetting of the clay buffer in the deposition holes of a KBS-3 repository is very slow if the water pressure is low and that complete water saturation can take several tens of years if the initial degree of water saturation of the buffer clay and the ability of the rock to give off water are low. It has therefore been asked whether injection of water can raise the degree of water saturation and if a high water pressure in the nearfield can have the same effect. The present report describes attempts to moisten highly compacted blocks of MX-80 clay with a dry density of 1510 kg/m 3 by injecting water under a pressure of 650 kPa through a perforated injection pipe for 3 and 20 minutes, respectively. The interpretation was made by determining the water content of a number of samples located at different distances from the pipe. An attempt to interpret the pattern of distribution of injected uranium acetate solution showed that the channels into which the solution went became closed in a few minutes and that dispersion in the homogenized clay gave low U-concentrations. The result was that the water content increased from about 9 to about 11-12 % within a distance of about 1 centimeter from the injection pipe and to slightly more than 9 % at a distance of about 4-5 cm almost independently of the injection time. Complete water saturation corresponds to a water content of about 30 % and the wetting effect was hence small from a practical point of view. By use of microstructural models it can be shown that injected water enters only the widest channels that remain after the compaction and that these channels are quickly closed by expansion of the hydrating surrounding clay. Part of the particles that are thereby released become transported by the flowing water and cause clogging of the channels, which is

  12. Facility for generating crew waste water product for ECLSS testing

    Science.gov (United States)

    Buitekant, Alan; Roberts, Barry C.

    1990-01-01

    An End-use Equipment Facility (EEF) has been constructed which is used to simulate water interfaces between the Space Station Freedom Environmental Control and Life Support Systems (ECLSS) and man systems. The EEF is used to generate waste water to be treated by ECLSS water recovery systems. The EEF will also be used to close the water recovery loop by allowing test subjects to use recovered hygiene and potable water during several phases of testing. This paper describes the design and basic operation of the EEF.

  13. Large-leak sodium-water reaction analysis for steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakano, K; Shindo, Y; Hori, M

    1975-07-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  14. Large-leak sodium-water reaction analysis for steam generators

    International Nuclear Information System (INIS)

    Sakano, K.; Shindo, Y.; Hori, M.

    1975-01-01

    The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

  15. Liquid radioactive waste processing system for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    This Standard sets forth design, construction, and performance requirements, with due consideration for operation, of the Liquid Radioactive Waste Processing System for pressurized water reactor plants for design basis inputs. For the purpose of this Standard, the Liquid Radioactive Waste Processing System begins at the interfaces with the reactor coolant pressure boundary and the interface valve(s) in lines from other systems, or at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material; and it terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system, and at the point of recycle back to storage for reuse

  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. Pressure transient analysis in single and two-phase water by finite difference methods

    International Nuclear Information System (INIS)

    Berry, G.F.; Daley, J.G.

    1977-01-01

    An important consideration in the design of LMFBR steam generators is the possibility of leakage from a steam generator water tube. The ensuing sodium/water reaction will be largely controlled by the amount of water available at the leak site, thus analysis methods treating this event must have the capability of accurately modeling pressure transients through all states of water occurring in a steam generator, whether single or two-phase. The equation systems of the present model consist of the conservation equations together with an equation of state for one-dimensional homogeneous flow. These equations are then solved using finite difference techniques with phase considerations and non-equilibrium effects being treated through the equation of state. The basis for water property computation is Keenan's 'fundamental equation of state' which is applicable to single-phase water at pressures less than 1000 bars and temperatures less than 1300 0 C. This provides formulations allowing computation of any water property to any desired precision. Two-phase properties are constructed from values on the saturation line. The use of formulations permits the direct calculation of any thermodynamic property (or property derivative) to great precision while requiring very little computer storage, but does involve considerable computation time. For this reason an optional calculation scheme based on the method of 'transfinite interpolation' is included to give rapid computation in selected regions with decreased precision. The conservation equations were solved using the second order Lax-Wendroff scheme which includes wall friction, allows the formation of shocks and locally supersonic flow. Computational boundary conditions were found from a method-of-characteristics solution at the reservoir and receiver ends. The local characteristics were used to interpolate data from inside the pipe to the boundary

  18. A Numerical Investigation on the Effect of Gas Pressure on the Water Saturation of Compacted Bentonite-Sand Samples

    Directory of Open Access Journals (Sweden)

    Jiang-Feng Liu

    2017-01-01

    Full Text Available In deep geological disposal for high-level radioactive waste, the generated gas can potentially affect the sealing ability of bentonite buffers. There is a competition between water and gas: the former provides sealing by swelling bentonite, and the latter attempts to desaturate the bentonite buffer. Thus, this study focused on numerically modelling the coupling effects of water and gas on the water saturation and sealing efficiency of compacted bentonite-sand samples. Different gas pressures were applied to the top surface of an upper sample, whereas the water pressure on the bottom side of the lower sample was maintained at 4 MPa. The results indicated that gas pressure did not significantly affect the saturation of the bentonite-sand sample until 2 MPa. At 2 MPa, the degree of water saturation of the upper sample was close to 1.0. As the gas pressure increased, this influence was more apparent. When the gas pressure was 6 MPa or higher, it was difficult for the upper sample to become fully saturated. Additionally, the lower sample was desaturated due to the high gas pressure. This indicated that gas pressure played an important role in the water saturation process and can affect the sealing efficiency of bentonite-based buffer materials.

  19. A review on water fault diagnosis of PEMFC associated with the pressure drop

    International Nuclear Information System (INIS)

    Pei, Pucheng; Li, Yuehua; Xu, Huachi; Wu, Ziyao

    2016-01-01

    Highlights: • Reviewed the effect factors and estimations of pressure drop associated with water fault diagnosis. • Reviewed pressure drop-based water fault diagnosis using different indicators. • Deviation of pressure drop is used frequently to diagnose water fault. • Reviewed recovery strategies based on pressure drop used in commercial PEMFC. • Merits, demerits and application prospects of pressure drop-based water fault diagnosis are discussed. - Abstract: The pressure difference between the inlet and outlet of the reactant in fuel cells is called the pressure drop, which is related to the water amount inside the fuel cells. In recent years there have been many studies that used the pressure drop to detect the water content and diagnose water fault of proton exchange membrane fuel cells (PEMFCs). To our knowledge, there has not been a systematic review of these studies. In this paper, the effect variables of pressure drop are reviewed firstly. Then estimations of the theoretical pressure drop are reviewed mainly based on the following four aspects: Bernoulli’s equation, two-phase flow multiplier, Darcy’s law and artificial intelligence. Afterward, the water fault diagnosis based on the pressure drop using the following six indicators are reviewed: indicator of direct pressure drop, its deviation, frequency, multiplier, the ratio of pressure drop to flow rate and the flooding degree. In addition, the strategies of water fault recovery are also summarized. Finally the merits, demerits and application prospects of pressure drop-based water fault diagnosis are presented.

  20. Steam condensation behavior of high pressure water's blow down directly into water in containment under LOCA

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Ishida, Toshihisa; Yoritsune, Tsutomu; Kasahara, Y.

    1995-01-01

    JAERI has been conducting a design study of an advanced type Marine Reactor X (MRX) for merchant ships. By employing 'Integral type PWR', In-vessel type control rod drive systems', 'Water filled containment system' and 'Decay heat removal system by natural convection', MRX achieved a compact, light weight and highly safe plant. Experiments on steam condensation behavior of high pressure water's blow down into water have been conducted in order to investigate a major safety issue related to the design decision of 'Water filled containment system'. (author)

  1. THE EFFECTS OF SWIRL GENERATOR HAVING WINGS WITH HOLES ON HEAT TRANSFER AND PRESSURE DROP IN TUBE HEAT EXCHANGER

    Directory of Open Access Journals (Sweden)

    Zeki ARGUNHAN

    2006-02-01

    Full Text Available This paper examines the effect of turbulance creators on heat transfer and pressure drop used in concentric heat exchanger experimentaly. Heat exchanger has an inlet tube with 60 mm in diameter. The angle of swirl generators wings is 55º with each wing which has single, double, three and four holes. Swirl generators is designed to easily set to heat exchanger entrance. Air is passing through inner tube of heat exhanger as hot fluid and water is passing outer of inner tube as cool fluid.

  2. Neutron physics of a high converting advanced pressurized water reactor

    International Nuclear Information System (INIS)

    Berger, H.D.

    1985-01-01

    The neutron physics of an APWR are analysed by single pin-cell calculations as well as two-dimensional whole-reactor computations. The calculational methods of the two codes employed for this study, viz. the cell code SPEKTRA and the diffusion-burnup code DIBU, are presented in detail. The APWR-investigations carried out concentrate on the void coefficient characteristics of tight UO 2 /PuO 2 -lattices, control rod worths, burnup behaviour and spatial power distributions in APWR cores. The principal physics design differences between advanced pressurized water reactors and present-day PWRs are identified and discussed. (orig./HP) [de

  3. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    Khair, H. O. M.

    2012-06-01

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

  4. Contribution to a neutronic calculation scheme for pressurized water reactors

    International Nuclear Information System (INIS)

    Martin Del Campo, C.

    1987-01-01

    This research thesis aims at developing and validating the set of data and codes which build up the neutron computation scheme of pressurized water reactors. More precisely, it focuses on the improvement of the precision of calculation of command clusters (absorbing components which can be inserted into the core to control the reactivity), and on the modelling of reflector representation (material placed around the core and reflecting back the escaping neutrons). For the first case, a precise calculation is performed, based on the transport theory. For the second case, diffusion constants obtained in the previous case and simplified equations are used to reduce the calculation cost

  5. Advanced fuels for plutonium management in pressurized water reactors

    International Nuclear Information System (INIS)

    Vasile, A.; Dufour, Ph.; Golfier, H.; Grouiller, J.P.; Guillet, J.L.; Poinot, Ch.; Youinou, G.; Zaetta, A.

    2003-01-01

    Several fuel concepts are under investigation at CEA with the aim of manage plutonium inventories in pressurized water reactors. This options range from the use of mature technologies like MOX adapted in the case of MOX-EUS (enriched uranium support) and COmbustible Recyclage A ILot (CORAIL) assemblies to more innovative technologies using IMF like DUPLEX and advanced plutonium assembly (APA). The plutonium burning performances reported to the electrical production go from 7 to 60 kg (TW h) -1 . More detailed analysis covering economic, sustainability, reliability and safety aspects and their integration in the whole fuel cycle would allow identifying the best candidate

  6. Environment sensitive cracking in light water reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Haenninen, H.; Aho-Mantila, I.

    1985-01-01

    The purpose of the paper is to review the available methods and the most promising future possibilities of preventive maintenance to counteract the various forms of environment sensitive cracking of pressure boundary materials in light water reactors. Environment sensitive cracking is considered from the metallurgical, mechanical and environmental point of view. The main emphasis is on intergranular stress corrosion cracking of austenitic stainless steels and high strenght Ni-base alloys, as well as on corrosion fatigue of low alloy and stainless steels. Finally, some general ideas how to predict, reduce or eliminate environment sensitive cracking in service are presented

  7. Effect of Filters on the Noise Generated by Continuous Positive Airway Pressure Delivered via a Helmet

    Directory of Open Access Journals (Sweden)

    Ricardo Hernández-Molina

    2017-01-01

    Full Text Available Introduction: One of the problems that the delivery of continuous positive airway pressure (CPAP via a helmet poses is the generation of noise. The objective of our study was to assess the effect that the use of filter has on sound pressure levels generated by the delivery of positive airway pressure at different gas flow rates. Materials and Methods: Sound pressure levels generated by neonatal helmet CPAP delivery were measured at different gas flows (20, 30, and 40 l/min with and without a breathing filter. Noise intensity was measured by installing microphones in the inner ear of dummy heads wearing helmets. Results: The sound pressure level increased by 38% at a gas flow of 40 l/min, as compared to a gas flow of 20 l/min {74 dBA [interquartile range (IQR 2,2] vs 52 dBA (IQR 5,9, respectively}. Using the breathing filter as a diffuser has a variety of effects on sound pressure levels according to the gas flow rate. Conclusion: The intensity of the noise generated by helmet delivery of positive airway pressure depends on the type of helmet used, gas flow, and use or not of a diffuser filter. Breathing filters with gas flows over 30 l/min might not be recommended since they would not attenuate but will rather amplify sound pressure.

  8. Construction management of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bohra, S.A.; Sharma, P.D.

    2006-01-01

    Pandit Jawaharlal Nehru and Dr. Homi J. Bhabha, the visionary architects of Science and Technology of modern India foresaw the imperative need to establish a firm base for indigenous research and development in the field of nuclear electricity generation. The initial phase has primarily focused on the technology development in a systematic and structured manner, which has resulted in establishment of strong engineering, manufacturing and construction base. The nuclear power program started with the setting up of two units of boiling light water type reactors in 1969 for speedy establishment of nuclear technology, safety culture, and development of operation and maintenance manpower. The main aim at that stage was to demonstrate (to ourselves, and indeed to the rest of the world) that India, inspite of being a developing country, with limited industrial infrastructure and low capacity power grids, could successfully assimilate the high technology involved in the safe and economical operation of nuclear power reactors. The selection of a BWR was in contrast to the pressurized heavy water reactors (PHWR), which was identified as the flagship for the first stage of India's nuclear power program. The long-term program in three stages utilizes large reserves of thorium in the monazite sands of Kerala beaches in the third stage with first stage comprising of series of PHWR type plants with a base of 10,000 MW. India has at present 14 reactors in operation 12 of these being of PHWR type. The performance of operating units of 2720 MW has improved significantly with an overall capacity factor of about 90% in recent times. The construction work on eight reactor units with installed capacity of 3960 MW (two PHWRs of 540 MW each, four PHWRs of 220 MW each and two VVERs of 1000 MW each) is proceeding on a rapid pace with project schedules of less than 5 years from first pour of concrete. This is being achieved through advanced construction technology and management. Present

  9. TRAC-PF1: an advanced best-estimate computer program for pressurized water reactor analysis

    International Nuclear Information System (INIS)

    Liles, D.R.; Mahaffy, J.H.

