WorldWideScience

Sample records for water pressure

  1. 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

  2. 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)

  3. 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.

  4. 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…

  5. 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

  6. 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.

  7. 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)

  8. 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

  9. 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.

  10. 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

  11. 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

  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. 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)

  17. 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.

  18. 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).

  19. 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.…

  20. 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

  1. 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.

  2. 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.

  3. 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.

  4. 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

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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.

  10. 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.

  11. 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.)

  12. 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.

  13. 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

  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. 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)

  16. 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.

  17. 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....

  18. 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.

  19. 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

  20. 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

  1. 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.

  2. 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)

  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. 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.)

  6. 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

  7. 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

  8. 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.)

  9. 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)

  10. 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

  11. 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

  12. 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

  13. 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.

  14. 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.

  15. 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…

  16. 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)

  17. 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...

  18. 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

  19. 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....

  20. 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

  1. 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).

  2. 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

  3. 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

  4. 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

  5. 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.

  6. 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.

  7. 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.

  8. 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

  9. 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.

  10. 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...

  11. 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.)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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...

  17. 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…

  18. 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.

  19. 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)

  20. 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...

  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. 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.

  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. 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

  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. 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

  10. 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.)

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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

  17. 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

  18. 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

  19. 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...

  20. 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.

  1. 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

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. 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

  7. 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)

  8. 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

  9. 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

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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.

  15. 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.

  16. 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.

  17. 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

  18. 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

  19. 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)

  20. 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

  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. 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

  4. 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

  5. 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)

  6. 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)

  7. 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.)

  8. 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

  9. 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)

  10. 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)

  11. 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

  12. 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.)

  13. 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)

  14. 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)

  15. 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)

  16. 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

  17. 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

  18. 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

  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. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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)

  9. 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.

  10. 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.)

  11. 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

  12. 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...

  13. 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)

  14. 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.

  15. 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

  16. 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)

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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.

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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)

  7. 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

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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)

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

    Science.gov (United States)

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

    2003-01-01

    Advances in polymer chemistry have led to the development of monomers and initiation agents that enable propagating free-radical polymerization fronts to exist. These fronts are driven by the exothermicity of the polymerization reaction and the transport of heat from the polymerized product to the reactant monomer/solvent/initiator solution. The thermal energy transported to the reactant solution causes the initiator to decompose, yielding free radicals, which start the free radical polymerization process as discussed in recent reviews. The use of polymerization processes based on propagating fronts has numerous applications. Perhaps the most important of these is that it enables rapid curing of polymers without external heating since the polymerization process itself provides the high temperatures necessary to initiate and sustain polymerization. This process also enables more uniform curing of arbitrarily thick samples since it does not rely on heat transfer from an external source, which will necessarily cause the temperature history of the sample to vary with distance from the surface according to a diffusion-like process. Frontal polymerization also enables filling and sealing of structures having cavities of arbitrary shape without having to externally heat the structure. Water at atmospheric pressure is most convenient solvent to employ and the most important for practical applications (because of the cost and environmental issues associated with DMSO and other solvents). Nevertheless, to our knowledge, steady, self-propagating polymerization fronts have not been reported in water at atmospheric pressure. Currently, polymerization fronts require a high boiling point solvent (either water at high pressures or an alternative solvent such as dimethyl sulfoxide (DMSO) (boiling point 189 C at atmospheric pressure.) Early work on frontal polymerization, employed pressures up to 5000 atm in order to avoid boiling of the monomer/solvent/initiator solution. High

  14. 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.

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. 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.)

  20. 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

  1. 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

  2. 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

  3. 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)

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. 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)

  10. 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

  11. 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)

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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.

  3. 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.

  4. 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

  5. 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.

  6. 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)

  7. 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

  8. 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

  9. 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...

  10. 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

  11. 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

  12. 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

  13. 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

  14. 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)

  15. 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.

  16. 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

  17. 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.

  18. 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

  19. 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

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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

  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. 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.

  10. 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

  11. 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)

  12. 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

  13. 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)

  14. Contributions to the energetical role of advanced pressurized water reactors

    International Nuclear Information System (INIS)

    Seifritz, W.

    1984-06-01

    Three articles written by the author in the past two years and all concerned with energy production of the Advanced Pressurized Water Reactor (APWR) are collated to form this report. The standpoint is made that the APWR using mixed oxide-fuel (about 8% Pusub(fiss) and Usub(depl)) would entail similar capital costs as a PWR, but with conversion rates of 0.85< C<0.95. Given this assumption the analysis shows that the high specific plutonium inventories (being proportional to the conversion ratio in the interested range) result in a strong damping effect on the growth rate, the effective uranium utilization, and on the total nuclear power level. Over one century an APWR strategy is 'far from equilibrium' and to describe this dynamic situation using static APWR data is not appropriate. If nuclear fission energy is to play a substantial and not just a marginal role in providing energy for the future (in the region of 5 - 8 TWsub(el) in 50 years time, corresponding to a share of 20 - 25% of the total world demand), clearly the fast breeder strategy offers the better chance of achieving the goal. (Auth.)

  15. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.

    2016-09-06

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  16. High protein flexibility and reduced hydration water dynamics are key pressure adaptive strategies in prokaryotes

    KAUST Repository

    Martinez, N.; Michoud, Gregoire; Cario, A.; Ollivier, J.; Franzetti, B.; Jebbar, M.; Oger, P.; Peters, J.

    2016-01-01

    Water and protein dynamics on a nanometer scale were measured by quasi-elastic neutron scattering in the piezophile archaeon Thermococcus barophilus and the closely related pressure-sensitive Thermococcus kodakarensis, at 0.1 and 40 MPa. We show that cells of the pressure sensitive organism exhibit higher intrinsic stability. Both the hydration water dynamics and the fast protein and lipid dynamics are reduced under pressure. In contrast, the proteome of T. barophilus is more pressure sensitive than that of T. kodakarensis. The diffusion coefficient of hydration water is reduced, while the fast protein and lipid dynamics are slightly enhanced with increasing pressure. These findings show that the coupling between hydration water and cellular constituents might not be simply a master-slave relationship. We propose that the high flexibility of the T. barophilus proteome associated with reduced hydration water may be the keys to the molecular adaptation of the cells to high hydrostatic pressure.

  17. Thermo-hydraulic behavior of saturated steam-water mixture in pressure vessel during injection of cold water

    International Nuclear Information System (INIS)

    Aya, Izuo; Kobayashi, Michiyuki; Inasaka, Fujio; Nariai, Hideki.

    1983-01-01

    The thermo-hydraulic behavior of saturated steam water mixture in a pressure vessel during injection of cold water was experimentally investigated with the Facility for Mixing Effect of Emergency Core Cooling Water. The dimensions of the pressure vessel used in the experiments were 284mm ID and 1,971mm height. 11 experiments were conducted without blowdown in order to comprehend the basic process excluding the effect of blowdown at injection of cold water. The initial pressure and water level, the injection flow rate and the size of injection nozzle were chosen as experimental parameters. Temperatures and void fractions at 6 elevations as well as pressure in the pressure vessel were measured, and new data especially on the pressure undershoot just after the initation of water injection and the vertical distribution of temperature and void fraction were gotten. The transients of pressure, average temperature and void fraction were caluculated using single-volume analysis code BLODAC-1V which is based on thermal equilibrium and so-called bubble gradient model. Some input parameters included in the analysis code were evaluated through the comparison of analysis with experimental data. Moreover, the observed pressure undershoot which is evaluated to be induced by a time lag of vapourization in water due to thermal nonequilibrium, was also discussed with the aid of another simple analysis model. (author)

  18. Spreading pressures of water and n-propanol on polymer surfaces

    NARCIS (Netherlands)

    Busscher, H.J.; Kip, Gerhardus A.M.; van Silfhout, Arend; Arends, J.

    1986-01-01

    Spreading pressures of water and n-propanol on polytetrafluoroethylene (PTFE), polystyrene (PS), polymethylmethacrylate (PMMA), polycarbonate (PC), and glass are determined from ellipsometrically measured adsorption isotherms by graphical integration, yielding for water 9, 37, 26, 33, and 141

  19. Modeling and diagnostic techniques applicable to the analysis of pressure noise in pressurized water reactors and pressure-sensing systems

    International Nuclear Information System (INIS)

    Mullens, J.A.; Thie, J.A.

    1984-01-01

    Pressure noise data from a PWR are interpreted by means of a computer-implemented model. The model's parameters, namely hydraulic impedances and noise sources, are either calculated or deduced from fits to data. Its accuracy is encouraging and raises the possibility of diagnostic assistance for nuclear plant monitoring. A number of specific applications of pressure noise in the primary system of a PWR and in a pressure sensing system are suggested

  20. On the pressure field of nonlinear standing water waves

    Science.gov (United States)

    Schwartz, L. W.

    1980-01-01

    The pressure field produced by two dimensional nonlinear time and space periodic standing waves was calculated as a series expansion in the wave height. The high order series was summed by the use of Pade approximants. Calculations included the pressure variation at great depth, which was considered to be a likely cause of microseismic activity, and the pressure distribution on a vertical barrier or breakwater.

  1. Living PSA issues in France on pressurized water reactors

    International Nuclear Information System (INIS)

    Dewailly, J.; Deriot, S.; Dubreuil Chambardel, A.; Francois, P.; Magne, L.

    1993-09-01

    Two Probabilistic Safety Assessments (PSAs) carried out in France on 900 and 1300 MWe Pressurized Water Reactor units ended in 1990. These PSAs determined the core damage frequency for all plant operating conditions ranging from cold shutdown for refuelling to full power operation. Since 1990, these PSAs have been used increasingly as tools for applications such as accident precursor analysis, risk-based Technical Specifications, and maintenance optimization. In turn, these applications are used to enhance the initial PSAs. The notion of a ''living'' PSA which can be used and updated is slowly taking form. The accident precursor analysis consists in applying PSA event trees to obtain quick information on the potential consequences of a precursor event and on the corresponding probabilities of occurrence. A feedback on PSAs is provided by comparing them with actual operating incidents. The computation of the allowed outage time during power operation, based on the computerized models of Probabilistic Safety Assessments, requires adjustments: calculation of hourly risk of core damage under different reactor conditions without equipment unavailabilities. The proposed method also turns out to be an aid in determining the safe shutdown condition and procedure. Furthermore, when introducing a sufficient level of detail, PSA reliability models make it possible to compute contributions and to perform sensitivity studies in order to highlight those components for which a maintenance effort should be made. From the experience acquired up to now, there was felt to be a strong need to create guidelines for using PSAs that would simplify their implementation by the experts in charge of determining Technical Specifications, of maintenance programs, etc. who are not generally specialists in PSAs. For this purpose, it is necessary to improve the intelligibility of the models made in order for them to be used and to offer user's guides adapted to each application. Documents

  2. 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.

  3. Effect of water pressure on absorbency of hydroentangled greige cotton nonwoven fabrics

    Science.gov (United States)

    A studied has been conducted to determine the effect of water pressure in a commercial-grade Fleissner MiniJet hydroentanglement system on the absorbency of greige (non-bleached) cotton lint-based nonwoven fabric. The study has shown that a water pressure of 125 Bar or higher on only two high-pressu...

  4. Behavior of pressure rise and condensation caused by water evaporation under vacuum at high temperature

    International Nuclear Information System (INIS)

    Takase, Kazuyuki; Kunugi, Tomoaki; Yamazaki, Seiichiro; Fujii, Sadao

    1998-01-01

    Pressure rise and condensation characteristics during the ingress-of-coolant event (ICE) in fusion reactors were investigated using the preliminary ICE apparatus with a vacuum vessel (VV), an additional tank (AT) and an isolation valve (IV). A surface of the AT was cooled by water at RT. The high temperature and pressure water was injected into the VV which was heated up to 250degC and pressure and temperature transients in the VV were measured. The pressure increased rapidly with an injection time of the water because of the water evaporation. After the IV was opened and the VV was connected with the AT, the pressure in the VV decreased suddenly. From a series of the experiments, it was confirmed that control factors on the pressure rise were the flushing evaporation and boiling heat transfer in the VV, and then, condensation of the vapor after was effective to the depressurization in the VV. (author)

  5. 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

  6. 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.

  7. 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)

  8. Clean-up system for pool water in pressure suppression chamber and operation method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Hirabayashi, Kentaro; Kinoshita, Shoichiro

    1996-09-17

    Pool water in a pressure suppression chamber of a BWR type reactor is sucked by a pump of an after-heat removing system. The pool water pressurized here is sent to the pressure suppression chamber by way of a heat exchanger and a test line backwarding pipeline to stir the pool water in the pressure suppression chamber. Further, the pool water pressurized by the pump is sent to the pressure suppression chamber by way of a filtration desalting device and an exit pipe to purify the pool water. Upon cleaning of pipelines before the start of a periodical test, the pool water sucked by the pump is sent to the filtration desalting device and recovered to the pressure suppression chamber. This can reduce the amount of impurities carried to the suppression chamber. After the cleaning of the pipelines, pool water is passed through the test line backwarding pipeline, so that the pool water can be stirred at the same time. (I.N.)