    1984-02-01

    The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos National Laboratory to provide advanced best-estimate predictions of postulated accidents in light water reactors. The TRAC-PF1 program provides this capability for pressurized water reactors and for many thermal-hydraulic experimental facilities. The code features either a one-dimensional or a three-dimensional treatment of the pressure vessel and its associated internals; a two-phase, two-fluid nonequilibrium hydrodynamics model with a noncondensable gas field; flow-regime-dependent constitutive equation treatment; optional reflood tracking capability for both bottom flood and falling-film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions. This report describes the thermal-hydraulic models and the numerical solution methods used in the code. Detailed programming and user information also are provided

  10. Design of the US-CRBRP sodium/water reaction pressure relief system

    International Nuclear Information System (INIS)

    Kruger, G.B.; Murdock, T.B.; Rodwell, E.; Sane, J.O.

    1976-01-01

    Protection against intermediate sodium system overpressure from the sodium/water reaction associated with large leaks within the CRBRP Steam Generators is provided by the sodium/water reaction pressure relief system (SWRPRS). This system consists of rupture disks connected to the intermediate sodium piping adjacent to the inlet to the superheater and outlet from the evaporator modules. The rupture discs relieve into piping that leads to reaction produce separator tanks, which in turn are vented to a centrifugal separator and flare stack arranged to burn hydrogen gas exhausting into the atmosphere. Analyses have been conducted using the TRANSWRAP Computer Code to predict the system pressures and flow rates during the large leak event. Experimental tests to be conducted in the large leak test rig (LLTR) will be used to confirm the analysis techniques used in the design

  11. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  12. Pressurized water reactor iodine spiking behavior under power transient conditions

    International Nuclear Information System (INIS)

    Ho, J.C.

    1992-01-01

    The most accepted theory explaining the cause of pressurized water reactor iodine spiking is steam formation and condensation in damaged fuel rods. The phase transformation of the primary coolant from water to steam and back again is believed to cause the iodine spiking phenomenon. But due to the complex nature of the phenomenon, a comprehensive model of the behavior has not yet been successfully developed. This paper presents a new model based on an empirical approach, which gives a first-order estimation of the peak iodine spiking magnitude. Based on the proposed iodine spiking model, it is apparent that it is feasible to derive a correlation using the plant operating data base to monitor and control the peak iodine spiking magnitude

  13. Disintegration of liquid metals by low pressure water blasting

    International Nuclear Information System (INIS)

    Heshmatpour, B.; Copeland, G.L.

    1981-01-01

    The feasibility of disintegrating metals by a low cost system and subsequently incorporating them into grout mixtures has been demonstrated. A low pressure water blasting technique consisting of multiple nozzles and a converging-line jet stream was developed to disintegrate liquid metals and produce coarse metal powder and shot. Molten iron resulted in spherical shot, while copper, aluminum, and tin produced irregular shaped particles. The particle size was between 0.05 and 3 mm (0.002 and 0.1 in.), and about half the particles were smaller than 1 mm (0.04 in.) in all cases. The water consumption was rather low, while the production rate was relatively high. The method proved to be simple and reliable. The coarse metal powders were suspendable in grout fluids, indicating that they are probably disposable by the shale hydrofracture technique

  14. Some local dilution transient in a pressurized water reactor

    International Nuclear Information System (INIS)

    Jacobson, S.

    1989-01-01

    Reactivity accidents are important in the safety analysis of a pressurized water reactor. In this anlysis ejected control rod, steam line break, start of in-active loop and boron dilution accidents are usually dealt with. However, in the analysis is not included what reactivity excursions might happen when a zone,depleted of boron passes the reactor core. This thesis investigates during what operation and emergency conditions diluted zones might exist in a pressurized water reactor and what should be the maximum volumes for then. The limiting transport means are also established in terms of reactivty addition, for the depleted zones. In order to describe the complicated mixing process in the reactor vessel during such a transportation, a typical 3-loop reactor vessel has been modulated by means of TRAC-PF1's VESSEL component. Three cases have been analysed. In the first case the reactor is in a cold condition and the ractor coolant has boron concentration of 2000 ppm. To the reactor vessel is injected an clean water colume of 14 m 3 . In the two other cases the reactor is close to hot shutdown and borated to 850 ppm. To the reactor vessel is added 41 and 13 m 3 clean water, respectively. In the thesis is shown what spatial distribution the depleted zone gets when passing through the reactor vessel in the three cases. The boron concentration in the first case did not decrease the values which would bring the reactor to critical condition. For case two was shown by means of TRAC's point kinetics model that the reactor reaches prompt criticality after 16.03 seconds after starting of the reactor coolant pump. Another prompt criticality occured two seconds later. The total energy developed during the two power escalations were about 55 GJ. A comparision with the criteria used to evaluate the ejected control rod reactivity transient showed that none of these criteria were exceeded. (64 figs.)

  15. Sodium/water reactions in steam generators of liquid metal fast breeder reactors

    International Nuclear Information System (INIS)

    Hori, M.

    1980-01-01

    The status of the research and development on sodium/water reactions resulting from the leakage of water into sodium in LMFBR steam generators is reviewed. The importance of sodium/water reaction phenomena in the design and operation of steam generators is discussed. The effects of sodium/water reactions are evaluated and methods of protection against these phenomena are surveyed. The products of chemical reactions between sodium and water under steam generator conditions are H 2 , NaOH, Na 2 O and NaH. Together with the temperature rise due to the associated exothermic reaction, these reaction products cause effects such as self-wastage, single- and multi-target wastage, and rapid pressure increase, depending on the size of the leak hole or the magnitude of leak rate. As for the wastage phenomena of small leaks, the effects of various factors have been studied and experimental correlations, as well as some theoretical work, have been performed. To investigate the pressure phenomena of a large leak, large-scale tests have been conducted by various organizations, and the computer codes to analyse these phenomena have been developed and verified by experiments. In the design of steam generators, an initial failure up to a hypothetical double-ended guillotine rupture of a single heat transfer tube is widely used as the design basis leak. Protection systems for LMFBR plants consist of leak detection devices, leak termination devices, and reaction pressure relief devices. From analyses based on research and development activities, as well as from experience with leaks in steam generator test loops and reactor plants, it has been confirmed that protection systems can satisfactorily be designed to accommodate leak incidents in LMFBR plants. (author)

  16. Bioinspired Bifunctional Membrane for Efficient Clean Water Generation.

    Science.gov (United States)

    Liu, Yang; Lou, Jinwei; Ni, Mengtian; Song, Chengyi; Wu, Jianbo; Dasgupta, Neil P; Tao, Peng; Shang, Wen; Deng, Tao

    2016-01-13

    Solving the problems of water pollution and water shortage is an urgent need for the sustainable development of modern society. Different approaches, including distillation, filtration, and photocatalytic degradation, have been developed for the purification of contaminated water and the generation of clean water. In this study, we explored a new approach that uses solar light for both water purification and clean water generation. A bifunctional membrane consisting of a top layer of TiO2 nanoparticles (NPs), a middle layer of Au NPs, and a bottom layer of anodized aluminum oxide (AAO) was designed and fabricated through multiple filtration processes. Such a design enables both TiO2 NP-based photocatalytic function and Au NP-based solar-driven plasmonic evaporation. With the integration of these two functions into a single membrane, both the purification of contaminated water through photocatalytic degradation and the generation of clean water through evaporation were demonstrated using simulated solar illumination. Such a demonstration should also help open up a new strategy for maximizing solar energy conversion and utilization.

  17. Apparatus for localizing disturbances in pressurized water reactors (PWR)

    International Nuclear Information System (INIS)

    Sykora, D.

    1989-01-01

    The invention according to CS-PS 177386, entitled ''Apparatus for increasing the efficiency and passivity of the functioning of a bubbling-vacuum system for localizing disturbances in nuclear power plants with a pressurized water reactor'', concerns an important area of nuclear power engineering that is being developed in the RGW member countries. The invention solves the problems of increasing the reliability and intensification during the operation of the above very important system for guaranteeing the safety of the standard nuclear power plants of Soviet design. The essence of the invention consists in the installation of a simple passively operating supplementary apparatus. Consequently, the following can be observed in the system: first an improvement and simultaneous increase in the reliability of its function during the critical transition period, which follows the filling of the second space with air from the first space; secondly, elimination of the hitherto unavoidable initiating role of the active sprinkler-condensation device present; thirdly, a more effective performance and subjection of the elements to disintegration of the water flowing from the bubbling condenser into the first space; and fourthly, an enhanced utilization of the heat-conducting ability of the water reservoir of the bubbling condenser. Representatives of the supplementary apparatus are autonomous and local secondary systems of the sprinkler-sprayer without an insert, which spray the water under the effect of gravity. 1 fig

  18. Process for water-gas generation from degassed combustibles

    Energy Technology Data Exchange (ETDEWEB)

    1906-05-23

    A process for water-gas generation in a continuous operation from degassed combustibles in the lower part of a vertical exterior-heated retort, whose middle part can serve to degas the combustibles, is described. It is characterized in that the water vapor employed is obtained by vaporizing water in the upper part of the retort by means of the waste heat from the heating gases, which had effected the coking of the combustibles before the water-gas recovery or after the latter.

  19. Some problems of leaks in sodium-water steam generator

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Sergeev, G.V.; Sednev, A.R.; Makarov, V.M.

    1976-01-01

    The paper contains data on wastage of steam generator structural materials and high-nickel alloys in the zone of water leakage into sodium as well as investigation results for self-enlargement of water leaks into sodium through defects in these materials. It is shown that the rate of material damage in the zone of sodium-water reaction and in the channel with water leaking-out decreases with increasing nickel content in steels and strongly depends on sodium temperature. The paper presents experimentally obtained dependences of leakage self-enlargement rates on sodium temperature and leakage size

  20. Acceptance test procedure for High Pressure Water Jet System

    International Nuclear Information System (INIS)

    Crystal, J.B.

    1995-01-01

    The overall objective of the acceptance test is to demonstrate a combined system. This includes associated tools and equipment necessary to perform cleaning in the 105 K East Basin (KE) for achieving optimum reduction in the level of contamination/dose rate on canisters prior to removal from the KE Basin and subsequent packaging for disposal. Acceptance tests shall include necessary hardware to achieve acceptance of the cleaning phase of canisters. This acceptance test procedure will define the acceptance testing criteria of the high pressure water jet cleaning fixture. The focus of this procedure will be to provide guidelines and instructions to control, evaluate and document the acceptance testing for cleaning effectiveness and method(s) of removing the contaminated surface layer from the canister presently identified in KE Basin. Additionally, the desired result of the acceptance test will be to deliver to K Basins a thoroughly tested and proven system for underwater decontamination and dose reduction. This report discusses the acceptance test procedure for the High Pressure Water Jet

  1. Effect of Stand-Off Distance on Impact Pressure of High Speed Water Jets

    Science.gov (United States)

    Sittiwong, Wuttichai; Seehanam, Wirapan; Pianthong, Kulachate; Matthujak, Anirut

    2010-06-01

    High speed liquid jets may be applied to jet cutting, drilling and cleaning. Recently, in the automotive industries, the spray injection pressure becomes higher and higher to enhance the fuel mixing for the improved combustion efficiency. However, the ultra high injection pressure may cause the damage to the nozzle and also the combustion chamber. In the medical application, the high speed liquid injection might be applied for the drug delivery through the skin where the needle is not required anymore. From the above mentioned application, the investigation on the impact pressure of the high speed liquid jet relative to the stand-off distant is significant. The high speed liquid jets are generated by the projectile impact driven method. The high speed projectile is launched by the horizontal single stage powder gun. The experimental study focuses on the stand-off between 1.5 cm to 6.0 cm, while the nozzle contains approximately 1.5cm3 of water in its cavity. The nozzle conical angles are 30° and 60° with the orifice diameter of 0.7 mm. The jet velocities are measured by laser beam interruptions method. The target material is the Polymethyl Methacrylate (PMMA) which the impact pressure is measured by using a piezoelectric Polyvinylidene Fluoride (PVDF) film. From the experiments, the maximum water jet velocity of 2290 m/s can be obtained from the 30° conical angle nozzle. The maximum impact pressures of nozzle conical angle of 30° and 60° are 3.4 GPa and 2.6 GPa respectively, at stand-off distance 3 cm. However, at the stand-off distance more than 3 cm, the impact pressure significantly decreases, because of aerodynamic drag, jets core break-up, and atomization of the water.

  2. Experimental study on vapor explosion induced by pressure pulse in coarse mixing of hot molten metal and water

    International Nuclear Information System (INIS)

    Inoue, A.; Tobita, Y.; Aritomi, M.; Takahashi, M.; Matsuzaki, M.

    2004-01-01

    An experimental study was done to investigate characteristics of metal-water interaction, when a mount of hot liquid metal is injected into the water. The test section is a vertical shock tube of 60mm in inner diameter and 1200mm in length. A special injector which is designed to inject hot metal of controlled volume and flow rate is attached at the top of the tube. When the hot metal is injected in the water and comes down at a position of the test vessel, a trigger pressure pulse is generated at the bottom of the test tube. Local transient pressures along the tube are measured by piezo pressure transducers. The following items were investigated in the experiment; 1) The criteria to cause a vapor explosion, 2) Transient behaviors and propagation characteristics of pressure wave in the mixing region. 3) Effects of triggering pulse, injection temperature and mass of hot molten metal on the peak pressure. The probability of the vapor explosion jumped when the interface temperature at the molten metal-water direct contact is higher than the homogeneous nucleation temperature of water and the triggering pulse becomes larger than 0.9MPa. Two types of the pressure propagation modes are observed, one is the detonative mode with a sharp rise and other is usual pressure mode with a mild rise. (author)

  3. Influence of water depth on the sound generated by air-bubble vibration in the water musical instrument

    Science.gov (United States)

    Ohuchi, Yoshito; Nakazono, Yoichi

    2014-06-01

    We have developed a water musical instrument that generates sound by the falling of water drops within resonance tubes. The instrument can give people who hear it the healing effect inherent in the sound of water. The sound produced by falling water drops arises from air- bubble vibrations. To investigate the impact of water depth on the air-bubble vibrations, we conducted experiments at varying values of water pressure and nozzle shape. We found that air-bubble vibration frequency does not change at a water depth of 50 mm or greater. Between 35 and 40 mm, however, the frequency decreases. At water depths of 30 mm or below, the air-bubble vibration frequency increases. In our tests, we varied the nozzle diameter from 2 to 4 mm. In addition, we discovered that the time taken for air-bubble vibration to start after the water drops start falling is constant at water depths of 40 mm or greater, but slower at depths below 40 mm.

  4. Pressure pulses generated by gas released from a breached fuel element

    International Nuclear Information System (INIS)

    Wu, T.S.

    1979-01-01

    In experimental measurements of liquid pressure pulses generated by rapid release of gas from breached fuel elements in a nuclear reactor, different peak pressures were observed at locations equidistant from the origin of the release. Using the model of a submerged spherical bubble with a nonstationary center, this analysis predicts not only that the peak pressure would be higher at a point in front of the advancing bubble than that at a point the same distance behind the bubble origin, but also that the pressure pulse in front of the bubble reaches its peak later than the pulse behind the origin

  5. High Pressure Water Stripping Using Multi-Orifice Nozzles

    Science.gov (United States)

    Hoppe, David

    1999-01-01

    The use of multi-orifice rotary nozzles greatly increases the speed and stripping effectiveness of high pressure water blasting systems, but also greatly increases the complexity of selecting and optimizing the operating parameters. The rotational speed of the nozzle must be coupled with its transverse velocity as it passes across the surface of the substrate being stripped. The radial and angular positions of each orifice must be included in the analysis of the nozzle configuration. Orifices at the outer edge of the nozzle head move at a faster rate than the orifices located near the center. The energy transmitted to the surface from the impact force of the water stream from an outer orifice is therefore spread over a larger area than energy from an inner orifice. Utilizing a larger diameter orifice in the outer radial positions increases the total energy transmitted from the outer orifice to compensate for the wider distribution of energy. The total flow rate from the combination of all orifices must be monitored and should be kept below the pump capacity while choosing orifice to insert in each position. The energy distribution from the orifice pattern is further complicated since the rotary path of all the orifices in the nozzle head pass through the center section. All orifices contribute to the stripping in the center of the path while only the outer most orifice contributes to the stripping at the edge of the nozzle. Additional orifices contribute to the stripping from the outer edge toward the center section. With all these parameters to configure and each parameter change affecting the others, a computer model was developed to track and coordinate these parameters. The computer simulation graphically indicates the cumulative affect from each parameter selected. The result from the proper choices in parameters is a well designed, highly efficient stripping system. A poorly chosen set of parameters will cause the nozzle to strip aggressively in some areas

  6. Research on axial total pressure distributions of sonic steam jet in subcooled water

    International Nuclear Information System (INIS)

    Wu Xinzhuang; Li Wenjun; Yan Junjie

    2012-01-01

    The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

  7. Computer code for the analysis of destructive pressure generation process during a fuel failure accident, PULSE-2

    International Nuclear Information System (INIS)

    Fujishiro, Toshio

    1978-03-01

    The computer code PULSE-2 has been developed for the analysis of pressure pulse generation process when hot fuel particles come into contact with the coolant in a fuel rod failure accident. In the program, it is assumed that hot fuel fragments mix with the coolant instantly and homogeneously in the failure region. Then, the rapid vaporization of the coolant and transient pressure rise in failure region, and the movement of ejected coolant slugs are calculated. The effect of a fuel-particle size distribution is taken into consideration. Heat conduction in the fuel particles and heat transfer at fuel-coolant interface are calculated. Temperature, pressure and void fraction in the mixed region are calculated from the average enthalpy. With physical property subroutines for liquid sodium and water, the model is usable for both LMFBR and LWR conditions. (auth.)

  8. Wastage of Steam Generator Tubes by Sodium-Water Reaction

    International Nuclear Information System (INIS)

    Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

    2010-01-01

    The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

  9. Secondary-side feed-and-bleed effectiveness in pressurized water reactors

    International Nuclear Information System (INIS)

    Annunziato, A.

    1994-01-01

    An analysis is presented of physical phenomena occurring during secondary-side feed-and-bleed, which is being considered as an accident management procedure for pressurized water reactors. Problem areas related to the reliability of this procedure are identified and discussed. Secondary-side feed-and-bleed, as examined in the LOBI (which is the Light Water Reactor Off-Normal Behavior Investigation) integral system test facility, was not successful because of a delay in the secondary-side heat removal caused by the release of stored heat from the steam generator downcomer wall. Countercurrent flow limiting in the hot leg was another major phenomenon that can influence the effectiveness of this procedure. The discussion of the experimental results is complemented by relevant calculations by RELAP5/MOD3. In general, it was found that the effectiveness of the feed-and-bleed procedure is maximized if the number of steam generators used to implement it is reduced

  10. Development status and application prospect of supercritical-pressure light water cooled reactor

    International Nuclear Information System (INIS)

    Li Manchang; Wang Mingli

    2006-01-01

    The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

  11. Detection of gaseous heavy water leakage points in CANDU 6 pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Park, T-K.; Jung, S-H.

    1996-01-01

    During reactor operation, the heavy water filled primary coolant system in a CANDU 6 Pressurized Heavy Water (PHWR) may leak through routine operations of the plant via components, mechanical joints, and during inadvertent operations etc. Early detection of leak points is therefore important to maintain plant safety and economy. There are many independent systems to monitor and recover heavy water leakage in a CANDU 6 PHWR. Methodology for early detection based on operating experience from these systems, is investigated in this paper. In addition, the four symptoms of D 2 O leakage, the associated process for clarifying and verifying the leakage, and the probable points of leakage are discussed. (author)

  12. Support vector regression model based predictive control of water level of U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kavaklioglu, Kadir, E-mail: kadir.kavaklioglu@pau.edu.tr

    2014-10-15

    Highlights: • Water level of U-tube steam generators was controlled in a model predictive fashion. • Models for steam generator water level were built using support vector regression. • Cost function minimization for future optimal controls was performed by using the steepest descent method. • The results indicated the feasibility of the proposed method. - Abstract: A predictive control algorithm using support vector regression based models was proposed for controlling the water level of U-tube steam generators of pressurized water reactors. Steam generator data were obtained using a transfer function model of U-tube steam generators. Support vector regression based models were built using a time series type model structure for five different operating powers. Feedwater flow controls were calculated by minimizing a cost function that includes the level error, the feedwater change and the mismatch between feedwater and steam flow rates. Proposed algorithm was applied for a scenario consisting of a level setpoint change and a steam flow disturbance. The results showed that steam generator level can be controlled at all powers effectively by the proposed method.

  13. Innovated feed water distributing system of VVER steam generators

    International Nuclear Information System (INIS)

    Matal, O.; Sousek, P.; Simo, T.; Lehota, M.; Lipka, J.; Slugen, V.

    2000-01-01

    Defects in feed water distributing system due to corrosion-erosion effects have been observed at many VVER 440 steam generators (SG). Therefore analysis of defects origin and consequently design development and testing of a new feed water distributing system were performed. System tests in-situ supported by calculations and comparison of measured and calculated data were focused on demonstration of long term reliable operation, definition of water flow and water chemical characteristics at the SG secondary side and their measurements and study of dynamic characteristics needed for the innovated feed water distributing system seismic features approval. The innovated feed water distributing system was installed in the SGs of two VVER units already. (author)

  14. Influence of the pressure holding time on strain generation in fuel injection lines

    International Nuclear Information System (INIS)

    Basara, Adis; Alt, Nicolas; Schluecker, Eberhard

    2011-01-01

    An influence of the pressure holding time on residual strain generation during the autofrettage process was studied experimentally for the first time in the present work. It is the state of the art that fuel injection lines are held at the autofrettage pressure for only a few seconds in an industrial production. In doing so, it is assumed that a desirable residual stress-strain pattern is generated. However, the results of the experimental investigations outlined in this work indicated that completion of the plastic deformation caused by the autofrettage process and generation of the desirable stress-strain pattern require a much longer period. As shown, a third-order polynomial equation best described the interdependence between the time required for the completion of the process, the corresponding autofrettage pressure and the generated strain state. The method presented can be used as a tool for the determination of the optimal autofrettage process parameters in industrial production of fuel injection lines.

  15. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    Energy Technology Data Exchange (ETDEWEB)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H. [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH{sub T} was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle.

  16. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    International Nuclear Information System (INIS)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H.

    2016-01-01

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH_T was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle

  17. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  18. Pressurized water reactor simulator. Workshop material. 2. ed

    International Nuclear Information System (INIS)

    2005-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and educational material and sponsors courses and workshops. The workshops are in two parts: techniques and tools for reactor simulator development. And the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 reactor department simulator from the Moscow Engineering and Physics Institute, the Russian Federation is presented in the IAEA Training Course Series No. 21, 2nd edition, 'WWER-1000 Reactor Simulator' (2005). Course material for workshops using a boiling water reactor simulator developed for the IAEA by Cassiopeia Technologies Incorporated of Canada (CTI) is presented in the IAEA publication: Training Course Series No.23, 2nd edition, 'Boiling Water Reactor Simulator' (2005). This report consists of course material for workshops using a pressurized water reactor simulator

  19. Steam generators in indirect-cycle water-cooled reactors

    International Nuclear Information System (INIS)

    Fajeau, M.

    1976-01-01

    In the indirect cycle water-cooled nuclear reactors, the steam generators are placed between the primary circuit and the turbine. They act both as an energy transmitter and as a leaktigh barrier against fission or corrosion products. Their study is thus very important from a performance and reliability point of view. Two main types are presented here: the U-tube and the once-through steam generators [fr

  20. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  1. The European Pressurized Water Reactor (EPR). State of the art after the preliminary design phase

    International Nuclear Information System (INIS)

    Bouteille, F.; Schneider, D.

    2002-01-01

    The European Pressurized Water Reactor (EPR) is an evolutionary development of the pressurized water reactor product lines built by Framatome and Siemens in France and Germany. Under the technical leadership of both nuclear power plant suppliers (now merged in Framatome ANP, a joint venture of AREVA and Siemens) the future-oriented plant concepts was developed in close cooperation with German and French utilities and in compliance with the European Utility Requirements. The EPR has safety features with which even extremely improbable, beyond design-basis events can be controlled and their effects can be limited to such an extent that no emergency response actions need be taken outside of the immediate plant site. This also means that safety systems prevent containment failure even in the improbable case of a core melt. This was confirmed by the French and German reactor safety authorities. The selected high thermal output also insures the economic viability of the innovative reactor concept, so that the power generation costs which can be achieved with the EPR will be absolutely competitive with those of fossil energy carriers. Framatome ANP has thus developed a pressurized water reactor ready for offer at the right time, which can completely fulfill the most rigorous requirements in terms of nuclear safety and economy. (Author)

  2. Simulation of pressurized water reactor in accidental state

    International Nuclear Information System (INIS)

    Chakir, E.

    1994-01-01

    The aim of this work is to develop the 1300 MWe 4 loops 'PWR' simulator called 'SATRAPE', witch the adopted physics modelisation allows a simplified neutronic calculation, and focus essentially on the reactor thermal hydraulic behavior in the case of the following accidents: - Loss of Coolant Accident (LOCA). - Steam Generator Tube Failure (SGTF). - Steam Line Break (SLB). In case of the 'LOCA' or 'SLB' accident, this modelisation enables the calculation of the pressure and the temperature in the containment building, and also the debit of the released dose in this latter in case of the 'LOCA' accident. The adopted models are relatively simple so as to allow an explicit resolve. In SATRAPE, two graphical interfaces enables to launch orders, whereas the other permits to visualize, the principal state variables of installations. The results obtained show a very good consistency with the envisaged commonly scenario at the time of the considered accidents. 33 refs., 52 figs., 1 tab. (author)

  3. Diagnostic system for primary circuits of pressurized-water reactors

    International Nuclear Information System (INIS)

    Liska, J.; Majer, J.