  9. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

  10. Evaporation of tungsten in vacuum at low hydrogen and water vapor pressures

    International Nuclear Information System (INIS)

    Andrievskij, R.A.; Galkin, E.A.; Khromonozhkin, V.V.

    1981-01-01

    The results of experimental investigations of tungsten evaporation rates in the temperature range 1650-2500 K, partial hydrogen and water vapours pressures 1x10 -5 -10 Pa are presented. Experi-- mental plant, equipment employed and radiometric technique of tungsten evaporation study are described. The dependences of evaporation rate and probabilities of tungsten oxidation by residual vacuum water vapours and dependences of tungsten evaporation rate on partial hydrogen and water vapours pressures are determined [ru

  11. 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

  12. Effect of Water Cut on Pressure Drop of Oil (D130) -Water Flow in 4″Horizontal Pipe

    Science.gov (United States)

    Basha, Mehaboob; Shaahid, S. M.; Al-Hems, Luai M.

    2018-03-01

    The oil-water flow in pipes is a challenging subject that is rich in physics and practical applications. It is often encountered in many oil and chemical industries. The pressure gradient of two phase flow is still subject of immense research. The present study reports pressure measurements of oil (D130)-water flow in a horizontal 4″ diameter stainless steel pipe at different flow conditions. Experiments were carried out for different water cuts (WC); 0-100%. Inlet oil-water flow rates were varied from 4000 to 8000 barrels-per-day in steps of 2000. It has been found that the frictional pressure drop decreases for WC = 0 - 40 %. With further increase in WC, friction pressure drop increases, this could be due to phase inversion.

  13. Pressure-driven ballistic Kelvin's water dropper for energy harvesting

    NARCIS (Netherlands)

    Xie, Yanbo; de Boer, Hans L.; van den Berg, Albert; Sprenkels, A.J.; Eijkel, Jan C.T.

    2014-01-01

    In this paper, we introduce a microfluidic-based self-excited energy conversion system inspired by Kelvin's water dropper but driven by inertia instead of gravity. Two micro water jets are produced by forcing water through two micropores by overpressure. The jets break up into microdroplets which

  14. Impact of radiation embrittlement on integrity of pressure vessel supports for two PWR [pressurized-water-reactor] plants

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Pennell, W.E.; Robinson, G.C.; Nanstad, R.K.

    1988-01-01

    Recent pressure-vessel surveillance data from the High Flux Isotope Reactor (HFIR) indicate an embrittlement fluence-rate effect that is applicable to the evaluation of the integrity of light-water reactor (LWR) pressure vessel supports. A preliminary evaluation using the HFIR data indicated increases in the nil ductility transition temperature at 32 effective full-power years (EFPY) of 100 to 130/degree/C for pressurized-water-reactor (PWR) vessel supports located in the cavity at midheight of the core. This result indicated a potential problem with regard to life expectancy. However, an accurate assessment required a detailed, specific-plant, fracture-mechanics analysis. After a survey and cursory evaluation of all LWR plants, two PWR plants that appeared to have a potential problem were selected. Results of the analyses indicate minimum critical flaw sizes small enough to be of concern before 32 EFPY. 24 refs., 16 figs., 7 tabs

  15. Pressure suppression containment system for boiling water reactor

    Science.gov (United States)

    Gluntz, Douglas M.; Nesbitt, Loyd B.

    1997-01-01

    A system for suppressing the pressure inside the containment of a BWR following a postulated accident. A piping subsystem is provided which features a main process pipe that communicates the wetwell airspace to a connection point downstream of the guard charcoal bed in an offgas system and upstream of the main bank of delay charcoal beds which give extensive holdup to offgases. The main process pipe is fitted with both inboard and outboard containment isolation valves. Also incorporated in the main process pipe is a low-differential-pressure rupture disk which prevents any gas outflow in this piping whatsoever until or unless rupture occurs by virtue of pressure inside this main process pipe on the wetwell airspace side of the disk exceeding the design opening (rupture) pressure differential. The charcoal holds up the radioactive species in the noncondensable gas from the wetwell plenum by adsorption, allowing time for radioactive decay before the gas is vented to the environs.

  16. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  17. Reducing water losses via intelligent pressure management; Reduzierung von Wasserverlusten durch intelligentes Druckmanagement

    Energy Technology Data Exchange (ETDEWEB)

    Oppinger, Peter [VAG-Armaturen GmbH, Mannheim (Germany). Marketing

    2008-03-15

    Leaks in water pipes and leaking municipal water-transmission and piping systems, particularly in developing and threshold countries account for water-losses of up to 50% of the water supplied by the waterworks. This article examines three different solutions for effective pressure management on the basis of an intelligent control system, by means of which water-losses can be reduced to a stable and economically rational level. (orig.)

  18. Methods and means for reducing pressure in systems for fire fighting and water spraying in mines

    Energy Technology Data Exchange (ETDEWEB)

    Kozlyuk, A I; Grin' , G V; Yushchenko, Yu N

    1986-01-01

    Valves are evaluated used in water systems for fire fighting and dust suppression in underground black coal mines in the USSR. Specifications of the KR-2, the KR-3 and the R-86 pressure-reducing valves used in deep mines are analyzed. The valves are characterized by low reliability, low capacity and low pressure reducing range. Therefore groups (parallel arrangement) of pressure-reducing valves are used. Using valve groups increases equipment cost. The pressure-reducing systems should consist of no more than 2 valves. The VNIIGD Institute developed the RKGD pressure-reducing valve with the following specifications: inlet pressure 6.87 MPa, outlet pressure from 0.98 to 2.45 MPa, water discharge 100 m/sup 3//h). The RKGD valves are characterized by high reliability but extremely high weight. Therefore, the VNIIGD Institute developed a modified version of pressure-reducing valve, called the PRK (with maximum inlet pressure of 5 MPa, outlet pressure ranging from 0.5 to 1.5 MPa, water discharge 80 m/sup 3//h and weighing 5 kg). Design of the PRK pressure-reducing valve is shown.

  19. The nuclear physical method for high pressure steam manifold water level gauging and its error

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    A new method, which is non-contact on measured water level, for measuring high pressure steam manifold water level with nuclear detection technique is introduced. This method overcomes the inherent drawback of previous water level gauges based on other principles. This method can realize full range real time monitoring on the continuous water level of high pressure steam manifold from the start to full load of boiler, and the actual value of water level can be obtained. The measuring errors were analysed on site. Errors from practical operation in Tianjin Junliangcheng Power Plant and in laboratory are also presented

  20. Theory of the Maxwell pressure tensor and the tension in a water bridge.

    Science.gov (United States)

    Widom, A; Swain, J; Silverberg, J; Sivasubramanian, S; Srivastava, Y N

    2009-07-01

    A water bridge refers to an experimental "flexible cable" made up of pure de-ionized water, which can hang across two supports maintained with a sufficiently large voltage difference. The resulting electric fields within the de-ionized water flexible cable maintain a tension that sustains the water against the downward force of gravity. A detailed calculation of the water bridge tension will be provided in terms of the Maxwell pressure tensor in a dielectric fluid medium. General properties of the dielectric liquid pressure tensor are discussed along with unusual features of dielectric fluid Bernoulli flows in an electric field. The "frictionless" Bernoulli flow is closely analogous to that of a superfluid.

  1. Response of the water status of soybean to changes in soil water potentials controlled by the water pressure in microporous tubes

    Science.gov (United States)

    Steinberg, S. L.; Henninger, D. L.

    1997-01-01

    Water transport through a microporous tube-soil-plant system was investigated by measuring the response of soil and plant water status to step change reductions in the water pressure within the tubes. Soybeans were germinated and grown in a porous ceramic 'soil' at a porous tube water pressure of -0.5 kpa for 28 d. During this time, the soil matric potential was nearly in equilibrium with tube water pressure. Water pressure in the porous tubes was then reduced to either -1.0, -1.5 or -2.0 kPa. Sap flow rates, leaf conductance and soil, root and leaf water potentials were measured before and after this change. A reduction in porous tube water pressure from -0.5 to -1.0 or -1.5 kPa did not result in any significant change in soil or plant water status. A reduction in porous tube water pressure to -2.0 kPa resulted in significant reductions in sap flow, leaf conductance, and soil, root and leaf water potentials. Hydraulic conductance, calculated as the transpiration rate/delta psi between two points in the water transport pathway, was used to analyse water transport through the tube-soil-plant continuum. At porous tube water pressures of -0.5 to-1.5 kPa soil moisture was readily available and hydraulic conductance of the plant limited water transport. At -2.0 kPa, hydraulic conductance of the bulk soil was the dominant factor in water movement.

  2. 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,

  3. Gray water recycle: Effect of pretreatment technologies on low pressure reverse osmosis treatment

    Science.gov (United States)

    Gray water can be a valuable source of water when properly treated to reduce the risks associated with chemical and microbial contamination to acceptable levels for the intended reuse application. In this study, the treatment of gray water using low pressure reverse osmosis (RO) filtration after pre...

  4. Check experiment of the high pressure water washing technology used to the decommissioning of reactor

    International Nuclear Information System (INIS)

    Han Jianping; Hou Yongming; Fu Yunshan

    2004-01-01

    High pressure water washing technology has been widely applied in the field of the decommissioning of nuclear facilities, and it is used to wash the sump for craft conveyance, the craft workshop, the hermetic sump, and some other nuclear equipment as well. The authors have got a set of technical data correlated with high pressure water washing technology by comparing the situations between the test before and after the washing work. At the same time, authors also improve the technique on some special cases, which made the high pressure water washing technology more perfect in the field of the decommissioning of nuclear facilities. (authors)

  5. Increase in gas output by active modification of the water pressure regime

    Energy Technology Data Exchange (ETDEWEB)

    Zakirov, S N; Gordon, V Y; Kondrat, R M; Kravtsov, N A; Somov, B Y

    1981-01-01

    Based on gas-hydrodynamic calculations made on a planar model formation, two variants of formation working are examined. In the first variant, the modern ideology of working gas fields with a water pressure regime are simulated. In the second variant, working of the formation is modeled according to the suggested ideology of active modification of the water-pressure regime by operating the flooded gas wells. The calculations made indicate the efficiency of active modification of the water pressure regime from the viewpoint of controlling the fund of E wells, and most important, maximizing the final coefficient of gas bed output.

  6. Light Water Reactor-Pressure Vessel Surveillance project computer system

    International Nuclear Information System (INIS)

    Merriman, S.H.

    1980-10-01

    A dedicated process control computer has been implemented for regulating the metallurgical Pressure Vessel Wall Benchmark Facility (PSF) at the Oak Ridge Research Reactor. The purpose of the PSF is to provide reliable standards and methods by which to judge the radiation damage to reactor pressure vessel specimens. Benchmark data gathered from the PSF will be used to improve and standardize procedures for assessing the remaining safe operating lifetime of aging reactors. The computer system controls the pressure vessel specimen environment in the presence of gamma heating so that in-vessel conditions are simulated. Instrumented irradiation capsules, in which the specimens are housed, contain temperature sensors and electrical heaters. The computer system regulates the amount of power delivered to the electrical heaters based on the temperature distribution within the capsules. Time-temperature profiles are recorded along with reactor conditions for later correlation with specimen metallurgical changes

  7. Dew-point measurements at high water vapour pressure

    Science.gov (United States)

    Lomperski, S.; Dreier, J.

    1996-05-01

    A dew-point meter capable of measuring humidity at high vapour pressure and high temperature has been constructed and tested. Humidity measurements in pure steam were made over the temperature range 100 - 1500957-0233/7/5/003/img1C and a vapour pressure range of 1 - 4 bar. The dew-point meter performance was assessed by comparing measurements with a pressure transmitter and agreement between the two was within 0957-0233/7/5/003/img2% relative humidity. Humidity measurements in steam - air mixtures were also made and the dew-point meter readings were compared to those of a zirconia oxygen sensor. For these tests the dew-point meter readings were generally within 0957-0233/7/5/003/img2% relative humidity of the oxygen sensor measurements.

  8. Extended fuel cycle operation for pressurized water reactor plants

    International Nuclear Information System (INIS)

    Silvestri, G.J. Jr.

    1978-01-01

    A nuclear steam turbine power plant system having an arrangement therein for extended fuel cycle operation is described. The power plant includes a turbine connected at its inlet to a source of motive fluid having a predetermined pressure associated therewith. The turbine has also connected thereto an extraction conduit which extracts steam from a predetermined location therein for use in an associated apparatus. A bypass conduit is provided between a point upstream of the inlet and the extraction conduit. A flow control device is provided within the bypass conduit and opens when the pressure of the motive steam supply drops beneath the predetermined pressure as a result of reactivity loss within the nuclear reactor. Opening of the bypass conduit provides flow to the associated apparatus and at the same time provides an increased flow orifice to maintain fluid flow rate at a predetermined level

  9. Topic 1. Steels for light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Brynda, J.; Kepka, M.; Barackova, L.; Vacek, M.; Havel, S.; Cukr, B.; Protiva, K.; Petrman, I.; Tvrdy, M.; Hyspecka, L.; Mazanec, K.; Kupca, L.; Brezina, M.