    1983-01-01

    The diagnostic system monitors the reactor, the main circulating pipe, the main circulating pump, the main shut-off valve, the steam generator and the pressurizer. Diagnostic signals are obtained from the sensors designed for operation measurements and from sensors for special diagnostic purposes. The following operations are carried out: detection of dangerous dynamic stress of components, detection of damage to functional surfaces of components, detection of occurrence and propagation of defects in component materials, detection of loose particles and foreign bodies, detection of coolant leakage, detection of coolant boiling in the core and detection of impermissible non-homogeneities of fields of physical quantities in the core. The diagnostic system comprises: monitoring, classification of properly investigated effects, periodical tracing and long-term tracing. The operational diagnostics system developed by the SKODA Concern consists of a vibration monitoring system, a spectral analysis system and a central evaluation system. (M.D.)

  4. Contemporary pressurized water reactor technology in the world

    International Nuclear Information System (INIS)

    Komarek, A.

    1991-01-01

    The recent political events enabled Czechoslovak industrial companies to come into direct contact with leading western companies involved in pressurized water ractor technology. A survey is presented of the present situation at the world market of PWR type nuclear power plant suppliers and suppliers of fuel cycle services. Information is given on the potential bids for the next Czechoslovak nuclear power plants with PWR reactors. Economic aspects of the potential bids are presented including some considerations about the participation of the Czechoslovak nuclear industry as a supplier of the reactor for the future power plants. Main technical parameters are listed of PWR units with an output about 1000 MW supplied by Westinghouse EC, ABB -Combustion Engineering and Siemens AG. At present, the bids for new Czechoslovak nuclear power plants are being evaluated. No information on terms of the bids actually coming from foreign companies is used in the article. (Z.S.). 9 figs., 5 tabs

  5. Computer code for simulating pressurized water reactor core

    International Nuclear Information System (INIS)

    Serrano, A.M.B.

    1978-01-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numerically. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistance added to the film coefficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (author)

  6. Code for the core simulation in pressurized water reactors

    International Nuclear Information System (INIS)

    Serrano, M.A.B.

    1978-08-01

    A computer code was developed for the simulation of the steady-state and transient behaviour of the average channel of a Pressurizer Water Reactor core. Point kinetics equations were used with the reactivity calculated for average temperatures in the channel with the fuel and moderator temperature feedbacks. The radial heat conduction equation in the fuel was solved numericaly. For calculating the thermodynamic properties of the coolant, the fundamental equations of conservation (mass, energy and momentum) were solved. The gap and clad were treated as a resistence added to the film coeficient. The fuel system equations were decoupled from the coolant equations. The program permitted the changes in the heat transfer correlations and the flow patterns along the coolant channel. Various test were performed to determine the steady-state and transient response employing the PWR core simulator developed, obtaining results with adequate precision. (Author) [pt

  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. 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

  9. Testing of a portable ultrahigh pressure water decontamination system (UHPWDS)

    International Nuclear Information System (INIS)

    Lovell, A.; Dahlby, J.

    1996-02-01

    This report describes the tests done with a portable ultrahigh pressure water decontamination system (UHPWDS) on highly radioactively contaminated surfaces. A small unit was purchased, modified, and used for in-situ decontamination to change the waste level of the contaminated box from transuranic (TRU) waste to low- level waste (LLW). Low-level waste is less costly by as much as a factor of five or more if compared with TRU waste when handling, storage, and disposal are considered. The portable unit we tested is commercially available and requires minimal utilities for operation. We describe the UHPWDS unit itself, a procedure for its use, the results of the testing we did, and conclusions including positive and negative aspects of the UHPWDS

  10. Thermal Hydraulic Integral Effect Tests for Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Baek, Won Pil; Song, C. H.; Kim, Y. S.

    2007-02-01

    The objectives of the project are to construct a thermal-hydraulic integral effect test facility and to perform the tests for design, operation, and safety regulation of pressurized water reactors. In the first phase of this project (1997.8∼2002.3), the basic technology for thermal-hydraulic integral effect tests was established and the basic design of the test facility was accomplished. In the second phase (2002.4∼2005.2), an optimized design of the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation) was established and the construction of the facility was almost completed. In the third phase (2005.3∼2007.2), the construction and commission tests of the ATLAS are to be completed and some first-phase tests are to be conducted

  11. Neurocontrol of Pressurized Water Reactors in Load-Follow Operations

    International Nuclear Information System (INIS)

    Lin Chaung; Shen Chihming

    2000-01-01

    The neurocontrol technique was applied to control a pressurized water reactor (PWR) in load-follow operations. Generalized learning or direct inverse control architecture was adopted in which the neural network was trained off-line to learn the inverse model of the PWR. Two neural network controllers were designed: One provided control rod position, which controlled the axial power distribution, and the other provided the change in boron concentration, which adjusted core total power. An additional feedback controller was designed so that power tracking capability was improved. The time duration between control actions was 15 min; thus, the xenon effect is limited and can be neglected. Therefore, the xenon concentration was not considered as a controller input variable, which simplified controller design. Center target strategy and minimum boron strategy were used to operate the reactor, and the simulation results demonstrated the effectiveness and performance of the proposed controller

  12. Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

    International Nuclear Information System (INIS)

    Tingkuan, C.; Xuanzheng, C.

    1987-01-01

    It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

  13. Reducing energy consumption and leakage by active pressure control in a water supply system

    NARCIS (Netherlands)

    Bakker, M.; Rajewicz, T.; Kien, H.; Vreeburg, J.H.G.; Rietveld, L.C.

    2013-01-01

    WTP Gruszczyn supplies drinking water to a part of the city of Pozna?, in the Midwest of Poland. For the optimal automatic pressure control of the clear water pumping station, nine pressure measuring points were installed in the distribution network, and an active pressure control model was

  14. Basal interstitial water pressure in laboratory debris flows over a rigid bed in an open channel

    Directory of Open Access Journals (Sweden)

    N. Hotta

    2012-08-01

    Full Text Available Measuring the interstitial water pressure of debris flows under various conditions gives essential information on the flow stress structure. This study measured the basal interstitial water pressure during debris flow routing experiments in a laboratory flume. Because a sensitive pressure gauge is required to measure the interstitial water pressure in shallow laboratory debris flows, a differential gas pressure gauge with an attached diaphragm was used. Although this system required calibration before and after each experiment, it showed a linear behavior and a sufficiently high temporal resolution for measuring the interstitial water pressure of debris flows. The values of the interstitial water pressure were low. However, an excess of pressure beyond the hydrostatic pressure was observed with increasing sediment particle size. The measured excess pressure corresponded to the theoretical excess interstitial water pressure, derived as a Reynolds stress in the interstitial water of boulder debris flows. Turbulence was thought to induce a strong shear in the interstitial space of sediment particles. The interstitial water pressure in boulder debris flows should be affected by the fine sediment concentration and the phase transition from laminar to turbulent debris flow; this should be the subject of future studies.

  15. Analytical study on water hammer pressure in pressurized conduits with a throttled surge chamber for slow closure

    Directory of Open Access Journals (Sweden)

    Yong-liang Zhang

    2010-06-01

    Full Text Available This paper presents an analytical investigation of water hammer in a hydraulic pressurized pipe system with a throttled surge chamber located at the junction between a tunnel and a penstock, and a valve positioned at the downstream end of the penstock. Analytical formulas of maximum water hammer pressures at the downstream end of the tunnel and the valve were derived for a system subjected to linear and slow valve closure. The analytical results were then compared with numerical ones obtained using the method of characteristics. There is agreement between them. The formulas can be applied to estimating water hammer pressure at the valve and transmission of water hammer pressure through the surge chamber at the junction for a hydraulic pipe system with a surge chamber.

  16. Water-bearing, high-pressure Ca-silicates

    Science.gov (United States)

    Németh, Péter; Leinenweber, Kurt; Ohfuji, Hiroaki; Groy, Thomas; Domanik, Kenneth J.; Kovács, István J.; Kovács, Judit S.; Buseck, Peter R.

    2017-07-01

    Water-bearing minerals provide fundamental knowledge regarding the water budget of the mantle and are geophysically significant through their influence on the rheological and seismic properties of Earth's interior. Here we investigate the CaO-SiO2-H2O system at 17 GPa and 1773 K, corresponding to mantle transition-zone condition, report new high-pressure (HP) water-bearing Ca-silicates and reveal the structural complexity of these phases. We document the HP polymorph of hartrurite (Ca3SiO5), post-hartrurite, which is tetragonal with space group P4/ncc, a = 6.820 (5), c = 10.243 (8) Å, V = 476.4 (8) Å3, and Z = 4, and is isostructural with Sr3SiO5. Post-hartrurite occurs in hydrous and anhydrous forms and coexists with larnite (Ca2SiO4), which we find also has a hydrous counterpart. Si is 4-coordinated in both post-hartrurite and larnite. In their hydrous forms, H substitutes for Si (4H for each Si; hydrogrossular substitution). Fourier transform infrared (FTIR) spectroscopy shows broad hydroxyl absorption bands at ∼3550 cm-1 and at 3500-3550 cm-1 for hydrous post-hartrurite and hydrous larnite, respectively. Hydrous post-hartrurite has a defect composition of Ca2.663Si0.826O5H1.370 (5.84 weight % H2O) according to electron-probe microanalysis (EPMA), and the Si deficiency relative to Ca is also observed in the single-crystal data. Hydrous larnite has average composition of Ca1.924Si0.851O4H0.748 (4.06 weight % H2O) according to EPMA, and it is in agreement with the Si occupancy obtained using X-ray data collected on a single crystal. Superlattice reflections occur in electron-diffraction patterns of the hydrous larnite and could indicate crystallographic ordering of the hydroxyl groups and their associated cation defects. Although textural and EPMA-based compositional evidence suggests that hydrous perovskite may occur in high-Ca-containing (or low silica-activity) systems, the FTIR measurement does not show a well-defined hydroxyl absorption band for this

  17. Portable device for generation of ultra-pure water vapor feeds

    Science.gov (United States)

    Velin, P.; Stenman, U.; Skoglundh, M.; Carlsson, P.-A.

    2017-11-01

    A portable device for the generation of co-feeds of water vapor has been designed, constructed, and evaluated for flexible use as an add-on component to laboratory chemical reactors. The vapor is formed by catalytic oxidation of hydrogen, which benefits the formation of well-controlled minute concentrations of ultra-pure water. Analysis of the effluent stream by on-line mass spectrometry and Fourier transform infrared spectroscopy confirms that water vapor can be, with high precision, generated both rapidly and steadily over extended periods in the range of 100 ppm to 3 vol. % (limited by safety considerations) using a total flow of 100 to 1500 ml/min at normal temperature and pressure. Further, the device has been used complementary to a commercial water evaporator and mixing system to span water concentrations up to 12 vol. %. Finally, an operando diffuse reflective infrared Fourier transform spectroscopic measurement of palladium catalysed methane oxidation in the absence and presence of up to 1.0 vol. % water has been carried out to demonstrate the applicability of the device for co-feeding well-controlled low concentrations of water vapor to a common type of spectroscopic experiment. The possibilities of creating isotopically labeled water vapor as well as using tracer gases for dynamic experiments are discussed.

  18. Set up for simultaneous water desalination and power generation

    International Nuclear Information System (INIS)

    Hasan, S.W.; Mookhi, M.B.; Sadiq, M.A.; Hasan, Z.; Zaidi, S.I.; Shah, W.A.

    2010-01-01

    Instead of following the conventional fuel oriented power generation methods and dissipating its heat into environment, we evaporate saline water into steam and use its energy to generate power. Using this scheme would make sea water usable in power generation which at the moment is only being used for cooling purposes in the power plants. The steam used for generating electricity is eventually collected, condensed and used for potable purposes. The proposed scheme may be seen as Steam Power Generation with additional feature of desalination. We set up an experimental test bed in order to calculate the electric power available using this scheme. To ensure safety for human consumption, we also perform chemical tests on the desalinated water to see whether it is fit to be used for drinking and agricultural purposes. Our conclusions are based on actual experiments and laboratory tests; procedures outlined here may be used at larger scale for more in-depth analyses. We also highlight future extensions and modifications in this work. (author)

  19. Pressured drilling riser design for drilling in ultra deep water with surface bop

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, J.; Morrison, D.; Efthymiou, M.; Lo, K.H. [Shell Global Solutions, 78 - Velizy Villacoublay (France); Magne, E.; Leach, C. [Shell Internationale Exploration and Production (Netherlands)

    2002-12-01

    In conventional drilling with a semi-submersible rig valuable rig time is used to run and retrieve the BOP and its accessories on the seabed, and this time increases with water depth. Furthermore, use of the conventional sub-sea BOP requires a large-diameter riser, which requires substantial rig storage and deck load capacity prior to installation. It also requires high riser-tensioning capacity or additional buoyancy. Thus as the water depth increases, it leads to a need for heavy duty 4. and 5. generation rigs with escalation in costs. The high cost of deep-water drill rigs is leading to the development of Surface BOP technology. In this development, the BOP is placed above sea level and the riser is simply a continuation of the casing (typical diameter 13-3/8''). This eliminates the need for a heavy 21'' riser and for running the BOP to the sea bed and retrieving it. Moreover, the reduced tension requirement for the smaller riser extends the water depth capability of 3. generation drilling semi-submersibles, enabling them to drill in deeper waters. A critical success factor for this development is the ability to design the riser/casing to withstand high internal pressures due to well kicks, in addition to environmental loads, and to restrict vessel offsets within certain limits so as not to overload the riser under the prevailing weather conditions. This paper addresses the design considerations of a pressured drilling riser that can be used with a surface BOP in deep-water. Key design issues that are sensitive to ultra-deep-water applications are discussed. The technical aspects of using (disposable) standard casing with threaded connector for the drilling riser are discussed, with a particular emphasis on the connector fatigue-testing program to quantify the stress concentration factor for fatigue design. Emerging composite material offers some alternatives to the steel riser when drilling in ultra-deep water Design issues related to the

  20. Water leak detection in steam generator of SUPER PHENIX

    International Nuclear Information System (INIS)

    Brunet, M.; Garnaud, P.; Ghaleb, D.; Kong, N.

    1988-01-01

    With the intent of detecting water leaks inside steam generators, we developed a third system, called acoustic detector, to complement hydrogen detectors and rupture disks (burst disks). The role of the acoustic system is to enable rapid intervention in the event of a leak growing rapidly which could rupture neighbouring tubes. In such a case, the detectable flow rate of the leak varies from a few tens of g/s to a few hundred g/s. At the SUPER PHENIX, three teams work in [20-100 kHz] and CEA/STA* [50-300 kHz]. The simulation of water leaks in the steam generator by the argon injections performed to date at 50% of the rated power has shown promising results. An anomaly in the evolution of the background noise at more than 50% loading of one of the two instrumented steam generators would make difficult any extrapolation to full power behaviour. (author)

  1. Behavior of stainless steels in pressurized water reactor primary circuits

    International Nuclear Information System (INIS)

    Féron, D.; Herms, E.; Tanguy, B.

    2012-01-01

    Stainless steels are widely used in primary circuits of pressurized water reactors (PWRs). Operating experience with the various grades of stainless steels over several decades of years has generally been excellent. Nevertheless, stress corrosion failures have been reported in few cases. Two main factors contributing to SCC susceptibility enhancement are investigated in this study: cold work and irradiation. Irradiation is involved in the stress corrosion cracking and corrosion of in-core reactor components in PWR environment. Irradiated assisted stress corrosion cracking (IASCC) is a complex and multi-physics phenomenon for which a predictive modeling able to describe initiation and/or propagation is not yet achieved. Experimentally, development of initiation smart tests and of in situ instrumentation, also in nuclear reactors, is an important axis in order to gain a better understanding of IASCC kinetics. A strong susceptibility for SCC of heavily cold worked austenitic stainless steels is evidenced in hydrogenated primary water typical of PWRs. It is shown that for a given cold-working procedure, SCC susceptibility of austenitic stainless steels materials increases with increasing cold-work. Results have shown also strong influences of the cold work on the oxide layer composition and of the maximum stress on the time to fracture.

  2. Taipower's reload safety evaluation methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Ping-Hue; Yang, Y.S.

    1996-01-01

    For Westinghouse pressurized water reactors (PWRs) such as Taiwan Power Company's (TPC's) Maanshan Units 1 and 2, each of the safety analysis is performed with conservative reload related parameters such that reanalysis is not expected for all subsequent cycles. For each reload cycle design, it is required to perform a reload safety evaluation (RSE) to confirm the validity of the existing safety analysis for fuel cycle changes. The TPC's reload safety evaluation methodology for PWRs is based on 'Core Design and Safety Analysis Package' developed by the TPC and the Institute of Nuclear Energy Research (INER), and is an important portion of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors'. The Core Management System (CMS) developed by Studsvik of America, the one-dimensional code AXINER developed by TPC, National Tsinghua University and INER, and a modified version of the well-known subchannel core thermal-hydraulic code COBRAIIIC are the major computer codes utilized. Each of the computer models is extensively validated by comparing with measured data and/or vendor's calculational results. Moreover, parallel calculations have been performed for two Maanshan reload cycles to validate the RSE methods. The TPC's in-house RSE tools have been applied to resolve many important plant operational issues and plant improvements, as well as to verify the vendor's fuel and core design data. (author)

  3. Imaging optical probe for pressurized steam-water environment

    International Nuclear Information System (INIS)

    Donaldson, M.R.; Pulfrey, R.E.

    1979-01-01

    An air-cooled imaging optical probe, with an outside diameter of 25.4 mm, has been developed to provide high resolution viewing of flow regimes in a steam-water environment at 343 0 C and 15.2 MPa. The design study considered a 3-m length probe. A 0.3-m length probe prototype was fabricated and tested. The optical probe consists of a 3.5-mm diameter optics train surrounded by two coaxial coolant flow channels and two coaxial insulating dead air spaces. With air flowing through the probe at 5.7 g/s, thermal analysis shows that no part of the optics train will exceed 93 0 C when a 3-m length probe is immersed in a 343 0 C environment. Computer stress analysis plus actual tests show that the probe can operate successfully with conservative safety factors. The imaging optical probe was tested five times in the design environment at the semiscale facility at the INEL. Two-phase flow regimes in the high temperature, high pressure, steam-water blowdown and reflood experiments were recorded on video tape for the first time with the imaging optical probe

  4. Makeup water system performance and impact on PWR steam generator corrosion

    International Nuclear Information System (INIS)

    Bell, M.J.; Sawocha, S.G.; Smith, L.A.

    1984-01-01

    The object of this EPRI-funded project was to assess the possible relation of pressurized water reactor (PWR) steam generator corrosion at fresh water sites to makeup water impurity ingress. Makeup water system design, operation and performance reviews were based on site visits, plant design documents, performance records and grab sample analyses. Design features were assessed in terms of their effect on makeup system performance. Attempts were made to correlate the makeup plant source water, system design characteristics, and typical makeup water qualities to steam generator corrosion observations, particularly intergranular attack (IGA). Direct correlations were not made since many variables are involved in the corrosion process and in the case of IGA, the variables have not been clearly established. However, the study did demonstrate that makeup systems can be a significant source of contaminants that are suspected to lead to both IGA and denting. Additionally, it was noted that typical makeup system performance with respect to organic removal was not good. The role of organics in steam generator damage has not been quantified and may deserve further study

  5. Forward and pressure retarded osmosis: potential solutions for global challenges in energy and water supply.

    Science.gov (United States)

    Klaysom, Chalida; Cath, Tazhi Y; Depuydt, Tom; Vankelecom, Ivo F J

    2013-08-21

    Osmotically driven membrane processes (ODMP) have gained renewed interest in recent years and they might become a potential solution for the world's most challenging problems of water and energy scarcity. Though the concept of utilizing osmotic pressure difference between high and low salinity streams across semipermeable membranes has been explored for several decades, lack of optimal membranes and draw solutions hindered competition between forward osmosis (FO) and pressure retarded osmosis (PRO) with existing water purification and power generation technologies, respectively. Driven by growing global water scarcity and by energy cost and negative environmental impacts, novel membranes and draw solutions are being developed for ODMPs, mass and heat transfer in osmotic process are becoming better understood, and new applications of ODMPs are emerging. Therefore, OMDPs might become promising green technologies to provide clean water and clean energy from abundantly available renewable resources. This review focuses primarily on new insights into osmotic membrane transport mechanisms and on novel membranes and draw solutions that are currently being developed. Furthermore, the effects of operating conditions on the overall performance of osmotic membranes will be highlighted and future perspectives will be presented.

  6. Chaotic behavior of water column oscillator simulating pressure balanced injection system in passive safety reactor

    International Nuclear Information System (INIS)

    Morimoto, Y.; Madarame, H.; Okamoto, K.

    2001-01-01

    Japan Atomic Energy Research Institute (JAERI) proposed a passive safety reactor called the System-integrated Pressurized Water Reactor (SPWR). In a loss of coolant accident, the Pressurizing Line (PL) and the Injection Line (IL) are passively opened. Vapor generated by residual heat pushes down the water level in the Reactor Vessel (RV). When the level is lower than the inlet of the PL, the vapor is ejected into the Containment Vessel (CV) through the PL. Then boronized water in the CV is injected into the RV through the IL by the static head. In an experiment using a simple apparatus, gas ejection and water injection were found to occur alternately under certain conditions. The gas ejection interval was observed to fluctuate considerably. Though stochastic noise affected the interval, the experimental results suggested that the large fluctuation was produced by an inherent character in the system. A set of piecewise linear differential equations was derived to describe the experimental result. The large fluctuation was reproduced in the analytical solution. Thus it was shown to occur even in a deterministic system without any source of stochastic noise. Though the derived equations simulated the experiment well, they had ten independent parameters governing the behavior of the solution. There appeared chaotic features and bifurcation, but the analytical model was too complicated to examine the features and mechanism of bifurcation. In this study, a new simple model is proposed which consists of a set of piecewise linear ordinary differential equations with only four independent parameters. (authors)

  7. Use of pressure manifestations following the water plasma expansion for phytomass disintegration.

    Science.gov (United States)

    Maroušek, Josef; Kwan, Jason Tai Hong

    2013-01-01

    A prototype capable of generating underwater high-voltage discharges (3.5 kV) coupled with water plasma expansion was constructed. The level of phytomass disintegration caused by transmission of the pressure shockwaves (50-60 MPa) followed by this expansion was analyzed using gas adsorption techniques. The dynamics of the external surface area and the micropore volume on multiple pretreatment stages of maize silage and sunflower seeds was approximated with robust analytical techniques. The multiple increases on the reaction surface were manifest in up to a 15% increase in cumulative methane production, which was itself manifest in the overall acceleration of the anaerobic fermentation process. Disintegration of the sunflower seeds allowed up to 45% higher oil yields using the same operating pressure.

  8. Numerical simulation of large systems: application to a pressurized water reactor

    International Nuclear Information System (INIS)

    Tallec, Michele.

    1981-10-01

    This note describes the design of a pressurized water reactor power plant simulator using a minicomputer. It contains the description of the models used to simulate the dynamic behavior of the various components of the nuclear power station (i.e. the reactor core, two steam generators, the pressurizer and the control systems associated with them); the algorithms used to integrate the resulting system of algebraic differential equations; the solution of problems associated with the use of a mini-computer; the control deck outlay designed and the variables shown on it to the user; and finally the description of tests made to validate the models used and the results obtained for various transients using plant signal is presented. These results are compared to corresponding plant signals and outputs of other, already existing models [fr

  9. Use of CPXSD for generation of effective fast multigroup libraries for pressure vessel fluence calculations

    International Nuclear Information System (INIS)

    Alpan, F. Arzu; Haghighat, Alireza

    2008-01-01

    Multigroup (i.e., broad-group) libraries play a significant role in the accuracy of transport calculations. There are several broad-group libraries available for particular applications. For example the 47-neutron (26 fast groups), 20-gamma-group BUGLE libraries are commonly used for light water reactor shielding and pressure vessel dosimetry problems. However, there is no publicly available methodology to construct group structures for a problem and objective of interest. Therefore, we have developed the Contribution and Point-wise Cross-Section Driven (CPXSD) methodology, which constructs effective fine-and broad-group structures. In this paper, we use the CPXSD methodology to construct broad-group structures for fast neutron dosimetry problems. It is demonstrated that the broad-group libraries generated from CPXSD constructed group structures, while only 14 groups (rather than 26 groups) in the fast energy range are in good agreement (similar to 1 %-2 %) with the fine-group library from which they were derived, in reaction rate calculations.

  10. Generating of low energy intensive ion streams in conditions of low pressure

    International Nuclear Information System (INIS)

    Zinoviev, D.V.; Tseluyko, A.F.; Chunadra, A.G.; Yunakov, N.N.

    2000-01-01

    In the work the method of forming of low energy ion streams near the sample surface with separating the generation area of plasma and the acceleration area of ion is offered.It allows to lower pressure in acceleration area essentially (0.01 Pa and below).The separating of the areas takes place at the expense of vacuum resistance in a plasma generating device.The dependence of plasma parameters on exterior parameters of the device is determined and the way of the further decreasing of working pressure in the modification area up to 10 -3 - 10 -4 Pa are shown

  11. Generation and maintenance of low effective pressures due to fluid flow in fractured rocks

    Science.gov (United States)

    Garagash, D.; Brantut, N.; Schubnel, A.; Bhat, H. S.

    2017-12-01

    The pore fluid pressure is expected to increase with increasing depth in the crust, primarily due to gravity forces. Because direct measurements are impossible beyond a few kilometers depths, the pore pressure gradient is often assumed to be linear (e.g., hydrostatic). However, a number of processes can severely modify the fluid pressure distribution in the crust. Here, we investigate the effect of fluid flow coupled to nonlinear permeability-effective pressure relationship. We performed a set of laboratory fluid flow experiments on thermally cracked Westerly granite at confining pressures up to 200 MPa and pore fluid pressures up to 120 MPa. Fluid flow was generated by imposing very strong pore pressure differences, up to 120 MPa, between the ends of the sample. The vertical fluid pressure distribution inside the sample was inferred by a set of 8 radial strain gauges, and an array of 10 P- and S-wave transducers. When the effective stress is kept near zero at one end of the sample and maintained high at the other end, the steady-state pore pressure profile is nonlinear. The effective stress, as inferred from the strain gauge array, remains close to zero through 2/3 of the sample, and increases sharply near the drained end of the sample. The ultrasonic data are used to build a vertical P- and S-wave velocity structure. The wave velocity profiles are consistent with a nonlinear relationship between wave velocity and effective pressure, as expected in thermally cracked granite. Taken together, our experimental data confirm the theoretical prediction that near zero effective stress can be generated through significant sections of rocks as a response to an imposed fluid flow. This has strong implications for the state of stress of the Earth's crust, especially around major continental transform faults that act as conduits for deep volatiles.