    1980-01-01

    Part 1 of the Proceedings consists of papers on the criteria for the selection and comparison of the properties of steel for pressure vessels and on the metallurgy of the said steels, the selection of suitable material for internal tubing systems, the manufacture of high-alloy steels for WWER components, the mechanical and metallurgical properties of steel 22K for WWER 440 pressure components, and of steel 10MnNi2Mo for the WWER primary coolant circuit, and the metallographic assessment of steel 0Kh18N10T. (J.P.)

  10. Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

    Science.gov (United States)

    Li, Hong-Li; Wang, Yong-Yang; Zhang, Qian; Wang, Pu; Zhang, Ming-Xiang; Yu, Fei-Hai

    2015-01-01

    Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum) with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot) and two levels of water depth (30 cm and 70 cm). Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

  11. Vegetative Propagule Pressure and Water Depth Affect Biomass and Evenness of Submerged Macrophyte Communities.

    Directory of Open Access Journals (Sweden)

    Hong-Li Li

    Full Text Available Vegetative propagule pressure may affect the establishment and structure of aquatic plant communities that are commonly dominated by plants capable of clonal growth. We experimentally constructed aquatic communities consisting of four submerged macrophytes (Hydrilla verticillata, Ceratophyllum demersum, Elodea nuttallii and Myriophyllum spicatum with three levels of vegetative propagule pressure (4, 8 and 16 shoot fragments for communities in each pot and two levels of water depth (30 cm and 70 cm. Increasing vegetative propagule pressure and decreasing water level significantly increased the growth of the submerged macrophyte communities, suggesting that propagule pressure and water depth should be considered when utilizing vegetative propagules to re-establish submerged macrophyte communities in degraded aquatic ecosystems. However, increasing vegetative propagule pressure and decreasing water level significantly decreased evenness of the submerged macrophyte communities because they markedly increased the dominance of H. verticillata and E. nuttallii, but had little impact on that of C. demersum and M. spicatum. Thus, effects of vegetative propagule pressure and water depth are species-specific and increasing vegetative propagule pressure under lower water level can facilitate the establishment success of submerged macrophyte communities.

  12. Water dynamics and population pressure in the Nepalese Himalayas.

    Science.gov (United States)

    Schreier, H; Shah, P B

    1996-10-01

    The authors investigate the impact of water shortages, especially water for irrigation, on development in Nepal. "The problems associated with hydropower development will be illustrated by using the Kulekhani watershed project as a case study." The possible future effects on food supplies and health are discussed. excerpt

  13. Low-pressure water-cooled inductively coupled plasma torch

    Science.gov (United States)

    Seliskar, Carl J.; Warner, David K.

    1988-12-27

    An inductively coupled plasma torch is provided which comprises an inner tube, including a sample injection port to which the sample to be tested is supplied and comprising an enlarged central portion in which the plasma flame is confined; an outer tube surrounding the inner tube and containing water therein for cooling the inner tube, the outer tube including a water inlet port to which water is supplied and a water outlet port spaced from the water inlet port and from which water is removed after flowing through the outer tube; and an r.f. induction coil for inducing the plasma in the gas passing into the tube through the sample injection port. The sample injection port comprises a capillary tube including a reduced diameter orifice, projecting into the lower end of the inner tube. The water inlet is located at the lower end of the outer tube and the r.f. heating coil is disposed around the outer tube above and adjacent to the water inlet.

  14. Dielectric Relaxation of Bound Water versus Soil Matric Pressure

    NARCIS (Netherlands)

    Hilhorst, M.A.; Dirksen, C.; Kampers, F.W.H.; Feddes, R.A.

    2001-01-01

    The electrical permittivity of soil is a function of the water content, which facilitates water content measurements. The permittivity of soil is also a function of the frequency of the applied electric field. This frequency dependence can be described by the relationship between the dielectric

  15. Effects of water compressibility on the pressure fluctuation prediction in pump turbine

    International Nuclear Information System (INIS)

    Yin, J L; Wang, D Z; Wang, L Q; Wu, Y L; Wei, X Z

    2012-01-01

    The compressible effect of water is a key factor in transient flows. However, it is always neglected in the unsteady simulations for hydraulic machinery. In light of this, the governing equation of the flow is deduced to combine the compressibility of water, and then simulations with compressible and incompressible considerations to the typical unsteady flow phenomenon (Rotor stator interaction) in a pump turbine model are carried out and compared with each other. The results show that water compressibility has great effects on the magnitude and frequency of pressure fluctuation. As the operating condition concerned, the compressibility of water will induce larger pressure fluctuation, which agrees better with measured data. Moreover, the lower frequency component of the pressure signal can only be captured with the combination of water compressibility. It can be concluded that water compressibility is a fatal factor, which cannot be neglected in the unsteady simulations for pump turbines.

  16. 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

  17. Study on the pressure self-adaptive water-tight junction box in underwater vehicle

    Directory of Open Access Journals (Sweden)

    Haocai Huang

    2012-09-01

    Full Text Available Underwater vehicles play a very important role in underwater engineering. Water-tight junction box (WJB is one of the key components in underwater vehicle. This paper puts forward a pressure self-adaptive water-tight junction box (PSAWJB which improves the reliability of the WJB significantly by solving the sealing and pressure problems in conventional WJB design. By redundancy design method, the pressure self-adaptive equalizer (PSAE is designed in such a way that it consists of a piston pressure-adaptive compensator (PPAC and a titanium film pressure-adaptive compensator (TFPAC. According to hydro-mechanical simulations, the operating volume of the PSAE is more than or equal to 11.6 % of the volume of WJB liquid system. Furthermore, the required operating volume of the PSAE also increases as the gas content of oil, hydrostatic pressure or temperature difference increases. The reliability of the PSAWJB is proved by hyperbaric chamber tests.

  18. More recent developments for the ultrasonic testing of light water reactor pressure vessels

    International Nuclear Information System (INIS)

    Seiger, H.; Engl, G.

    1976-01-01

    The development of an ultrasonic testing method for the inspection from the outside of the areas close to the cladding of the spherical fields of holes of light water reactor pressure vessels is described

  19. LINEAR KERNEL SUPPORT VECTOR MACHINES FOR MODELING PORE-WATER PRESSURE RESPONSES

    Directory of Open Access Journals (Sweden)

    KHAMARUZAMAN W. YUSOF

    2017-08-01

    Full Text Available Pore-water pressure responses are vital in many aspects of slope management, design and monitoring. Its measurement however, is difficult, expensive and time consuming. Studies on its predictions are lacking. Support vector machines with linear kernel was used here to predict the responses of pore-water pressure to rainfall. Pore-water pressure response data was collected from slope instrumentation program. Support vector machine meta-parameter calibration and model development was carried out using grid search and k-fold cross validation. The mean square error for the model on scaled test data is 0.0015 and the coefficient of determination is 0.9321. Although pore-water pressure response to rainfall is a complex nonlinear process, the use of linear kernel support vector machine can be employed where high accuracy can be sacrificed for computational ease and time.

  20. The Oxidation Rate of SiC in High Pressure Water Vapor Environments

    Science.gov (United States)

    Opila, Elizabeth J.; Robinson, R. Craig

    1999-01-01

    CVD SiC and sintered alpha-SiC samples were exposed at 1316 C in a high pressure burner rig at total pressures of 5.7, 15, and 25 atm for times up to 100h. Variations in sample emittance for the first nine hours of exposure were used to determine the thickness of the silica scale as a function of time. After accounting for volatility of silica in water vapor, the parabolic rate constants for Sic in water vapor pressures of 0.7, 1.8 and 3.1 atm were determined. The dependence of the parabolic rate constant on the water vapor pressure yielded a power law exponent of one. Silica growth on Sic is therefore limited by transport of molecular water vapor through the silica scale.

  1. 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.

  2. Capital cost: pressurized water reactor plant. Commerical electric power cost studies

    International Nuclear Information System (INIS)

    1977-06-01

    The investment cost study for the 1139-MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume includes in addition to the foreword and summary, the plant description and the detailed cost estimate

  3. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description.

  4. Capital cost: pressurized water reactor plant. Commercial electric power cost studies

    International Nuclear Information System (INIS)

    1977-06-01

    The investment cost study for the 1139 MW(e) pressurized water reactor (PWR) central station power plant consists of two volumes. This volume contains the drawings, equipment list and site description

  5. Exploration of Impinging Water Spray Heat Transfer at System Pressures Near the Triple Point

    Science.gov (United States)

    Golliher, Eric L.; Yao, Shi-Chune

    2013-01-01

    The heat transfer of a water spray impinging upon a surface in a very low pressure environment is of interest to cooling of space vehicles during launch and re-entry, and to industrial processes where flash evaporation occurs. At very low pressure, the process occurs near the triple point of water, and there exists a transient multiphase transport problem of ice, water and water vapor. At the impingement location, there are three heat transfer mechanisms: evaporation, freezing and sublimation. A preliminary heat transfer model was developed to explore the interaction of these mechanisms at the surface and within the spray.

  6. Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

    International Nuclear Information System (INIS)

    Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

    2003-01-01

    The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

  7. Influence of pressure on acoustic and rheologic parameters in water solutions of laury sodium sulfate

    Science.gov (United States)

    Khamidov, B. T.; Lezhnev, N. B.

    1995-10-01

    Ultrasonic velocity and density in water solutions of lauril sodium sulphate at frequency 36 MHz, within the range of pressures from 0.1 to 105 MPa at temperature T equals 293 K were measured. According to data of ultrasonic velocity and density under high pressures there was calculated adiabatic compressibility in objects studied from pressure. It was found out that the region of critical concentration of micelle formation has been shifted to the zone of much more low concentrations.

  8. Troubled waters: Growing climate and population pressures in ...

    International Development Research Centre (IDRC) Digital Library (Canada)

    2014-04-09

    Apr 9, 2014 ... Across India, urban growth is pushing water demands and pollution on adjacent ... Neither climate models nor rainfall data show any significant changes ... “In India, we have strong civil society organizations,” says Srinivasan.

  9. Comparison of ASME pressure–temperature limits on the fracture probability for a pressurized water reactor pressure vessel

    International Nuclear Information System (INIS)

    Chou, Hsoung-Wei; Huang, Chin-Cheng

    2017-01-01

    Highlights: • P-T limits based on ASME K_I_a curve, K_I_C curve and RI method are presented. • Probabilistic and deterministic methods are used to evaluate P-T limits on RPV. • The feasibility of substituting P-T curves with more operational is demonstrated. • Warm-prestressing effect is critical in determining the fracture probability. - Abstract: The ASME Code Section XI-Appendix G defines the normal reactor startup (heat-up) and shut-down (cool-down) operation limits according to the fracture toughness requirement of reactor pressure vessel (RPV) materials. This paper investigates the effects of different pressure-temperature limit operations on structural integrity of a Taiwan domestic pressurized water reactor (PWR) pressure vessel. Three kinds of pressure-temperature limits based on different fracture toughness requirements – the K_I_a fracture toughness curve of ASME Section XI-Appendix G before 1998 editions, the K_I_C fracture toughness curve of ASME Section XI-Appendix G after 2001 editions, and the risk-informed revision method supplemented in ASME Section XI-Appendix G after 2013 editions, respectively, are established as the loading conditions. A series of probabilistic fracture mechanics analyses for the RPV are conducted employing ORNL’s FAVOR code considering various radiation embrittlement levels under these pressure-temperature limit conditions. It is found that the pressure-temperature operation limits which provide more operational flexibility may lead to higher fracture risks to the RPV. The cladding-induced shallow surface breaking flaws are the most critical and dominate the fracture probability of the RPV under pressure-temperature limit transients. Present study provides a risk-informed reference for the operation safety and regulation viewpoint of PWRs in Taiwan.

  10. Development of a quasi-adiabatic calorimeter for the determination of the water vapor pressure curve.

    Science.gov (United States)

    Mokdad, S; Georgin, E; Hermier, Y; Sparasci, F; Himbert, M

    2012-07-01

    Progress in the knowledge of the water saturation curve is required to improve the accuracy of the calibrations in humidity. In order to achieve this objective, the LNE-CETIAT and the LNE-CNAM have jointly built a facility dedicated to the measurement of the saturation vapor pressure and temperature of pure water. The principle is based on a static measurement of the pressure and the temperature of pure water in a closed, temperature-controlled thermostat, conceived like a quasi-adiabatic calorimeter. A copper cell containing pure water is placed inside a temperature-controlled copper shield, which is mounted in a vacuum-tight stainless steel vessel immersed in a thermostated bath. The temperature of the cell is measured with capsule-type standard platinum resistance thermometers, calibrated with uncertainties below the millikelvin. The vapor pressure is measured by calibrated pressure sensors connected to the cell through a pressure tube whose temperature is monitored at several points. The pressure gauges are installed in a thermostatic apparatus ensuring high stability of the pressure measurement and avoiding any condensation in the tubes. Thanks to the employment of several technical solutions, the thermal contribution to the overall uncertainty budget is reduced, and the remaining major part is mainly due to pressure measurements. This paper presents a full description of this facility and the preliminary results obtained for its characterization.

  11. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  12. Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1977-01-01

    A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

  13. Pressure Heads and Simulated Water Uptake Patterns for a Severely Stressed Bean Crop

    NARCIS (Netherlands)

    Durigon, A.; Santos, dos M.A.; Lier, van Q.D.; Metselaar, K.

    2012-01-01

    In modeling, actual crop transpiration as a function of soil hydraulic conditions is usually estimated from a water content or pressure head dependent reduction function. We compared the performance of the empirical pressure head based reduction function of Feddes (FRF) and a more physically based

  14. Photolytic removal of DBPs by medium pressure UV in swimming pool water

    OpenAIRE

    Hansen, Kamilla Marie Speht; Zortea, R.; Piketty, A.; Rodriguez Vega, S.; Andersen, Henrik Rasmus

    2013-01-01

    Medium pressure UV is used for controlling the concentration of combined chlorine (chloramines) in many public swimming pools. Little is known about the fate of other disinfection by-products (DBPs) in UV treatment. Photolysis by medium pressure UV treatment was investigated for 12 DBPs reported to be found in swimming pool water: chloroform, bromodichloromethane, dibromochloromethane, bromoform, dichloroacetonitrile, bromochloroacetonitrile, dibromoacetronitrile, trichloroacetonitrile, trich...

  15. 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

  16. Standard technical specifications for Westinghouse pressurized water reactors (revision issued Fall 1981). Technical report

    International Nuclear Information System (INIS)

    Virgilio, M.J.

    1981-11-01

    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 Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public

  17. Noise analysis method for monitoring the moderator temperature coefficient of pressurized water reactors: Neural network calibration

    International Nuclear Information System (INIS)

    Thomas, J.R. Jr.; Adams, J.T.

    1994-01-01

    A neural network was trained with data for the frequency response function between in-core neutron noise and core-exit thermocouple noise in a pressurized water reactor, with the moderator temperature coefficient (MTC) as target. The trained network was subsequently used to predict the MTC at other points in the same fuel cycle. Results support use of the method for operating pressurized water reactors provided noise data can be accumulated for several fuel cycles to provide a training base

  18. Application of linearized model to the stability analysis of the pressurized water reactor

    International Nuclear Information System (INIS)

    Li Haipeng; Huang Xiaojin; Zhang Liangju

    2008-01-01

    A Linear Time-Invariant model of the Pressurized Water Reactor is formulated through the linearization of the nonlinear model. The model simulation results show that the linearized model agrees well with the nonlinear model under small perturbation. Based upon the Lyapunov's First Method, the linearized model is applied to the stability analysis of the Pressurized Water Reactor. The calculation results show that the methodology of linearization to stability analysis is conveniently feasible. (authors)

  19. Variation of Pore Water Pressure in Tailing Sand under Dynamic Loading

    Directory of Open Access Journals (Sweden)

    Jia-xu Jin

    2018-01-01

    Full Text Available Intense vibration affects the pore water pressure in a tailing dam, with the tendency to induce dam liquefaction. In this study, experiments were performed wherein model tailing dams were completely liquefied by sustained horizontal dynamic loading to determine the effects of the vibration frequency, vibration amplitude, and tailing density on the pore water pressure. The results revealed four stages in the increase of the tailing pore water pressure under dynamic loading, namely, a slow increase, a rapid increase, inducement of structural failure, and inducement of complete liquefaction. A lower frequency and smaller amplitude of the vibration were found to increase the time required to achieve a given pore water pressure in dense tailings. Under the effect of these three factors—vibration frequency and amplitude and tailing density—the tailing liquefaction time varied nonlinearly with the height from the base of the tailing dam, with an initial decrease followed by an increase. The pore pressure that induced structural failure also gradually decreased with increasing height. The increase in the tailing pore pressure could be described by an S-shaped model. A complementary multivariate nonlinear equation was also derived for predicting the tailing pore water pressure under dynamic loading.

  20. Enriched uranium cycles in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Mazzola, A.

    1994-01-01

    A study was made on the substitution of natural uranium with enriched and on plutonium recycle in unmodified PHWRs (pressure vessel reactor). Results clearly show the usefulness of enriched fuel utilisation for both uranium ore consumption (savings of 30% around 1.3% enrichment) and decreasing fuel cycle coasts. This is also due to a better plutonium exploitation during the cycle. On the other hand plutonium recycle in these reactors via MOX-type fuel appears economically unfavourable under any condition

  1. EFFECTS OF PRESSURE AND TEMPERATURE ON ULTRAFILTRATION HOLLOW FIBER MEMBRANE IN MOBILE WATER TREATMENT SYSTEM

    Directory of Open Access Journals (Sweden)

    ROSDIANAH RAMLI

    2016-07-01

    Full Text Available In Sabah, Malaysia, there are still high probability of limited clean water access in rural area and disaster site. Few villages had been affected in Pitas due to improper road access, thus building a water treatment plant there might not be feasible. Recently, Kundasang area had been affected by earthquake that caused water disruption to its people due to the damage in the underground pipes and water tanks. It has been known that membrane technology brought ease in making mobile water treatment system that can be transported to rural or disaster area. In this study, hollow fiber membrane used in a mobile water treatment system due to compact and ease setup. Hollow fiber membrane was fabricated into small module at 15 and 30 fibers to suit the mobile water treatment system for potable water production of at least 80 L/day per operation. The effects of transmembrane pressure (TMP and feed water temperature were investigated. It was found that permeate flux increases by more than 96% for both 15 and 30 fiber bundles with increasing pressure in the range of 0.25 to 3.0 bar but dropped when the pressure reached maximum. Lower temperature of 17 to 18˚C increase the water viscosity by 15% from normal temperature of water at 24˚C, making the permeate flux decreases. The fabricated modules effectively removed 96% turbidity of the surface water sample tested.

  2. 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

  3. 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)

  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. Physical model of lean suppression pressure oscillation phenomena: steam condensation in the light water reactor pressure suppression system (PSS)

    International Nuclear Information System (INIS)

    McCauley, E.W.; Holman, G.S.; Aust, E.; Schwan, H.; Vollbrandt, J.

    1980-01-01

    Using the results of large scale multivent tests conducted by GKSS, a physical model of chugging is developed. The unique combination of accurate digital data and cinematic data has provided the derivation of a detailed, quantified correlation between the dynamic physical variables and the associated two-phase thermo-hydraulic phenomena occurring during lean suppression (chugging) phases of the loss-of-coolant accident in a boiling water reactor pressure suppression system

  6. Type GQS-1 high pressure steam manifold water level monitoring system

    International Nuclear Information System (INIS)

    Li Nianzu; Li Beicheng; Jia Shengming

    1993-10-01

    The GQS-1 high pressure steam manifold water level monitoring system is an advanced nuclear gauge that is suitable for on-line detecting and monitor in high pressure steam manifold water level. The physical variable of water level is transformed into electrical pulses by the nuclear sensor. A computer is equipped for data acquisition, analysis and processing and the results are displayed on a 14 inch color monitor. In addition, a 4 ∼ 20 mA output current is used for the recording and regulation of water level. The main application of this gauge is for on-line measurement of high pressure steam manifold water level in fossil-fired power plant and other industries

  7. Investigation of temperature fluctuation phenomena in a stratified steam-water two-phase flow in a simulating pressurizer spray pipe of a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Miyoshi, Koji, E-mail: miyoshi.koj@inss.co.jp; Takenaka, Nobuyuki; Ishida, Taisuke; Sugimoto, Katsumi

    2017-05-15

    Highlights: • Thermal hydraulics phenomena were discussed in a spray pipe of pressurizer. • Temperature fluctuation was investigated in a stratified steam-water two-phase. • Remarkable liquid temperature fluctuations were observed in the liquid layer. • The observed temperature fluctuations were caused by the internal gravity wave. • The temperature fluctuations decreased with increasing dissolved oxygen. - Abstract: Temperature fluctuation phenomena in a stratified steam-water two-phase flow in a horizontal rectangular duct, which simulate a pressurizer spray pipe of a pressurized water reactor, were studied experimentally. Vertical distributions of the temperature and the liquid velocity were measured with water of various dissolved oxygen concentrations. Large liquid temperature fluctuations were observed when the water was deaerated well and dissolved oxygen concentration was around 10 ppb. The large temperature fluctuations were not observed when the oxygen concentration was higher. It was shown that the observed temperature fluctuations were caused by the internal gravity wave since the Richardson numbers were larger than 0.25 and the temperature fluctuation frequencies were around the Brunt-Väisälä frequencies in the present experimental conditions. The temperature fluctuations decreased by the non-condensable gas since the non-condensable gas suppressed the condensation and the temperature difference in the liquid layer was small.

  8. Water solubility of synthetic pyrope at high temperature and pressure up to 12GPa

    Science.gov (United States)

    Huang, S.; Chen, J.

    2012-12-01

    Water can be incorporated into normally anhydrous minerals as OH- defects and transported into the mantle. Its existence in the mantle may affect property of minerals, such as elasticity, electrical conductivity and rheological properties. As the secondary mineral in the mantle, garnet has not been extensively studied for its water solubility and there is discrepancies among the existing experiments on the water solubility in the garnet change at pressures and temperatures. Geiger et al., 1991 investigated water content in synthetic pyrope and concluded 0.02wt% to 0.07wt% OH- substitution. Lu et al., 1997 found 198ppm water in the Dora Miara pyrope at 100Kbar and 1000°C. Withers et al., 1998 claimed that water solubility in pyrope reached 1000ppm at 5GPa and then decreased as pressure increasing; above 7GPa, no water was detected. Mookherjee et al., 2009 also explored pyrope-rich garnet, which contains water up to 0.1%wt at 5-9GPa and temperatures 1373K-1473K. Here we report a study of water solubility of synthetic single crystal pyrope at pressures 4-12GPa and temperature 1000°C. Single crystals of pyrope were synthesized using multi-anvil press and water contents in these samples were measured using FTIR. We have observed OH- peak at 3650 cm-1 along this pressure range, although Withers, 1998 reported water contents decrease to undetectable level above 7GPa. Water solubility in pyrope will be reported as a function of pressure up to 12 GPa at 1000°C.

  9. French studies and research program in pressurized water reactor safety

    International Nuclear Information System (INIS)

    Duco, J.

    1986-06-01

    The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena [fr

  10. 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

  11. Pressure balanced type membrane covered polarographic oxygen detectors for use in high temperature-high pressure water, (1)

    International Nuclear Information System (INIS)

    Nakayama, Norio; Uchida, Shunsuke

    1984-01-01

    A pressure balanced type membrane covered polarographic oxygen detector was developed to determine directly oxygen concentrations in high temperature, high pressure water without cooling and pressure reducing procedures. The detector is characterized by the following features: (1) The detector body and the membrane for oxygen penetration are made of heat resistant resin. (2) The whole detector body is contained in a pressure chamber where interior and exterior pressures of the detector are balanced. (3) Thermal expansion of the electrolyte is absorbed by deformation of a diaphragm attached to the detector bottom. (4) The effect of dissolved Ag + on the signal current is eliminated by applying a guard electrode. As a result of performance tests at elevated temperature, it was demonstrated that a linear relationship between oxygen concentration and signal current was obtained up to 285 0 C, which was stabilized by the guard electrode. The minimum O 2 concentration detectable was 0.03ppm (9.4 x 10 -7 mol/kg). (author)

  12. 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.

  13. High pressure water electrolysis for space station EMU recharge

    Science.gov (United States)

    Lance, Nick; Puskar, Michael; Moulthrop, Lawrence; Zagaja, John

    1988-01-01

    A high pressure oxygen recharge system (HPORS), is being developed for application on board the Space Station. This electrolytic system can provide oxygen at up to 6000 psia without a mechanical compressor. The Hamilton standard HPORS based on a solid polymer electrolyte system is an extension of the much larger and succesful 3000 psia system of the U.S. Navy. Cell modules have been successfully tested under conditions beyond which spacecraft may encounter during launch. The control system with double redundancy and mechanical backups for all electronically controlled components is designed to ensure a safe shutdown.

  14. Safety of pressurized water reactors: problems and corresponding studies

    International Nuclear Information System (INIS)

    Cogne, F.