  12. In-service diagnostic systems of steam generators, pressurizers and other components of WWER type nuclear power plants

    International Nuclear Information System (INIS)

    Matal, O.

    1988-01-01

    A detailed description is presented of the systems of vibration inspections and systems of determining residual service life, implemented as in-service diagnostic systems for steam generators and pressurizers at the Dukovany nuclear power plant. Low temperature accelerometers of the KD or KS type and high temperature accelerometers CA 91 are used as vibration sensors. In the system of vibration inspection a total of 64 vibration measuring chains of Czechoslovak make and design are installed in the power plant. Systems are being built for determining residual service life which consist of 75 special chains for heat monitoring with thermocouples installed on selected assemblies of the steam generators and the pressurizers serving to monitor and evaluate heat stress. Also included in the system for determining residual service life are 16 routes for water withdrawal from steam generators. Their purpose is to make in-service determinations of places of biggest concentrations of impurities in secondary water, to determine the biggest local chemical exposure of primary collector and heat exchange tube materials and to optimize the size and place of leachate withdrawal. (Z.M.). 2 figs., 2 tabs., 15 refs

  13. Robustness of parameter-less remote real-time pressure control in water distribution systems

    CSIR Research Space (South Africa)

    Page, Philip R

    2017-06-01

    Full Text Available One way of reducing water leakage, pipe bursts and water consumption in a water distribution system (WDS) is to manage the pressure to be as low as possible. This can be done by adjusting a pressure control valve (PCV) in real-time in order to keep...

  14. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    Energy Technology Data Exchange (ETDEWEB)

    Burdin, J.R.

    1995-10-25

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  15. Acceptance Test Report for the high pressure water jet system canister cleaning fixture

    International Nuclear Information System (INIS)

    Burdin, J.R.

    1995-01-01

    This Acceptance Test confirmed the test results and recommendations, documented in WHC-SD-SNF-DTR-001, Rev. 0 Development Test Report for the High Pressure Water Jet System Nozzles, for decontaminating empty fuel canisters in KE-Basin. Optimum water pressure, water flow rate, nozzle size and overall configuration were tested

  16. Thermodynamic performance analysis and algorithm model of multi-pressure heat recovery steam generators (HRSG) based on heat exchangers layout

    International Nuclear Information System (INIS)

    Feng, Hongcui; Zhong, Wei; Wu, Yanling; Tong, Shuiguang

    2014-01-01

    Highlights: • A general model of multi-pressure HRSG based on heat exchangers layout is built. • The minimum temperature difference is introduced to replace pinch point analysis. • Effects of layout on dual pressure HRSG thermodynamic performances are analyzed. - Abstract: Changes of heat exchangers layout in heat recovery steam generator (HRSG) will modify the amount of waste heat recovered from flue gas; this brings forward a desire for the optimization of the design of HRSG. In this paper the model of multi-pressure HRSG is built, and an instance of a dual pressure HRSG under three different layouts of Taihu Boiler Co., Ltd. is discussed, with specified values of inlet temperature, mass flow rate, composition of flue gas and water/steam parameters as temperature, pressure etc., steam mass flow rate and heat efficiency of different heat exchangers layout of HRSG are analyzed. This analysis is based on the laws of thermodynamics and incorporated into the energy balance equations for the heat exchangers. In the conclusion, the results of the steam mass flow rate, heat efficiency obtained for three heat exchangers layout of HRSGs are compared. The results show that the optimization of heat exchangers layout of HRSGs has a great significance for waste heat recovery and energy conservation

  17. Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

    International Nuclear Information System (INIS)

    Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

    2000-01-01

    As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

  18. Condensation effects in a pressurizer scaled from a pressurized water reactor

    International Nuclear Information System (INIS)

    Loomis, G.G.; Shaw, R.A.

    1985-01-01

    This paper presents results from an experimental investigation of phenomena associated with pressurizer auxiliary spray during an abnormal plant transient in a commercial PWR. If normal pressurizer spray is unavailable (main coolant pumps are off) or the pressurizer power operated relief valve cannot be used during abnormal transients, pressurizer auxiliary spray can be used to reduce primary system pressure. Results from both transient integral experiments involving pressurizer auxiliary spray during tube rupture and separate effects spray experiments are presented. The experimental investigation was conducted in the Semiscale MOD-2B facility. Phenomenon of interest that occurred in the pressurizer during the pressurized auxiliary spray was desuperheating of the pressurizer steam space and quenching of metal walls followed by dropwise condensation of the pressurizer steam. The data from both the transient integral experiments and the separate effects experiments are compared to RELAP5 computer calculations and the capability of existing models in the code is discussed

  19. Low pressure arc discharge lamp apparatus with magnetic field generating means

    Science.gov (United States)

    Grossman, Mark W.; George, William A.; Maya, Jakob

    1987-01-01

    A low-pressure arc discharge apparatus having a magnetic field generating means for increasing the output of a discharge lamp is disclosed. The magnetic field generating means, which in one embodiment includes a plurality of permanent magnets, is disposed along the lamp for applying a constant transverse magnetic field over at least a portion of the positive discharge column produced in the arc discharge lamp operating at an ambient temperature greater than about 25.degree. C.

  20. Low pressure arc discharge lamp apparatus with magnetic field generating means

    Science.gov (United States)

    Grossman, M.W.; George, W.A.; Maya, J.

    1987-10-06

    A low-pressure arc discharge apparatus having a magnetic field generating means for increasing the output of a discharge lamp is disclosed. The magnetic field generating means, which in one embodiment includes a plurality of permanent magnets, is disposed along the lamp for applying a constant transverse magnetic field over at least a portion of the positive discharge column produced in the arc discharge lamp operating at an ambient temperature greater than about 25 C. 3 figs.

  1. Pore water pressures and slope stability: a joint geophysical and geotechnical analysis

    International Nuclear Information System (INIS)

    Perrone, Angela; Lapenna, Vincenzo; Vassallo, Roberto; Maio, Caterina Di

    2008-01-01

    Slope stability is influenced by many factors, among which are subsoil structure and pore water pressure distribution. This paper presents a multi-disciplinary approach for the determination of these two factors and for the construction of a reliable model of the subsoil for the slope stability analysis. The case of a clay slope located in the Southern Apennines (Italy) is presented and discussed. Geophysical imaging (2D electrical resistivity tomography—ERT), in situ geotechnical monitoring (measurements of pore pressures and horizontal displacements) and laboratory geotechnical tests (for the determination of index, hydraulic and mechanical properties of soils) have been carried out. The comparison and the integration between ERT images and direct observations of the material extracted from boreholes have allowed us to reconstruct the subsoil stratigraphy with continuity. Thus, a reliable 2D model of the subsoil has been obtained, with well-defined boundaries on which it has been possible to apply appropriate hydraulic conditions. This geotechnical model has been used for studying the pore water pressure distribution and for analysing how the hydraulic boundary conditions—among which rain events—influence the slope stability. Our findings demonstrate the powerful skill of the ERT, if integrated with borehole data, to generate an accurate subsoil model. It is also evident that geophysical imaging can be a source of ambiguity and misjudgement if interpreted without a comparison with geotechnical data

  2. Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

    International Nuclear Information System (INIS)

    Additon, S.L.; Fontana, M.H.; Carter, J.C.

    1990-01-01

    This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

  3. Countercurrent Air-Water Flow in a Scale-Down Model of a Pressurizer Surge Line

    Directory of Open Access Journals (Sweden)

    Takashi Futatsugi

    2012-01-01

    Full Text Available Steam generated in a reactor core and water condensed in a pressurizer form a countercurrent flow in a surge line between a hot leg and the pressurizer during reflux cooling. Characteristics of countercurrent flow limitation (CCFL in a 1/10-scale model of the surge line were measured using air and water at atmospheric pressure and room temperature. The experimental results show that CCFL takes place at three different locations, that is, at the upper junction, in the surge line, and at the lower junction, and its characteristics are governed by the most dominating flow limitation among the three. Effects of inclination angle and elbows of the surge line on CCFL characteristics were also investigated experimentally. The effects of inclination angle on CCFL depend on the flow direction, that is, the effect is large for the nearly horizontal flow and small for the vertical flow at the upper junction. The presence of elbows increases the flow limitation in the surge line, whereas the flow limitations at the upper and lower junctions do not depend on the presence of elbows.

  4. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  5. A Receding Horizon Controller for the Steam Generator Water Level

    International Nuclear Information System (INIS)

    Na, Man Gyun; Lee, Yoon Joon

    2003-01-01

    In this work, the receding horizon control method was used to control the water level of nuclear steam generators and applied to two linear models and also a nonlinear model of steam generators. A receding horizon control method is to solve an optimization problem for finite future steps at current time and to implement the first optimal control input as the current control input. The procedure is then repeated at each subsequent instant. The dynamics of steam generators is very different according to power levels. The receding horizon controller is designed by using a reduced linear steam generator model fixed over a certain power range and applied to a Westinghouse-type (U-tube recirculating type) nuclear steam generator. The proposed controller designed at a fixed power level shows good performance for any other power level within this power range. The steam generator shows actually nonlinear characteristics. Therefore, the proposed algorithm is implemented for a nonlinear model of the nuclear steam generator to verify its real performance and also shows good responses

  6. High pressure, low pressure and hot water heating systems in hospitals. Hochdruck-, Niederdruck- und Warmwasserheizungsanlagen im Krankenhaus

    Energy Technology Data Exchange (ETDEWEB)

    Riedle, K [H. Riedle GmbH, Wiesbaden (Germany)

    1994-07-01

    In hospital nowadays the limitation of the use of steam boilers and their direct supply network to the possible minimum is aimed at when the heating system is exchanged or retrofitted. Independent of the fact whether high pressure or low pressure steam or hot water is used the optimum water treatment should be carried out with a minimum of chemical substances. Here hydroquinone, neutralizing amines, carbohydrazide, sodium sulphite and tannins can be used. The dimensioning of hot water heating circuits is shown with examples. (BWI)

  7. Effects of ambient temperature and water vapor on chamber pressure and oxygen level during low atmospheric pressure stunning of poultry.

    Science.gov (United States)

    Holloway, Paul H; Pritchard, David G

    2017-08-01

    The characteristics of the vacuum used in a low atmospheric pressure stunning system to stun (render unconscious) poultry prior to slaughter are described. A vacuum chamber is pumped by a wet screw compressor. The vacuum pressure is reduced from ambient atmospheric pressure to an absolute vacuum pressure of ∼250 Torr (∼33 kPa) in ∼67 sec with the vacuum gate valve fully open. At ∼250 Torr, the sliding gate valve is partially closed to reduce effective pumping speed, resulting in a slower rate of decreasing pressure. Ambient temperature affects air density and water vapor pressure and thereby oxygen levels and the time at the minimum total pressure of ∼160 Torr (∼21 kPa) is varied from ∼120 to ∼220 sec to ensure an effective stun within the 280 seconds of each cycle. The reduction in total pressure results in a gradual reduction of oxygen partial pressure that was measured by a solid-state electrochemical oxygen sensor. The reduced oxygen pressure leads to hypoxia, which is recognized as a humane method of stunning poultry. The system maintains an oxygen concentration of air always reduces the oxygen concentrations to a value lower than in dry air. The partial pressure of water and oxygen were found to depend on the pump down parameters due to the formation of fog in the chamber and desorption of water from the birds and the walls of the vacuum chamber. © The Author 2017. Published by Oxford University Press on behalf of Poultry Science Association.

  8. Chemically reactive species in liquids generated by atmospheric-pressure plasmas and their roles in plasma medicine

    Energy Technology Data Exchange (ETDEWEB)

    Hamaguchi, Satoshi [Center for Atomic and Molecular Technologies, Graduate School of Engineering, Osaka University, 2-1 Yamadaoka, Suita, Osaka 565-0871 (Japan)

    2013-07-11

    Plasmas whose gas temperatures are close to room temperature may be generated in ambient air or a gas at atmospheric pressure with the use of low-frequency high voltage or low-power radio-frequency (RF) or microwave power applied to electrodes. Such plasmas can serve as a powerful source of free radicals and/or chemically reactive species that arise from atoms and molecules of the ambient gas. Recently use of such plasmas for medical purposes has attracted much attention as they can be implemented in possible medical devices that can cause blood coagulation, heal wounds, facilitate angiogenesis, sterilize surgical devices as well as living tissues without harming healthy cells, and selectively inactivate cancer cells. Especially of interest among reactive species generated by atmospheric-pressure plasmas (APP) are reactive oxygen species (ROS) and reactive nitrogen species (RNS) that are generated in liquid phase. Since most living tissues and cells are immersed in liquids (such as blood or culture media), reactive species generated by APPs in the gas phase are transported to the liquid phase and possibly converted to different types of reactive species therein before causing some influence on the tissues or cells. In this study, the rate equations are solved to evaluate concentrations of various reactive species in pure water that are originated by plasma reactions in atmosphere and possible effects of such species (including ROS/RNS) on living tissues and cells are discussed.