    1976-01-01

    The author recalls the safety problems subject to researches in the CEA, either because of their importance or because studies made abroad were not sufficiently developed or were classified or in order to acquire an independent judgement when safety is concerned. Those problems and studies are submitted referring to the 3 existing shields between man and dangerous materials: fuel element can, thermal shield, (pressure vessel and pipes), biological shield of which the behaviour is studied in connection with outside aggressions such as earthquakes, plane crashes, chemical explosions.. [fr

  15. Sloshing of water in annular pressure-suppression pool of boiling water reactors under earthquake ground motions

    International Nuclear Information System (INIS)

    Aslam, M.; Godden, W.G.; Scalise, D.T.

    1979-10-01

    This report presents an analytical investigation of the sloshing response of water in annular-circular as well as simple-circular tanks under horizontal earthquake ground motions, and the results are verified with tests. This study was motivated because of the use of annular tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 120 ft inside and outside diameters and a water depth of 20 ft. The analysis was based upon potential flow theory and a computer program was written to obtain time-history plots of sloshing displacements of water and the dynamic pressures. Tests were carried out on 1/80th and 1/15th scale models under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding the natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found for relatively small displacements. The computer program gave satisfactory results as long as the maximum water surface displacements were less than 30 in., which is roughly the value obtained under full intensity of El Centro earthquake

  16. Failure Mode of the Water-filled Fractures under Hydraulic Pressure in Karst Tunnels

    Directory of Open Access Journals (Sweden)

    Dong Xin

    2017-06-01

    Full Text Available Water-filled fractures continue to grow after the excavation of karst tunnels, and the hydraulic pressure in these fractures changes along with such growth. This paper simplifies the fractures in the surrounding rock as flat ellipses and then identifies the critical hydraulic pressure values required for the occurrence of tensile-shear and compression-shear failures in water-filled fractures in the case of plane stress. The occurrence of tensile-shear fracture requires a larger critical hydraulic pressure than compression-shear failure in the same fracture. This paper examines the effects of fracture strike and lateral pressure coefficient on critical hydraulic pressure, and identifies compression-shear failure as the main failure mode of water-filled fractures. This paper also analyses the hydraulic pressure distribution in fractures with different extensions, and reveals that hydraulic pressure decreases along with the continuous growth of fractures and cannot completely fill a newly formed fracture with water. Fracture growth may be interrupted under the effect of hydraulic tensile shear.

  17. 46 CFR 52.01-110 - Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges...

    Science.gov (United States)

    2010-10-01

    ... § 52.01-110 Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure... 46 Shipping 2 2010-10-01 2010-10-01 false Water-level indicators, water columns, gauge-glass connections, gauge cocks, and pressure gauges (modifies PG-60). 52.01-110 Section 52.01-110 Shipping COAST...

  18. Low internal pressure in femtoliter water capillary bridges reduces evaporation rates.

    Science.gov (United States)

    Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

    2016-03-01

    Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges.

  19. Low internal pressure in femtoliter water capillary bridges reduces evaporation rates

    Science.gov (United States)

    Cho, Kun; Hwang, In Gyu; Kim, Yeseul; Lim, Su Jin; Lim, Jun; Kim, Joon Heon; Gim, Bopil; Weon, Byung Mook

    2016-01-01

    Capillary bridges are usually formed by a small liquid volume in a confined space between two solid surfaces. They can have a lower internal pressure than the surrounding pressure for volumes of the order of femtoliters. Femtoliter capillary bridges with relatively rapid evaporation rates are difficult to explore experimentally. To understand in detail the evaporation of femtoliter capillary bridges, we present a feasible experimental method to directly visualize how water bridges evaporate between a microsphere and a flat substrate in still air using transmission X-ray microscopy. Precise measurements of evaporation rates for water bridges show that lower water pressure than surrounding pressure can significantly decrease evaporation through the suppression of vapor diffusion. This finding provides insight into the evaporation of ultrasmall capillary bridges. PMID:26928329

  20. Capillary pressure as a unique function of electric permittivity and water saturation

    NARCIS (Netherlands)

    Plug, W.J.; Slob, E.; Van Turnhout, J.; Bruining, J.

    2007-01-01

    The relation between capillary pressure (Pc) and interfacial area has been investigated by measuring Pc and the electric permittivity at 100 kHz simultaneously as function of the water saturation, (Sw). Drainage and imbibition experiments have been conducted for sand-distilled water-gas (CO2/N2)

  1. Data quality assurance in pressure transducer-based automatic water level monitoring

    Science.gov (United States)

    Submersible pressure transducers integrated with data loggers have become relatively common water-level measuring devices used in flow or well water elevation measurements. However, drift, linearity, hysteresis and other problems can lead to erroneous data. Researchers at the USDA-ARS in Watkinsvill...

  2. Undermoderated spectrum MOX core study. Pressurized water-type breeder

    International Nuclear Information System (INIS)

    Tochihara, Hiroshi; Komano, Yasuo

    1998-01-01

    The purpose of this development is the advance of the PWR core. The conversion ratio (the breeding ratio) are examined. In the case of heavy water as the coolant, the breeding ratio can be achieved about 1.1 with using a hexagonal lattice and space about 1 mm assembly fuel. In the case of light water coolant, the breeding ratio becomes about 1.0, using a hexagonal lattice and fuel space about 0.5 mm fuel assembly. Here, it reports on the situation of the examination such as the nucleus design of core, the design of fuel assembly, the heat hydraulics design of the core, the structure design and so on. (author)

  3. High conversion ratio plutonium recycle in pressurized water reactors

    International Nuclear Information System (INIS)

    Edlund, M.C.

    1975-01-01

    The use of Pu light water reactors in such a way as to minimise the depletion of Pu needed for future use, and therefore to reduce projected demands for U ore and U enrichment is envisaged. Fuel utilisation in PWRs could be improved by tightly-packed fuel rod lattices with conversion ratios of 0.8 to 0.9 compared with ratios of about 0.5 in Pu recycle designs using fuel to water volume ratios of currently operating PWRs. A conceptual design for the Babcock and Wilcox Company reactors now in operation is presented and for illustrative purposes thermalhydraulic design considerations and the reactor physics are described. Principle considerations in the mechanical design of the fuel assemblies are the effect of hydraulic forces, thermal expansion, and fission gas release. The impact of high conversion ratio plutionium recycle in separative work and natural U requirements for PWRs likely to be in operation by 1985 are examined. (U.K.)

  4. Structure and dynamics of water confined in a graphene nanochannel under gigapascal high pressure: dependence of friction on pressure and confinement.

    Science.gov (United States)

    Yang, Lei; Guo, Yanjie; Diao, Dongfeng

    2017-05-31

    Recently, water flow confined in nanochannels has become an interesting topic due to its unique properties and potential applications in nanofluidic devices. The trapped water is predicted to experience high pressure in the gigapascal regime. Theoretical and experimental studies have reported various novel structures of the confined water under high pressure. However, the role of this high pressure on the dynamic properties of water has not been elucidated to date. In the present study, the structure evolution and interfacial friction behavior of water constrained in a graphene nanochannel were investigated via molecular dynamics simulations. Transitions of the confined water to different ice phases at room temperature were observed in the presence of lateral pressure at the gigapascal level. The friction coefficient at the water/graphene interface was found to be dependent on the lateral pressure and nanochannel height. Further theoretical analyses indicate that the pressure dependence of friction is related to the pressure-induced change in the structure of water and the confinement dependence results from the variation in the water/graphene interaction energy barrier. These findings provide a basic understanding of the dynamics of the nanoconfined water, which is crucial in both fundamental and applied science.

  5. Structure of liquid water at high pressures and temperatures

    CERN Document Server

    Eggert, J H; Loubeyre, P

    2002-01-01

    We report quantitatively accurate structure-factor and radial-distribution-function measurements of liquid water in a diamond-anvil cell (DAC) using x-ray diffraction. During the analysis of our diffraction data, we found it possible (and necessary) to also determine the density. Thus, we believe we present the first-ever diffraction-based determination of a liquid structure factor and equation of state in a DAC experiment.

  6. 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

    Multiple geophysical sensors were used to characterize the underwater pressure field and ground vibrations of a seismic water gun and its suitability to deter the movement of Asian carps (particularly the silver [Hypophthalmichthys molitrix] and bighead [Hypophthalmichthys nobilis] carps) while ensuring the integrity of surrounding structures. The sensors used to collect this information were blast-rated hydrophones, surface- and borehole-mounted geophones, and fixed accelerometers.Results from two separate studies are discussed in this report. The Brandon Road study took place in May 2014, in the Des Plaines River, in a concrete-walled channel downstream of the Brandon Road Lock and Dam near Joliet, Illinois. The Lemont study took place in June 2014, in a segment of the dolomite setblock-lined Chicago Sanitary and Ship Canal near Lemont, Illinois.Two criteria were evaluated to assess the potential deterrence to carp migration, and to minimize the expected effect on nearby structures from discharge of the seismic water gun. The first criterion was a 5-pound-per-square-inch (lb/in2) limit for dynamic underwater pressure variations. The second criterion was a maximum velocity and acceleration disturbance of 0.75 inch per second (in/s) for sensitive machinery (such as the lock gates and pumps) and 2.0 in/s adjacent to canal walls, respectively. The criteria were based on previous studies of fish responses to dynamic pressure variations, and effects of vibrations on the structural integrity of concrete walls.The Brandon Road study evaluated the magnitude and extent of the pressure field created by two water gun configurations in the concrete-walled channel downstream of the lock where channel depths ranged from 11 to 14 feet (ft). Data from a single 80-cubic-inch (in³) water gun set at 6 ft below water surface (bws) produced a roughly cylindrical 5-lb/in2 pressure field 20 ft in radius, oriented vertically, with the radius decreasing to less than 15 ft at the water

  7. Materials characterization for advanced pressurized water reactors: Pt. 2

    International Nuclear Information System (INIS)

    Little, E.A.; Gage, G.

    1994-01-01

    A compilation and overview is presented of the experimental techniques available for characterization of the microstructural changes induced by neutron irradiation of PWR pressure vessel steels, and directed towards monitoring of embrittlement processes by examination of surveillance samples from advanced reactor systems. The microstructural features of significance include copper precipitation, dislocation loop and/or microvoid matrix damage and grain boundary solute segregation. The techniques of transmission electron microscopy, field-emission gun scanning transmission electron microscopy, small angle neutron scattering, positron annihilation and field-ion microscopy have all developed to a degree of sophistication such that they are capable of providing detailed microstructural information in these areas, and afford considerable insight into embrittlement processes when used in combination. (author)

  8. 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)

  9. 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.)

  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. Water permeability of nanoporous graphene at realistic pressures for reverse osmosis desalination

    Energy Technology Data Exchange (ETDEWEB)

    Cohen-Tanugi, David; Grossman, Jeffrey C. [Department of Materials Science and Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States)

    2014-08-21

    Nanoporous graphene (NPG) shows tremendous promise as an ultra-permeable membrane for water desalination thanks to its atomic thickness and precise sieving properties. However, a significant gap exists in the literature between the ideal conditions assumed for NPG desalination and the physical environment inherent to reverse osmosis (RO) systems. In particular, the water permeability of NPG has been calculated previously based on very high pressures (1000–2000 bars). Does NPG maintain its ultrahigh water permeability under real-world RO pressures (<100 bars)? Here, we answer this question by drawing results from molecular dynamics simulations. Our results indicate that NPG maintains its ultrahigh permeability even at low pressures, allowing a permeate water flux of 6.0 l/h-bar per pore, or equivalently 1041 ± 20 l/m{sup 2}-h-bar assuming a nanopore density of 1.7 × 10{sup 13} cm{sup −2}.

  12. Drilling Performance of Rock Drill by High-Pressure Water Jet under Different Configuration Modes

    Directory of Open Access Journals (Sweden)

    Songyong Liu

    2017-01-01

    Full Text Available In the rock drilling progress, the resistant force results in tools failure and the low drilling efficiency; thus, it is necessary to reduce the tools failure and enhance the drilling efficiency. In this paper, different configuration modes of drilling performance assisted with water jet are explored based on the mechanism and experiment analysis of rock drilling assisted with water jet. Moreover, the rotary sealing device with high pressure is designed to achieve the axial and rotation movement simultaneously as well as good sealing effect under high-pressure water jet. The results indicate that the NDB and NFB have better effects on drilling performance compared with that of NSB. Moreover, the high-pressure water jet is helpful not only to reduce the drill rod deflection, but also to reduce the probability of drill rod bending and improve the drill rod service life.

  13. Molecular Dynamics Simulation of Water Nanodroplets on Silica Surfaces at High Air Pressures

    DEFF Research Database (Denmark)

    Zambrano, Harvey A; Jaffe, Richard Lawrence; Walther, Jens Honore

    2010-01-01

    e.g., nanobubbles. In the present work we study the role of air on the wetting of hydrophilic systems. We conduct molecular dynamics simulations of a water nanodroplet on an amorphous silica surface at different air pressures. The interaction potentials describing the silica, water, and air......Silicon dioxides-water systems are abundant in nature and play fundamental roles in a diversity of novel science and engineering applications. Although extensive research has been devoted to study the nature of the interaction between silica and water a complete understanding of the system has...... perform extensive simulations of the water- air equilibrium and calibrate the water-air interaction to match the experimental solubility of N2 and O2 in water. For the silica-water system we calibrate the water-silica interaction to match the experimental contact angle of 27º. We subsequently study...

  14. Sloshing of water in torus pressure-suppression pool of boiling water reactors under earthquake ground motions

    International Nuclear Information System (INIS)

    Aslam, M.; Godden, W.G.; Scalise, D.T.

    1978-08-01

    This report presents an analytical and experimental investigation into the sloshing of water in torus tanks under horizontal earthquake ground motions. This study was motivated because of the use of torus tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 140 ft inside and outside diameters, a 30 ft diameter section, and a water depth of 15 ft. A general finite element analysis was developed for all axisymmetric tanks and a computer program was written to obtain time-history plots of sloshing displacements of water and dynamic pressures. Tests were carried out on a 1/60th scale model under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found within the range of displacements studied. The computer program gave satisfactory results within a maximum range of sloshing displacements in the full-size prototype of 30 in. which is greater than the value obtained under the full intensity of the El Centro earthquake (N-S component 1940). The range of linear behavior was studied experimentally by subjecting the torus model to increasing intensities of the El Centro earthquake

  15. 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)

  16. Assessment of Analytic Water hammer Pressure Model of FAI/08-70

    International Nuclear Information System (INIS)

    Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won

    2016-01-01

    In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative

  17. Assessment of Analytic Water hammer Pressure Model of FAI/08-70

    Energy Technology Data Exchange (ETDEWEB)

    Park, Ju Yeop; Yoo, Seung Hun; Seul, Kwang-Won [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2016-10-15

    In evaluating water hammer effect on the safety related systems, methods developed by the US utility are likely to be adopted in Korea. For example, the US utility developed specific methods to evaluate pressure and loading transient on piping due to water hammer as in FAI/08-70. The methods of FAI/08-70 would be applied in Korea when any regulatory request on the evaluation of water hammer effect due to the non-condensable gas accumulation in the safety related systems. Specifically, FAI/08-70 gives an analytic model which can be used to analyze the maximum transient pressure and maximum transient loading on the piping of the safety-related systems due to the non-condensable induced water hammer effect. Therefore, it seems to be meaningful to review the FAI/08-70 methods and attempt to apply the methods to a specific case to see if they really give reasonable estimate before the application of FAI/08-70 methods to domestic nuclear power plants. In the present study, analytic water hammer pressure model of FAI/08-70 is reviewed in detail and the model is applied to the specific experiment of FAI/08-70 to see if the analytic water hammer pressure model really gives reasonable estimate of the peak water hammer pressure. Specifically, we assess the experiment 52A of FAI/08-70 which adopts flushed initial condition with a short rising piping length and a high level piping length of 51inch. The calculated analytic water hammer pressure peak shows a close agreement with the measured experimental data of 52A. Unfortunately, however, the theoretical value is a little bit less than that of the experimental value. This implies the analytic model of FAI/08-70 is not conservative.

  18. Development of a pressurized bipolar alkaline water electrolyzer

    Energy Technology Data Exchange (ETDEWEB)

    Neves Junior, Newton Pimenta; Pinto, Edgar A. de Godoi Rodrigues; Silva, Ennio Peres da; Rapelli, Rubia; Pinto, Cristiano da Silva [Universidade Estadual de Campinas (DFA/ IFGW/UNICAMP), SP (Brazil). Inst. de Fisica Gleb Wataghin. Dept. de Fisica Aplicada], Email: nevesjr@unicamp.br; Marin Neto, Antonio Jose; Lopes, Daniel Gabriel; Camargo, Joao Carlos; Ferreira, Paulo F.P. [Hydrogen Technology (HyTron), Campinas, SP (Brazil); Furlan, Andre Luis [Universidade Estadual de Campinas (DE/FEC/UNICAMP), SP (Brazil). Fac. de Engenharia Mecanica

    2010-07-01

    This paper reports the actual development status of a bipolar alkaline water electrolyzer with maximum production capacity of 1 m3/h of hydrogen and controlled by a PLC (Programmable Logic Controller), which also interfaces the electrolytic system with operators and other equipment, such as gas storage tanks, fuel cells and photovoltaic panels. The project also includes the construction of an electrolysis test bench to record electrical parameters (cathode, anode, separator and electrolyte potentials), the amount of produced gases and gas quality determined by gas chromatography. (author)

  19. Core construction in a pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Ueda, Makoto; Aoki, Katsutada.

    1975-01-01

    Object: To replace a centrally positioned fuel assembly of a fuel assembly unit with a reactor controlling machinery to decrease a distance between the fuel assemblies thereby saving use of heavy water and enhancing economy. Structure: A centrally positioned fuel assembly of a fuel assembly unit, which is composed of a plurality of fuel assemblies orderly arranged in lattice fashion, is replaced with a reactor controlling members such as control rods, poison tubes and the like to provide an arrangement of lattice-free type fuel assembly, thus reducing the pitch as small as possible. (Kamimura, M.)

  20. Structural integrity of water reactor pressure boundary components

    International Nuclear Information System (INIS)

    Loss, F.J.

    1977-01-01

    The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

  1. Simulation of a pressurized-water nuclear power station

    International Nuclear Information System (INIS)

    Larminaux, Robert; Ourmann, Michel

    1978-01-01

    Faced with the large programme of fitting out PWR nuclear power stations, Electricite de France have undertaken a series of studies with a view to ensuring the best possible adaptation of the secondary part -particularly the feed water heating section- to the nuclear boiler. In order to undertake such studies it has been necessary to finalize simulation models of the entire power station. So as to verify the validity of the models, experiment-calculation comparisons were made during transient operating states recorded at the Ardennes power station as well as during starting up trials at the Tihange I power station [fr

  2. Indirect desalination of Red Sea water with forward osmosis and low pressure reverse osmosis for water reuse

    KAUST Repository

    Yangali-Quintanilla, Victor; Li, Zhenyu; Valladares Linares, Rodrigo; Li, Qingyu; Amy, Gary L.

    2011-01-01

    The use of energy still remains the main component of the costs of desalting water. Forward osmosis (FO) can help to reduce the costs of desalination, and extracting water from impaired sources can be beneficial in this regard. Experiments with FO membranes using a secondary wastewater effluent as a feed water and Red Sea water as a draw solution demonstrated that the technology is promising. FO coupled with low pressure reverse osmosis (LPRO) was implemented for indirect desalination. The system consumes only 50% (~1.5 kWh/m3) of the energy used for high pressure seawater RO (SWRO) desalination (2.5-4 kWh/m3), and produces a good quality water extracted from the impaired feed water. Fouling of the FO membranes was not a major issue during long-term experiments over 14 days. After 10 days of continuous FO operation, the initial flux declined by 28%. Cleaning the FO membranes with air scouring and clean water recovered the initial flux by 98.8%. A cost analysis revealed FO per se as viable technology. However, a minimum average FO flux of 10.5 L/m2-h is needed to compete with water reuse using UF-LPRO, and 5.5 L/m2-h is needed to recover and desalinate water at less cost than SWRO. © 2011 Elsevier B.V.

  3. Indirect desalination of Red Sea water with forward osmosis and low pressure reverse osmosis for water reuse

    KAUST Repository

    Yangali-Quintanilla, Victor

    2011-10-01

    The use of energy still remains the main component of the costs of desalting water. Forward osmosis (FO) can help to reduce the costs of desalination, and extracting water from impaired sources can be beneficial in this regard. Experiments with FO membranes using a secondary wastewater effluent as a feed water and Red Sea water as a draw solution demonstrated that the technology is promising. FO coupled with low pressure reverse osmosis (LPRO) was implemented for indirect desalination. The system consumes only 50% (~1.5 kWh/m3) of the energy used for high pressure seawater RO (SWRO) desalination (2.5-4 kWh/m3), and produces a good quality water extracted from the impaired feed water. Fouling of the FO membranes was not a major issue during long-term experiments over 14 days. After 10 days of continuous FO operation, the initial flux declined by 28%. Cleaning the FO membranes with air scouring and clean water recovered the initial flux by 98.8%. A cost analysis revealed FO per se as viable technology. However, a minimum average FO flux of 10.5 L/m2-h is needed to compete with water reuse using UF-LPRO, and 5.5 L/m2-h is needed to recover and desalinate water at less cost than SWRO. © 2011 Elsevier B.V.

  4. Superheated steam annealing of pressurized water reactor vessel

    International Nuclear Information System (INIS)

    Porowski, J.S.

    1993-01-01

    Thermal annealing of an embrittled Reactor Pressure Shell is the only recognized means for recovering material properties lost due to long-term exposure of the reactor walls to radiation. Reduced toughness of the material during operation is a major concern in evaluations of structural integrity of older reactors. Extensive studies performed within programs related to life extension of nuclear plants have confirmed that the thermal treatment of 850 deg. F for 168 hours on irradiated material essentially recovers material properties lost due to neutron exposure. Dry and wet annealing methods have been considered. Wet annealing involves operating the reactor at near design temperatures and pressures. Since the temperature of wet annealing must be limited to vessel design temperature of 650 deg. F, only partial recovery of the lost properties is achieved. Thus dry annealing was selected as an alternative for future development and industrial implementation to extend the safe life of reactors. Dry thermal annealing consists of heating portions of the reactor vessel at a specific temperature for a given period of time using a high temperature heat source. The use of spent fuel assemblies, induction heating and resistance heating elements as well as the circulation of heated fluid were investigated as potential candidate methods. To date the use of resistance heating elements which are lowered into a dry empty reactor was considered to be the preferred method. In-depth research in the United States and practical applications of such a method in Russia have confirmed feasibility of the method. The method of using circulating superheated steam to anneal the vessel at 850 deg. F without complete removal of the reactor internals is described herein. After removing the reactor head and fuel, the core barrel along with the upper and lower core in PWRs is lifted to open an annular space between the reactor shell flange and the core barrel flange. The thermal shield can remain

  5. SATCAP-C : a program for thermal hydraulic design of pressurized water injection type capsule

    International Nuclear Information System (INIS)

    Harayama, Yasuo; Someya, Hiroyuki; Asoh, Tomokazu; Niimi, Motoji

    1992-10-01

    There are capsules called 'Pressure Water Injection Type Capsule' as a kind of irradiation devices at the Japan Materials Testing Reactor (JMTR). A type of the capsules is a 'Boiling Water Capsule' (usually named BOCA). The other type is a 'Saturated Temperature Capsule' (named SATCAP). When the water is kept at a constant pressure, the water temperature does not become higher than the saturated temperature so far as the water does not fully change to steam. These type capsules are designed on the basis of the conception of applying the water characteristic to the control of irradiation temperature of specimens in the capsules. In designing of the capsules in which the pressurized water is injected, thermal performances have to be understood as exactly as possible. It is not easy however to predict thermal performances such as axially temperature distribution of water injected in the capsule, because there are heat-sinks at both side of inner and outer of capsule casing as the result that the water is fluid. Then, a program (named SATCAP-C) for the BOCA and SATCAP was compiled to grasp the thermal performances in the capsules and has been used the design of the capsules and analysis of the data obtained from some actual irradiation capsules. It was confirmed that the program was effective in thermal analysis for the capsules. The analysis found out the values for heat transfer coefficients at various surfaces of capsule components and some thermal characteristics of capsules. (author)

  6. When immiscible becomes miscible-Methane in water at high pressures.

    Science.gov (United States)

    Pruteanu, Ciprian G; Ackland, Graeme J; Poon, Wilson C K; Loveday, John S

    2017-08-01

    At low pressures, the solubility of gases in liquids is governed by Henry's law, which states that the saturated solubility of a gas in a liquid is proportional to the partial pressure of the gas. As the pressure increases, most gases depart from this ideal behavior in a sublinear fashion, leveling off at pressures in the 1- to 5-kbar (0.1 to 0.5 GPa) range with solubilities of less than 1 mole percent (mol %). This contrasts strikingly with the well-known marked increase in solubility of simple gases in water at high temperature associated with the critical point (647 K and 212 bar). The solubility of the smallest hydrocarbon, the simple gas methane, in water under a range of pressure and temperature is of widespread importance, because it is a paradigmatic hydrophobe and occurs widely in terrestrial and extraterrestrial geology. We report measurements up to 3.5 GPa of the pressure dependence of the solubility of methane in water at 100°C-well below the latter's critical temperature. Our results reveal a marked increase in solubility between 1 and 2 GPa, leading to a state above 2 GPa where the maximum solubility of methane in water exceeds 35 mol %.