  9. Chemically reactive species in liquids generated by atmospheric-pressure plasmas and their roles in plasma medicine

    International Nuclear Information System (INIS)

    Hamaguchi, Satoshi

    2013-01-01

    Plasmas whose gas temperatures are close to room temperature may be generated in ambient air or a gas at atmospheric pressure with the use of low-frequency high voltage or low-power radio-frequency (RF) or microwave power applied to electrodes. Such plasmas can serve as a powerful source of free radicals and/or chemically reactive species that arise from atoms and molecules of the ambient gas. Recently use of such plasmas for medical purposes has attracted much attention as they can be implemented in possible medical devices that can cause blood coagulation, heal wounds, facilitate angiogenesis, sterilize surgical devices as well as living tissues without harming healthy cells, and selectively inactivate cancer cells. Especially of interest among reactive species generated by atmospheric-pressure plasmas (APP) are reactive oxygen species (ROS) and reactive nitrogen species (RNS) that are generated in liquid phase. Since most living tissues and cells are immersed in liquids (such as blood or culture media), reactive species generated by APPs in the gas phase are transported to the liquid phase and possibly converted to different types of reactive species therein before causing some influence on the tissues or cells. In this study, the rate equations are solved to evaluate concentrations of various reactive species in pure water that are originated by plasma reactions in atmosphere and possible effects of such species (including ROS/RNS) on living tissues and cells are discussed

  10. A moist air condensing device for sustainable energy production and water generation

    International Nuclear Information System (INIS)

    Ming, Tingzhen; Gong, Tingrui; Richter, Renaud K. de; Wu, Yongjia; Liu, Wei

    2017-01-01

    Highlights: • A novel device based upon a SCPP system is proposed for electricity production and water generation. • The collector is replaced by black tubes around the chimney. • The overall performance of SCPP for energy production and water generation was analyzed. • The system total energy efficiency of a SCPP with a height of 3000 m can be nearly 7%. - Abstract: A solar chimney power plant (SCPP) is not only a solar thermal application system to achieve output power, but also a device extracting freshwater from the humid air. In this article, we proposed a SCPP with collector being replaced by black tubes around the chimney to warm water and air. The overall performance of SCPP was analyzed by using a one-dimensional compressible fluid transfer model to calculate the system characteristic parameters, such as chimney inlet air velocity, the condensation level, amount of condensed water, output power, and efficiency. It was found that increasing the chimney inlet air temperature is an efficient way to increase chimney inlet air velocity and wind turbine output power. The operating conditions, such as air temperature and air relative humidity, have significant influence on the condensation level. For water generation, chimney height is the most decisive factor, the mass flow rate of condensed water decreases with increasing wind turbine pressure drop. To achieve the optimum peak output power by wind turbine, we should set the pressure drop factor as about 0.7. In addition, increasing chimney height is also an efficient way to improve the SCPP efficiency. Under ideal conditions, the system total efficiency of a SCPP with a height of 3000 m can be up to nearly 7%.

  11. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  12. An innovative technique for estimating water saturation from capillary pressure in clastic reservoirs

    Science.gov (United States)

    Adeoti, Lukumon; Ayolabi, Elijah Adebowale; James, Logan

    2017-11-01

    A major drawback of old resistivity tools is the poor vertical resolution and estimation of hydrocarbon when applying water saturation (Sw) from historical resistivity method. In this study, we have provided an alternative method called saturation height function to estimate hydrocarbon in some clastic reservoirs in the Niger Delta. The saturation height function was derived from pseudo capillary pressure curves generated using modern wells with complete log data. Our method was based on the determination of rock type from log derived porosity-permeability relationship, supported by volume of shale for its classification into different zones. Leverette-J functions were derived for each rock type. Our results show good correlation between Sw from resistivity based method and Sw from pseudo capillary pressure curves in wells with modern log data. The resistivity based model overestimates Sw in some wells while Sw from the pseudo capillary pressure curves validates and predicts more accurate Sw. In addition, the result of Sw from pseudo capillary pressure curves replaces that of resistivity based model in a well where the resistivity equipment failed. The plot of hydrocarbon pore volume (HCPV) from J-function against HCPV from Archie shows that wells with high HCPV have high sand qualities and vice versa. This was further used to predict the geometry of stratigraphic units. The model presented here freshly addresses the gap in the estimation of Sw and is applicable to reservoirs of similar rock type in other frontier basins worldwide.

  13. Correction of Pressure Drop in Steam and Water System in Performance Test of Boiler

    Science.gov (United States)

    Liu, Jinglong; Zhao, Xianqiao; Hou, Fanjun; Wu, Xiaowu; Wang, Feng; Hu, Zhihong; Yang, Xinsen

    2018-01-01

    Steam and water pressure drop is one of the most important characteristics in the boiler performance test. As the measuring points are not in the guaranteed position and the test condition fluctuation exsits, the pressure drop test of steam and water system has the deviation of measuring point position and the deviation of test running parameter. In order to get accurate pressure drop of steam and water system, the corresponding correction should be carried out. This paper introduces the correction method of steam and water pressure drop in boiler performance test.

  14. A technical learning on the Pressurized Water Nuclear Power Plants using animation

    International Nuclear Information System (INIS)

    Ito, Hajime; Tomohara, Yasutaka; Kubo, Setsuo; Ninomiya, Toshiaki

    2002-01-01

    The pressurized water nuclear power generation plants tends to reduce construction of its new plant from viewpoints of recent stabilization in power demand/supply balance, development of new siting points, and so on. And, together with reducing any opportunity to experience at site, generation alternation to younger engineers without such experiences is progressing. In order to carry out technical tradition with high quality , as it is important to understand experiences of troubles and so on as valuable inheritance to apply them to actual use, it can be thought, in doubt, to be one of solving measures to prepare some learning tools applying the newest technology. The Kansai Electric Co., Ltd. Developed a CAD software using animation and 3D pictures using a personal computer which is edited some processes of technical transition on nuclear energy as a reference on a shape of CD ROM as an object from initial period of nuclear power station to present APWR. (G.K.)

  15. Parametric studies for the nuclear design of high-conversion pressurized water reactors

    International Nuclear Information System (INIS)

    Axmann, J.; Oldekop, W.

    1987-01-01

    Undermoderated high-conversion pressurized water reactors with steel canning tubes offer the possibility of high burnup together with a comparatively low consumption of fissionable material; however, they require a relatively large inventory of fissionable material. The effects of different fuel compositions upon the specific consumption of fissionable material are investigated for a fixed burnup and moderator-to-fuel volume ratios varying between 0.5 and 2.0. Moreover, the required inventory of fissionable material is determined and the influence on the costs of electric power generation is shown. Further investigations deal with the neutron-physical effects of decreasing fuel rod diameters and the influence of differing steel additives. It appears that the parasitic neutron absorption by alloying constituents depends on the moderation level in a non-uniform manner and that the contribution of the fissionable material to the electric power generation costs is rather independent of the moderator-to-fuel volume ratio. (orig.) [de

  16. Characteristics and economy of the European reactor of pressurized water (EPR)

    International Nuclear Information System (INIS)

    Ortiz V, J.; Ramirez S, J.R.; Palacios H, J.C.

    2005-01-01

    The high current costs of the fossil fuels, have propitiated that the industries of electric power generation in the world reconsider the nuclear option as medium of generation. In Europe, the more recently contracted nuclear power plant is that of Olkiluoto-III in Finland that waits it enters in operation at the end of 2009. The reactor that will be installed in this power plant will be a prototype of pressurized water reactor of the companies AREVA and EDF. In this work they are described the reactor EPR and the major components of the nuclear power plant as well as the main characteristics of safety and the flexibility of the operation of the EPR. The supposed costs reported in different sources of information are also described and calculated with information provided by the manufacturer company. (Author)

  17. Evaluation of heat exchange performance for primary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Nakagawa, Shigeaki

    2006-01-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30 MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the primary pressurized water cooler (PPWC) and the intermediate heat exchanger (IHX). The heat exchangers in the primary cooling system are required the heat exchange performance to remove reactor generated heat 30 MW under the condition of reactor coolant outlet temperature 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance data of the rise-to-power-up test and the in-service operation for the PPWC in the main cooling system was evaluated. Moreover, the evaluated values were compared with the design values, and it is confirmed that PPWC has the required heat exchange performance in the design. (author)

  18. Evaluation of heat exchange performance for secondary pressurized water cooler in HTTR

    International Nuclear Information System (INIS)

    Tochio, Daisuke; Watanabe, Syuji; Saikusa, Akio; Oyama, Sunao; Nemoto, Takahiro; Hamamoto, Shinpei; Shinohara, Masanori; Isozaki, Minoru; Nakagawa, Shigeaki

    2006-02-01

    In High Temperature Engineering Test Reactor (HTTR), the rated thermal power of 30MW, the generated heat at reactor core is finally dissipated at the air-cooler by way of the heat exchangers of the primary cooling system, such as the intermediate heat exchanger (IHX) and the secondary pressurized water cooler (SPWC). The heat exchangers in the main cooling system are required the heat exchange performance to remove the reactor-generated-heat of 30MW under the condition of reactor coolant outlet temperature of 850degC/950degC. Therefore, the heat exchanges are required to satisfy the design criteria of heat exchange performance. In this report, heat exchange performance of the SPWC in the main cooling system was evaluated with the rise-to-power-up test and the in-service operation data. Moreover, evaluated value is compared with designed one, it is confirmed that the SPWC has required heat exchange performance. (author)

  19. Atmospheric-pressure electric discharge as an instrument of chemical activation of water solutions

    Science.gov (United States)

    Rybkin, V. V.; Shutov, D. A.

    2017-11-01

    Results of experimental studies and numerical simulations of physicochemical characteristics of plasmas generated in different types of atmospheric-pressure discharges (pulsed streamer corona, gliding electric arc, dielectric barrier discharge, glow-discharge electrolysis, diaphragmatic discharge, and dc glow discharge) used to initiate various chemical processes in water solutions are analyzed. Typical reactor designs are considered. Data on the power supply characteristics, plasma electron parameters, gas temperatures, and densities of active particles in different types of discharges excited in different gases and their dependences on the external parameters of discharges are presented. The chemical composition of active particles formed in water is described. Possible mechanisms of production and loss of plasma particles are discussed.

  20. Acoustic emission technique for leak detection in an end shield of a pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Kalyanasundaram, P.; Jayakumar, T.; Raj, B.

    1989-01-01

    This paper discusses the successful application of the Acoustic Emission Technique (AET) for detection and location of leak paths present on the inaccessible side of an end shield of a Pressurized Heavy Water Reactor (PHWR). The methodology was based on the fact that air and water leak AE signals have different characteristic features. Baseline data was generated from a sound end-shield of a PHWR for characterizing the background noise. A mock up end-shield system with saw cut leak paths was used to verify the validity of the methodology. It was found that air leak signals under pressurisation (as low as 3 psi) could be detected by frequency domain analysis. Signals due to air leaks from various locations of a defective end-shield were acquired and analysed. It was possible to detect and locate leak paths. Presence of detected leak paths were further confirmed by alternate test. (orig.)

  1. A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

    International Nuclear Information System (INIS)

    Mandelli, Diego; Prescott, Steven; Smith, Curtis; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua; Kinoshita, Robert

    2015-01-01

    In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code called NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins

  2. Effect of condenser water in-leakage on steam generator water chemistry

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.

    1978-01-01

    Corrosive environments may be generated within steam genrators from condenser cooling water in-leakage. Theoretical as well as experimental evaluation of the aggressiveness of such environments is being carried out for the condenser-cooling waters used at CANDU-PHW nuclear power stations. Calculations have shown that highly concentrated chloride solutions - acidic in the case of sea-water in-leakage, and alkaline in the rest of the cases considered - would be produced within the steam generator. Experiments in a model boiler showed that sea-water in-leakage caused rapid corrosion of carbon steel components when only AVT (all volatile treatment) was used for water chemistry control. Use of a non-volatile reagent, as in the congruent phosphate treatment, avoided the rapid corrosion of carbon steel. On the basis of our studies, congruent phosphate treatment during sea water in-leakage appears desirable. (author)

  3. Energy and exergy analysis of electricity generation from natural gas pressure reducing stations

    International Nuclear Information System (INIS)

    Neseli, Mehmet Alparslan; Ozgener, Onder; Ozgener, Leyla

    2015-01-01

    Highlights: • Forecasting the recoverable energy from natural gas pressure reduction stations. • Electricity generation through pressure reduction stations via turboexpanders. • A thermodynamics analysis of PRS. - Abstract: Electricity generation or power recovery through pressure reduction stations (PRS) for general use has not been realized in Izmir. The main objective of the present study was to do a case study for calculating electricity to be recovered in one natural gas pressure reduction stations in Izmir. It is the first forecasting study to obtain energy from natural gas pressure-reducing stations in Izmir. Energy can be obtained from natural gas PRS with turbo-expanders instead of using throttle valves or regulators from the PRS. The exergy performance of PRS with TE is evaluated in this study. Exergetic efficiencies of the system and components are determined to assess their individual performances. Based upon pressure change and volumetric flow rate, it can be obtained by recovering average estimated installed capacity and annual energy 494.24 kW, 4113.03 MW h, respectively. In terms of estimated installed capacity power and annual energy, the highest level is 764.88 kW, approximately 6365.34 MW h, in Aliaga PRS. Also it can be seen that CO 2 emission factor average value is 295.45 kg/MW h

  4. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert W.; Ng, Kim Choon; Sarathy, Mani

    2017-01-01

    decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed

  5. Soluble epoxide hydrolase in the generation and maintenance of high blood pressure in spontaneously hypertensive rats

    NARCIS (Netherlands)

    Koeners, Maarten P.; Wesseling, Sebastiaan; Ulu, Arzu; Lopez Sepulveda, Rocio; Morisseau, Christophe; Braam, Branko; Hammock, Bruce D.; Joles, Jaap A.