  7. Water vapor pressure over molten KH_2PO_4 and demonstration of water electrolysis at ∼300 °C

    International Nuclear Information System (INIS)

    Berg, R.W.; Nikiforov, A.V.; Petrushina, I.M.; Bjerrum, N.J.

    2016-01-01

    Highlights: • The vapor pressure over molten KH_2PO_4 was measured by Raman spectroscopy to be about 8 bars at ∼300 °C. • Raman spectroscopy shows that molten KH_2PO_4 under its own vapor pressure contains much dissolved water. • It is demonstrated spectroscopically that water electrolysis is possible in KH_2PO_4 electrolyte forming H_2 and O_2 at 300 °C. • Molten KH_2PO_4 is a possible electrolyte for water electrolysis. - Abstract: A new potentially high-efficiency electrolyte for water electrolysis: molten monobasic potassium phosphate, KH_2PO_4 or KDP has been investigated at temperatures ∼275–325 °C. At these temperatures, KH_2PO_4 was found to dissociate into H_2O gas in equilibrium with a melt mixture of KH_2PO_4−K_2H_2P_2O_7−KPO_3−H_2O. The water vapor pressure above the melt, when contained in a closed ampoule, was determined quantitatively vs. temperature by use of Raman spectroscopy with methane or hydrogen gas as an internal calibration standard, using newly established relative ratios of Raman scattering cross sections of water and methane or hydrogen to be 0.40 ± 0.02 or 1.2 ± 0.03. At equilibrium the vapor pressure was much lower than the vapor pressure above liquid water at the same temperature. Electrolysis was realized by passing current through closed ampoules (vacuum sealed quartz glass electrolysis cells with platinum electrodes and the electrolyte melt). The formation of mixtures of hydrogen and oxygen gases as well as the water vapor was detected by Raman spectroscopy. In this way it was demonstrated that water is present in the new electrolyte: molten KH_2PO_4 can be split by electrolysis via the reaction 2H_2O → 2H_2 + O_2 at temperatures ∼275–325 °C. At these temperatures, before the start of the electrolysis, the KH_2PO_4 melt gives off H_2O gas that pressurizes the cell according to the following dissociations: 2KH_2PO_4 ↔ K_2H_2P_2O_7 + H_2O ↔ 2KPO_3 + 2H_2O. The spectra show however that the water by

  8. Fibre optic sensors in pressurized water reactor alternators

    International Nuclear Information System (INIS)

    Favennec, J.M.; Piguet, M.

    1994-01-01

    Measurement in the electrical engine environment (alternator, transformer...) is identified as one of the two main applications of fibre optic sensors within EDF; the other application niche is the monitoring of civil works (dams, containment building of nuclear reactors...). At the EDF Research and Development Division, temperature and vibration fibre optic sensors were evaluated by the Metrology Service, since their use is under consideration for alternator monitoring. For alternator stator thermal monitoring, the BERTIN company developed a fibre optic sensor network. The optic coding technique is based on broadband source spectral modulation; the sensors are interrogated sequentially by electronic commutation. For alternator stator vibration monitoring, a fibre optic accelerometer was developed in the frame of a manufacturers and universities consortium supported by the French Research and Technology Ministry. The accelerometer is of cantilever beam type and its networking is possible by chromatic multiplexing. The Metrology Service evaluated these temperature and vibration sensors in order to verify their metrological characteristics (bias error, hysteresis, repeatability, resolution, noise, amplitude linearity, response time, frequency response, etc.) and to test their behaviour in harsh alternator environmental conditions (pressure, vibrations and temperature). Ageing and accidental condition resistance tests were also carried out. Temperature sensor test results were very satisfactory. An eight-sensor BERTIN prototype was installed on the Tricastin 1 alternator during the september 1993 nuclear station periodic stop. On the contrary, the accelerometers presented deficient metrological characteristics (shorter span than foreseen, low repeatability...). They need some improvements and could not be installed on alternators. (authors). 5 refs., 8 figs

  9. Pressure probe study of the water relations of Phycomyces blakesleeanus sporangiophores

    Science.gov (United States)

    Cosgrove, D. J.; Ortega, J. K.; Shropshire, W. Jr

    1987-01-01

    The physical characteristics which govern the water relations of the giant-celled sporangiophore of Phycomyces blakesleeanus were measured with the pressure probe technique and with nanoliter osmometry. These properties are important because they govern water uptake associated with cell growth and because they may influence expansion of the sporangiophore wall. Turgor pressure ranged from 1.1 to 6.6 bars (mean = 4.1 bars), and was the same for stage I and stage IV sporangiophores. Sporangiophore osmotic pressure averaged 11.5 bars. From the difference between cell osmotic pressure and turgor pressure, the average water potential of the sporangiophore was calculated to be about -7.4 bars. When sporangiophores were submerged under water, turgor remained nearly constant. We propose that the low cell turgor pressure is due to solutes in the cell wall solution, i.e., between the cuticle and the plasma membrane. Membrane hydraulic conductivity averaged 4.6 x 10(-6) cm s-1 bar-1, and was significantly greater in stage I sporangiophores than in stage IV sporangiophores. Contrary to previous reports, the sporangiophore is separated from the supporting mycelium by septa which prevent bulk volume flow between the two regions. The presence of a wall compartment between the cuticle and the plasma membrane results in anomalous osmosis during pressure clamp measurements. This behavior arises because of changes in solute concentration as water moves into or out of the wall compartment surrounding the sporangiophore. Theoretical analysis shows how the equations governing transient water flow are altered by the characteristics of the cell wall compartment.

  10. Derivation of Elastic Stress Concentration Factor Equations for Debris Fretting Flaws in Pressure Tubes of Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Jong Sung; Oh, Young Jin

    2014-01-01

    If volumetric flaws such as bearing pad fretting flaws and debris fretting flaws are detected in the pressure tubes of pressurized heavy water reactors during in-service inspection, the initiation of fatigue cracks and delayed hydrogen cracking from the detected volumetric flaws shall be assessed by using elastic stress concentration factors in accordance with CSA N285.8-05. The CSA N285.8-05 presents only an approximate formula based on linear elastic fracture mechanics for the debris fretting flaw. In this study, an engineering formula considering the geometric characteristics of the debris fretting flaw in detail was derived using two-dimensional finite element analysis and Kinectrics, Inc.'s engineering procedure with slight modifications. Comparing the application results obtained using the derived formula with the three-dimensional finite element analysis results, it is found that the results obtained using the derived formula agree well with the results of the finite element analysis

  11. 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.)

  12. Development of Extended Period Pressure-Dependent Demand Water Distribution Models

    Energy Technology Data Exchange (ETDEWEB)

    Judi, David R. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Mcpherson, Timothy N. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-03-20

    Los Alamos National Laboratory (LANL) has used modeling and simulation of water distribution systems for N-1 contingency analyses to assess criticality of water system assets. Critical components considered in these analyses include pumps, tanks, and supply sources, in addition to critical pipes or aqueducts. A contingency represents the complete removal of the asset from system operation. For each contingency, an extended period simulation (EPS) is run using EPANET. An EPS simulates water system behavior over a time period, typically at least 24 hours. It assesses the ability of a system to respond and recover from asset disruption through distributed storage in tanks throughout the system. Contingencies of concern are identified as those in which some portion of the water system has unmet delivery requirements. A delivery requirement is defined as an aggregation of water demands within a service area, similar to an electric power demand. The metric used to identify areas of unmet delivery requirement in these studies is a pressure threshold of 15 pounds per square inch (psi). This pressure threshold is used because it is below the required pressure for fire protection. Any location in the model with pressure that drops below this threshold at any time during an EPS is considered to have unmet service requirements and is used to determine cascading consequences. The outage area for a contingency is the aggregation of all service areas with a pressure below the threshold at any time during the EPS.

  13. Experimental Study on Peak Pressure of Shock Waves in Quasi-Shallow Water

    Directory of Open Access Journals (Sweden)

    Zhenxiong Wang

    2015-01-01

    Full Text Available Based on the similarity laws of the explosion, this research develops similarity requirements of the small-scale experiments of underwater explosions and establishes a regression model for peak pressure of underwater shock waves under experimental condition. Small-scale experiments are carried out with two types of media at the bottom of the water and for different water depths. The peak pressure of underwater shock waves at different measuring points is acquired. A formula consistent with the similarity law of explosions is obtained and an analysis of the regression precision of the formula confirms its accuracy. Significance experiment indicates that the influence of distance between measuring points and charge on peak pressure of underwater shock wave is the greatest and that of water depth is the least within the range of geometric parameters. An analysis of data from experiments with different media at the bottom of the water reveals an influence on the peak pressure, as the peak pressure of a shock wave in a body of water with a bottom soft mud and rocks is about 1.33 times that of the case where the bottom material is only soft mud.

  14. Flooding of a large, passive, pressure-tube light water reactor

    International Nuclear Information System (INIS)

    Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

    1997-01-01

    A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

  15. Spectroscopic and thermodynamic properties of molecular hydrogen dissolved in water at pressures up to 200 MPa

    Science.gov (United States)

    Borysow, Jacek; del Rosso, Leonardo; Celli, Milva; Moraldi, Massimo; Ulivi, Lorenzo

    2014-04-01

    We have measured the Raman Q-branch of hydrogen in a solution with water at a temperature of about 280 K and at pressures from 20 to 200 MPa. From a least-mean-square fitting analysis of the broad Raman Q-branch, we isolated the contributions from the four lowest individual roto-vibrational lines. The vibrational lines were narrower than the pure rotational Raman lines of hydrogen dissolved in water measured previously, but significantly larger than in the gas. The separations between these lines were found to be significantly smaller than in gaseous hydrogen and their widths were slightly increasing with pressure. The lines were narrowing with increasing rotational quantum number. The Raman frequencies of all roto-vibrational lines were approaching the values of gas phase hydrogen with increasing pressure. Additionally, from the comparison of the integrated intensity signal of Q-branch of hydrogen to the integrated Raman signal of the water bending mode, we have obtained the concentration of hydrogen in a solution with water along the 280 K isotherm. Hydrogen solubility increases slowly with pressure, and no deviation from a smooth behaviour was observed, even reaching thermodynamic conditions very close to the transition to the stable hydrogen hydrate. The analysis of the relative hydrogen concentration in solution on the basis of a simple thermodynamic model has allowed us to obtain the molar volume for the hydrogen gas/water solution. Interestingly, the volume relative to one hydrogen molecule in solution does not decrease with pressure and, at high pressure, is larger than the volume pertinent to one molecule of water. This is in favour of the theory of hydrophobic solvation, for which a larger and more stable structure of the water molecules is expected around a solute molecule.

  16. Determining soil hydrologic characteristics on a remote forest watershed by continuous monitoring of soil water pressures, rainfall and runoff.

    Science.gov (United States)

    L.R. Ahuja; S. A. El-Swaify

    1979-01-01

    Continuous monitoring of soil-water pressures, rainfall and runoff under natural conditions was tested as a technique for determining soil hydrologic characteristics of a remote forest watershed plot. A completely battery-powered (and thus portable) pressure transducer–scanner–recorder system was assembled for monitoring of soil-water pressures in...

  17. Crop yields response to water pressures in the Ebro basin in Spain: risk and water policy implications

    Science.gov (United States)

    Quiroga, S.; Fernández-Haddad, Z.; Iglesias, A.

    2011-02-01

    The increasing pressure on water systems in the Mediterranean enhances existing water conflicts and threatens water supply for agriculture. In this context, one of the main priorities for agricultural research and public policy is the adaptation of crop yields to water pressures. This paper focuses on the evaluation of hydrological risk and water policy implications for food production. Our methodological approach includes four steps. For the first step, we estimate the impacts of rainfall and irrigation water on crop yields. However, this study is not limited to general crop production functions since it also considers the linkages between those economic and biophysical aspects which may have an important effect on crop productivity. We use statistical models of yield response to address how hydrological variables affect the yield of the main Mediterranean crops in the Ebro river basin. In the second step, this study takes into consideration the effects of those interactions and analyzes gross value added sensitivity to crop production changes. We then use Montecarlo simulations to characterize crop yield risk to water variability. Finally we evaluate some policy scenarios with irrigated area adjustments that could cope in a context of increased water scarcity. A substantial decrease in irrigated land, of up to 30% of total, results in only moderate losses of crop productivity. The response is crop and region specific and may serve to prioritise adaptation strategies.

  18. 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

  19. Elasticity of water-saturated rocks as a function of temperature and pressure.

    Science.gov (United States)

    Takeuchi, S.; Simmons, G.

    1973-01-01

    Compressional and shear wave velocities of water-saturated rocks were measured as a function of both pressure and temperature near the melting point of ice to confining pressure of 2 kb. The pore pressure was kept at about 1 bar before the water froze. The presence of a liquid phase (rather than ice) in microcracks of about 0.3% porosity affected the compressional wave velocity by about 5% and the shear wave velocity by about 10%. The calculated effective bulk modulus of the rocks changes rapidly over a narrow range of temperature near the melting point of ice, but the effective shear modulus changes gradually over a wider range of temperature. This phenomenon, termed elastic anomaly, is attributed to the existence of liquid on the boundary between rock and ice due to local stresses and anomalous melting of ice under pressure.