    Koeners MP, Wesseling S, Ulu A, Sepulveda RL, Morisseau C, Braam B, Hammock BD, Joles JA. Soluble epoxide hydrolase in the generation and maintenance of high blood pressure in spontaneously hypertensive rats. Am J Physiol Endocrinol Metab 300: E691-E698, 2011. First published January 25, 2011; doi:

  6. On the combined effect of moisture diffusion and cyclic pore pressure generation in asphalt concrete

    NARCIS (Netherlands)

    Varveri, A.; Scarpas, A.; Collop, A.; Erkens, S.M.J.G.

    2014-01-01

    In this paper, a new moisture conditioning protocol which attempts to distinguish the contributions of long- and short-term moisture damage, i.e. moisture diffusion and cyclic pore pressure generation, in asphalt mixtures is presented. The capability of the proposed protocol to rank various asphalt

  7. Manufacture of the 300 MW steam generator and pressure stabilizer for Qinshan Nuclear Power Station

    International Nuclear Information System (INIS)

    Qian Yi; Miao Deming.

    1989-01-01

    A brief description of the manufacturing process of the steam generator and pressure stabilizer for 300 MWe Qinshan Nuclear Power Station in Shanghai Boiler Works is presented, with special emphasis on fabrication facilities, test procedures and technological evaluations during the manufaturing process-imcluding deep driling of tubesheets, welding of tubes to tube-sheets and tube rolling tests

  8. Microvascular pressure responses of second-generation rats chronically exposed to 2 g centrifugation

    Science.gov (United States)

    Richardson, D. R.; Knapp, C. F.

    1977-01-01

    Preliminary results are presented for a study aimed at a quantitative comparison of microvascular dynamics in second-generation rats reared in a 2-g force field produced by centrifugation with similar data from animals reared in a centrifuge that produced only a 1-g force. It is shown that the pressure distribution in the mesenteric microvasculature of the second generation of rats reared in a 2-g environment, as well as the animals' blood pressure response to epinephrine, are significantly different compared to their 1-g counterparts. In particular, 1-g and 2-g chronic centrifugation enhances the arterial blood pressure, and the 2-g force field attenuates the pressor effects of norepinephrine.

  9. Performance of materials in the component cooling water systems of pressurized water reactors

    International Nuclear Information System (INIS)

    Lee, B.S.

    1993-01-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

  10. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  11. Dynamic study of steam generation from low-grade waste heat in a zeolite–water adsorption heat pump

    International Nuclear Information System (INIS)

    Xue, Bing; Meng, Xiangrui; Wei, Xinli; Nakaso, Koichi; Fukai, Jun

    2015-01-01

    A novel zeolite–water adsorption heat pump system based on a direct-contact heat exchange method to generate steam from low-grade waste gas and water has been proposed and examined experimentally. Superheated steam (200 °C, 0.1 MPa) is generated from hot water (70–80 °C) and dry air (100–130 °C). A dynamic model for steam generation process is developed to describe local mass and heat transfer. This model features a three-phase calculation and a moving water–gas interface. The calculations are carried out in the zeolite–water and zeolite–gas regions. Model outputs are compared with experimental results for validation. The thermal response inside the reactor and mass of steam generated is well predicted. Numerical results show that preheat process with low-temperature steam is an effective method to achieve local equilibrium quickly, thus generation process is enhanced by prolonging the time and increasing mass of the generated steam. Besides, high-pressure steam generation up to 0.5 MPa is possible from the validated dynamic model. Future work could be emphasized on enhancing high-pressure steam generation with preheat process or mass recovery operation

  12. Application of Genetic Algorithm methodologies in fuel bundle burnup optimization of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Jayalal, M.L.; Ramachandran, Suja; Rathakrishnan, S.; Satya Murty, S.A.V.; Sai Baba, M.

    2015-01-01

    Highlights: • We study and compare Genetic Algorithms (GA) in the fuel bundle burnup optimization of an Indian Pressurized Heavy Water Reactor (PHWR) of 220 MWe. • Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are considered. • For the selected problem, Multi Objective GA performs better than Penalty Functions based GA. • In the present study, Multi Objective GA outperforms Penalty Functions based GA in convergence speed and better diversity in solutions. - Abstract: The work carried out as a part of application and comparison of GA techniques in nuclear reactor environment is presented in the study. The nuclear fuel management optimization problem selected for the study aims at arriving appropriate reference discharge burnup values for the two burnup zones of 220 MWe Pressurized Heavy Water Reactor (PHWR) core. Two Genetic Algorithm methodologies namely, Penalty Functions based GA and Multi Objective GA are applied in this study. The study reveals, for the selected problem of PHWR fuel bundle burnup optimization, Multi Objective GA is more suitable than Penalty Functions based GA in the two aspects considered: by way of producing diverse feasible solutions and the convergence speed being better, i.e. it is capable of generating more number of feasible solutions, from earlier generations. It is observed that for the selected problem, the Multi Objective GA is 25.0% faster than Penalty Functions based GA with respect to CPU time, for generating 80% of the population with feasible solutions. When average computational time of fixed generations are considered, Penalty Functions based GA is 44.5% faster than Multi Objective GA. In the overall performance, the convergence speed of Multi Objective GA surpasses the computational time advantage of Penalty Functions based GA. The ability of Multi Objective GA in producing more diverse feasible solutions is a desired feature of the problem selected, that helps the

  13. Fuzzy power control algorithm for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Y.J.; Lee, B.W.

    1994-01-01

    A fuzzy power control algorithm is presented for automatic reactor power control in a pressurized water reactor (PWR). Automatic power shape control is complicated by the use of control rods with a conventional proportional-integral-differential controller because it is highly coupled with reactivity compensation. Thus, manual shape controls are usually employed even for the limited capability needed for load-following operations including frequency control. In an attempt to achieve automatic power shape control without any design modifications to the core, a fuzzy power control algorithm is proposed. For the fuzzy control, the rule base is formulated based on a multiple-input multiple-output system. The minimum operation rule and the center of area method are implemented for the development of the fuzzy algorithm. The fuzzy power control algorithm has been applied to Yonggwang Nuclear Unit 3. The simulation results show that the fuzzy control can be adapted as a practical control strategy for automatic reactor power control of PWRs during the load-following operations

  14. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  15. Current state of research on pressurized water reactor safety

    International Nuclear Information System (INIS)

    Couturier, Jean; Schwarz, Michel; Roubaud, Sebastien; Lavarenne, Caroline; Mattei, Jean-Marie; Rigollet, Laurence; Scotti, Oona; Clement, Christophe; Lancieri, Maria; Gelis, Celine; Jacquemain, Didier; Bentaib, Ahmed; Nahas, Georges; Tarallo, Francois; Guilhem, Gilbert; Cattiaux, Gerard; Durville, Benoit; Mun, Christian; Delaval, Christine; Sollier, Thierry; Stelmaszyk, Jean-Marc; Jeffroy, Francois; Dechy, Nicolas; Chanton, Olivier; Tasset, Daniel; Pichancourt, Isabelle; Barre, Francois; Bruna, Gianni; Evrard, Jean-Michel; Gonzalez, Richard; Loiseau, Olivier; Queniart, Daniel; Vola, Didier; Goue, Georges; Lefevre, Odile

    2018-03-01

    For more than 40 years, IPSN then IRSN has conducted research and development on nuclear safety, specifically concerning pressurized water reactors, which are the reactor type used in France. This publication reports on the progress of this research and development in each area of study - loss-of-coolant accidents, core melt accidents, fires and external hazards, component aging, etc. -, the remaining uncertainties and, in some cases, new measures that should be developed to consolidate the safety of today's reactors and also those of tomorrow. A chapter of this report is also devoted to research into human and organizational factors, and the human and social sciences more generally. All of the work is reviewed in the light of the safety issues raised by feedback from major accidents such as Chernobyl and Fukushima Daiichi, as well as the issues raised by assessments conducted, for example, as part of the ten-year reviews of safety at French nuclear reactors. Finally, through the subjects it discusses, this report illustrates the many partnerships and exchanges forged by IRSN with public, industrial and academic bodies both within Europe and internationally

  16. Remerschen nuclear power station with BBR pressurized water reactor

    International Nuclear Information System (INIS)

    Hoffmann, J.P.

    1975-01-01

    On the basis of many decades of successful cooperation in the electricity supply sector with the German RWE utility, the Grand Duchy of Luxemburg and RWE jointly founded Societe Luxembourgeoise d'Energie Nucleaire S.A. (SENU) in 1974 in which each of the partners holds a fifty percent interest. SENU is responsible for planning, building and operating this nuclear power station. Following an international invitation for bids on the delivery and turnkey construction of a nuclear power station, the consortium of the German companies of Brown, Boveri and Cie. AG (BBC), Babcock - Brown Boveri Reaktor GmbH (BBR) and Hochtief AG (HT) received a letter of intent for the purchase of a 1,300 MW nuclear power station equipped with a pressurized water reactor. The 1,300 MW station of Remerschen will be largely identical with the Muelheim-Kaerlich plant under construction by the same consortium near Coblence on the River Rhine since early 1975. According to present scheduling, the Remerschen nuclear power station could start operation in 1981. (orig.) [de

  17. Numerical study of optimal equilibrium cycles for pressurized water reactors

    International Nuclear Information System (INIS)

    Mahlers, Y.P.

    2003-01-01

    An algorithm based on simulated annealing and successive linear programming is applied to solve equilibrium cycle optimization problems for pressurized water reactors. In these problems, the core reload scheme is represented by discrete variables, while the cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are treated as continuous variables. The enrichments are considered to be distinct in all feed fuel assemblies. The number of batches and their sizes are not fixed and also determined by the algorithm. An important feature of the algorithm is that all the parameters are determined by the solution of one optimization problem including both discrete and continuous variables. To search for the best reload scheme, simulated annealing is used. The optimum cycle length as well as uranium enrichment and loading of burnable poison in each feed fuel assembly are determined for each reload pattern examined using successive linear programming. Numerical results of equilibrium cycle optimization for various values of the effective price of electricity and fuel reprocessing cost are studied

  18. Software to study the control strategy of pressurized water reactor

    International Nuclear Information System (INIS)

    Oliveira, Jose Ricardo de

    2002-01-01

    The computational program, result of this work, is a tool developed for the study of the control of Pressurized Water Reactors (PWR) constituted by only one coolant loop. The implementation of a user friendly interface for input/output data, makes the program also suitable for training and teaching applications. As design premise, it was considered enough just the modeling of the primary circuit, using as interface with the secondary circuit, a simplified differential equation of the temperature associated with the secondary power. All the incorporated dynamic equations to the model were developed using basic laws of conservation, boundary conditions and hypotheses appropriated to the control study. To arrive to the final model, core thermal and hydraulic characteristics and design data were obtained from of the available bibliography and adapted for a conceptual peculiar design of a small PWR. The whole program and all input/output interfaces were developed using the software Matlab, version 5.L Sub-routines of numeric integration based on the Runge-Kutta 4 method were applied, to solve the set of ordinary differential equations. (author)

  19. Evolution of general design requirements for french pressurized water reactors

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Rollinger, F.

    1988-10-01

    The design of French pressurized water reactors is based first on deterministic principles, using the well-known defense in depth concept. This safety approach, basically reflected current American practice at that time, which consisted notably in designing engineered safeguard systems capable of limiting the consequences of accidents assumed to be credible despite the preventive measures taken. Further reflections have led to complete this approach, resulting in modifications to regulatory practice, mainly related to better practical assimilation of the problems arising during plant unit operation and reactor control after an accident and to the determination to enhance the overall consistency of the safety approach. As regards system redundancy, it should be noted that common cause failures can result in the total loss of a redundant system. System redundancy aspects will be dealt with in Chapter 2. As regards study of design basis accidents, attention was focused on the human intervention stage following automatic activation of protection and safeguard systems. This resulted, for all plant units, in the revision of operating procedures, accompanied by examination of the means required for their implementation. These subjects will be discussed in Chapter 3. Finally, as regards equipment classification, the range of equipment subjected to particular requirements, formerly limited to design basis safety classified equipment, was enlarged to include important for safety equipment. This subject will be dealt with in Chapter 5

  20. Design description of the European pressurized water reactor

    International Nuclear Information System (INIS)

    Leverenz, R.

    1999-01-01

    The EPR (the European Pressurized Water Reactor) is an evolutionary PWR developed by Nuclear Power International and its parent companies, Framatome and Siemens, in co-operation with Electricite de France and German Utilities. NPI can rely on the huge experience gained by its parent companies; they have constructed more than 100 nuclear power plants throughout the world. The total installed capacity exceeds 100,000 MW - about 25% of the total world-wide figure. Following the conceptual design phase of the so-called Common Product conducted by NPI, Framatome and Siemens, from 1989 through 1991, Electricite de France (EDF) and several major German utilities decided to merge their own development programmes, - the N4 Plus and REP 2000 projects on the French side and the further development of the KONVOI technology on the German side, - with the NPI project. From that time on, the NPI project became one single common development line for both countries. In parallel, EDF and the German utilities decided to establish, together with other European utilities, specifications that would represent common utility views on the design and performance of future nuclear power plants. These are documented in the European Utility Requirements (EURs). The basic design has been completed in 1997, and in 1998 a design optimization is being carried out with the goal to even increase the economic competitiveness of nuclear power. This paper provides a brief design description of the EPR. (author)