  20. 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.

  1. Depletion optimization of lumped burnable poisons in pressurized water reactors

    International Nuclear Information System (INIS)

    Kodah, Z.H.

    1982-01-01

    Techniques were developed to construct a set of basic poison depletion curves which deplete in a monotonical manner. These curves were combined to match a required optimized depletion profile by utilizing either linear or non-linear programming methods. Three computer codes, LEOPARD, XSDRN, and EXTERMINATOR-2 were used in the analyses. A depletion routine was developed and incorporated into the XSDRN code to allow the depletion of fuel, fission products, and burnable poisons. The Three Mile Island Unit-1 reactor core was used in this work as a typical PWR core. Two fundamental burnable poison rod designs were studied. They are a solid cylindrical poison rod and an annular cylindrical poison rod with water filling the central region.These two designs have either a uniform mixture of burnable poisons or lumped spheroids of burnable poisons in the poison region. Boron and gadolinium are the two burnable poisons which were investigated in this project. Thermal self-shielding factor calculations for solid and annular poison rods were conducted. Also expressions for overall thermal self-shielding factors for one or more than one size group of poison spheroids inside solid and annular poison rods were derived and studied. Poison spheroids deplete at a slower rate than the poison mixture because each spheroid exhibits some self-shielding effects of its own. The larger the spheroid, the higher the self-shielding effects due to the increase in poison concentration

  2. Light water reactor pressure isolation valve performance testing

    International Nuclear Information System (INIS)

    Neely, H.H.; Jeanmougin, N.M.; Corugedo, J.J.

    1990-07-01

    The Light Water Reactor Valve Performance Testing Program was initiated by the NRC to evaluate leakage as an indication of valve condition, provide input to Section XI of the ASME Code, evaluate emission monitoring for condition and degradation and in-service inspection techniques. Six typical check and gate valves were purchased for testing at typical plant conditions (550F at 2250 psig) for an assumed number of cycles for a 40-year plant lifetime. Tests revealed that there were variances between the test results and the present statement of the Code; however, the testing was not conclusive. The life cycle tests showed that high tech acoustic emission can be utilized to trend small leaks, that specific motor signature measurement on gate valves can trend and indicate potential failure, and that in-service inspection techniques for check valves was shown to be both feasible and an excellent preventive maintenance indicator. Life cycle testing performed here did not cause large valve leakage typical of some plant operation. Other testing is required to fully understand the implication of these results and the required program to fully implement them. (author)

  3. Pipe rupture hardware minimization in pressurized water reactor system

    International Nuclear Information System (INIS)

    Mukherjee, S.K.; Szyslowski, J.J.; Chexal, V.; Norris, D.M.; Goldstein, N.A.; Beaudoin, B.; Quinones, D.; Server, W.

    1987-01-01

    For much of the high energy piping in light water reactor systems, fracture mechanics calculations can be used to assure pipe failure resistance, thus allowing the elimination of excessive rupture restraint hardware both inside and outside containment. These calculations use the concept of leak-before-break (LBB) and include part-through-wall flaw fatigue crack propagation, through-wall flaw detectable leakage, and through-wall flaw stability analyses. Performing these analyses not only reduces initial construction, future maintenance, and radiation exposure costs, but the overall safety and integrity of the plant are improved since much more is known about the piping and its capabilities than would be the case had the analyses not been performed. This paper presents the LBB methodology applied at Beaver Valley Power Station - Unit 2 (BVPS-2); the application for two specific lines, one inside containment (stainless steel) and the other outside containment (ferritic steel), is shown in a generic sense using a simple parametric matrix. The overall results for BVPS-2 indicate that pipe rupture hardware is not necessary for stainless steel lines inside containment greater than or equal to 6-in (152 mm) nominal pipe size that have passed a screening criteria designed to eliminate potential problem systems (such as the feedwater system). Similarly, some ferritic steel lines as small as 3-in (76 mm) diameter (outside containment) can qualify for pipe rupture hardware elimination

  4. Taipower's transient analysis methodology for pressurized water reactors

    International Nuclear Information System (INIS)

    Huang, Pinghue

    1998-01-01

    The methodology presented in this paper is a part of the 'Taipower's Reload Design and Transient Analysis Methodologies for Light Water Reactors' developed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research. This methodology utilizes four computer codes developed or sponsored by Electric Power Research institute: system transient analysis code RETRAN-02, core thermal-hydraulic analysis code COBRAIIIC, three-dimensional spatial kinetics code ARROTTA, and fuel rod evaluation code FREY. Each of the computer codes was extensively validated. Analysis methods and modeling techniques were conservatively established for each application using a systematic evaluation with the assistance of sensitivity studies. The qualification results and analysis methods were documented in detail in TPC topical reports. The topical reports for COBRAIIIC, ARROTTA. and FREY have been reviewed and approved by the Atomic Energy Council (ABC). TPC 's in-house transient methodology have been successfully applied to provide valuable support for many operational issues and plant improvements for TPC's Maanshan Units I and 2. Major applications include the removal of the resistance temperature detector bypass system, the relaxation of the hot-full-power moderator temperature coefficient design criteria imposed by the ROCAEC due to a concern on Anticipated Transient Without Scram, the reduction of boron injection tank concentration and the elimination of the heat tracing, and the reduction of' reactor coolant system flow. (author)

  5. Early response of pressurized hot water in a pipe to a sudden break. Final report

    International Nuclear Information System (INIS)

    Alamgir, M.; Kan, C.Y.; Lienhard, J.H.

    1981-06-01

    Experimental and analytic studies that explain the details of early pressure variations during rapid depressurization in water-cooled reactors are presented as a means of assessing sudden break consequences in a coolant pipe. The report includes (1) a description of the experiment, (2) an analysis of the new bubble growth law for thermally controlled growth of vapor bubbles in an exponentially-varying pressure field, and (3) a review of previous studies and additional observations of blowdown behavior

  6. Application of the regulations on pressurized components or light water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Barthelemy, F.; Menjon, G.

    1977-01-01

    This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

  7. On-off controller for installation to test the pressurized water reactor material

    International Nuclear Information System (INIS)

    Zauq, M.H.

    1982-05-01

    This report describes the design of an ON-OFF controller based on the 6800 microprocessor in its assembly language and its interfacing with its environment (sensors, periphery, etc). The controller is meant to control the temperature and the pressure inside an experimental chamber in which the material under test is placed. The ''Design basis accident'' conditions (e.g., LOCA) for a pressurized water reactor are simulated in the experimental chamber [fr

  8. Comparison of air-charged and water-filled urodynamic pressure measurement catheters.

    Science.gov (United States)

    Cooper, M A; Fletter, P C; Zaszczurynski, P J; Damaser, M S

    2011-03-01

    Catheter systems are utilized to measure pressure for diagnosis of voiding dysfunction. In a clinical setting, patient movement and urodynamic pumps introduce hydrostatic and motion artifacts into measurements. Therefore, complete characterization of a catheter system includes its response to artifacts as well its frequency response. The objective of this study was to compare the response of two disposable clinical catheter systems: water-filled and air-charged, to controlled pressure signals to assess their similarities and differences in pressure transduction. We characterized frequency response using a transient step test, which exposed the catheters to a sudden change in pressure; and a sinusoidal frequency sweep test, which exposed the catheters to a sinusoidal pressure wave from 1 to 30 Hz. The response of the catheters to motion artifacts was tested using a vortex and the response to hydrostatic pressure changes was tested by moving the catheter tips to calibrated heights. Water-filled catheters acted as an underdamped system, resonating at 10.13 ± 1.03 Hz and attenuating signals at frequencies higher than 19 Hz. They demonstrated significant motion and hydrostatic artifacts. Air-charged catheters acted as an overdamped system and attenuated signals at frequencies higher than 3.02 ± 0.13 Hz. They demonstrated significantly less motion and hydrostatic artifacts than water-filled catheters. The transient step and frequency sweep tests gave comparable results. Air-charged and water-filled catheters respond to pressure changes in dramatically different ways. Knowledge of the characteristics of the pressure-measuring system is essential to finding the best match for a specific application. Copyright © 2011 Wiley-Liss, Inc.

  9. Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

    International Nuclear Information System (INIS)

    Lai, W.; Collins, E.K.

    1977-01-01

    This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

  10. Effect of pressurized water reactor environment on fatigue crack propagation, including hole times

    International Nuclear Information System (INIS)

    Bamford, W.H.

    1976-01-01

    Results are presented from an experimental program being conducted to investigate the effects of pressurized water reactor environment on the fatigue crack growth rate of pressure vessel steels. Tests were conducted on precracked WOL type specimens under load controlled conditions. The effects of R ratio, loading rates, and loading wave form are evaluated, and the results are compared for both forging and plate material, as well as weldments

  11. 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

  12. Evaluation of water cooled supersonic temperature and pressure probes for application to 1366 K flows

    Science.gov (United States)

    Lagen, Nicholas; Seiner, John M.

    1990-01-01

    Water cooled supersonic probes are developed to investigate total pressure, static pressure, and total temperature in high-temperature jet plumes and thereby determine the mean flow properties. Two probe concepts, designed for operation at up to 1366 K in a Mach 2 flow, are tested on a water cooled nozzle. The two probe designs - the unsymmetric four-tube cooling configuration and the symmetric annular cooling design - take measurements at 755, 1089, and 1366 K of the three parameters. The cooled total and static pressure readings are found to agree with previous test results with uncooled configurations. The total-temperature probe, however, is affected by the introduction of water coolant, and effect which is explained by the increased heat transfer across the thermocouple-bead surface. Further investigation of the effect of coolant on the temperature probe is proposed to mitigate the effect and calculate more accurate temperatures in jet plumes.

  13. The effect of KZK pressure equation on the sonoluminescence in water and fat tissues

    International Nuclear Information System (INIS)

    Gheshlaghi, M.; Sadighi-Bonabi, R.; Ghadirifar, A.

    2015-01-01

    The effect of the produced light flashes from sonoluminescence (SL) on the fat tissue and water is studied. By using KZK equation as an essential equation for calculating the thermal source in bio-liquids, the effective bubble parameters in quasi-adiabatic model are calculated and compared in these systems. It is noticed that the temperature and the intensity for fat tissue are about 30% and 38% less than the ones for water respectively. These results are almost in good agreement with the only experimental measurement denoting less SL temperature in bio-liquids which present more suitable condition for using SL in such applications. - Highlights: • Coupling of acoustic pressure and the pressure's KZK equation for using Sonoluminescence equations. • The Sonoluminescence parameters (temperature, pressure and intensity) are calculated and Compared for water and fat tissue. • The high-intensity radiation of Sonoluminescence bubble is used in medical applications

  14. The effect of KZK pressure equation on the sonoluminescence in water and fat tissues

    Energy Technology Data Exchange (ETDEWEB)

    Gheshlaghi, M. [Payame Noor University, P.O.B. 19395-3697, Tehran (Iran, Islamic Republic of); Sadighi-Bonabi, R., E-mail: Sadighi@sharif.ir [Department of Physics, Sharif University of Technology, 11365-91, Tehran (Iran, Islamic Republic of); Ghadirifar, A. [Islamic Azad University, Faculty of Mechanical Engineering, Mashhad (Iran, Islamic Republic of)

    2015-09-25

    The effect of the produced light flashes from sonoluminescence (SL) on the fat tissue and water is studied. By using KZK equation as an essential equation for calculating the thermal source in bio-liquids, the effective bubble parameters in quasi-adiabatic model are calculated and compared in these systems. It is noticed that the temperature and the intensity for fat tissue are about 30% and 38% less than the ones for water respectively. These results are almost in good agreement with the only experimental measurement denoting less SL temperature in bio-liquids which present more suitable condition for using SL in such applications. - Highlights: • Coupling of acoustic pressure and the pressure's KZK equation for using Sonoluminescence equations. • The Sonoluminescence parameters (temperature, pressure and intensity) are calculated and Compared for water and fat tissue. • The high-intensity radiation of Sonoluminescence bubble is used in medical applications.

  15. Solubility of hydrogen in water in a broad temperature and pressure range

    International Nuclear Information System (INIS)

    Baranenko, V.I.; Kirov, V.S.

    1989-01-01

    In the coolant of water-water reactors, as a result of radiolytic decomposition of water and chemical additives (hydrazine and ammonia) and saturation of the make-up water of the first loop with free hydrogen in order to suppress radiolysis, 30-60 ml/kg of hydrogen is present in normal conditions. On being released from the water, it is free to accumulate in micropores of the metals, resulting in hydrogen embrittlement; gas accumulates in stagnant zones, with deterioration in heat transfer in the first loop and corresponding difficulty in the use of the reactor and the whole reactor loop. To determine the amount of free hydrogen and hydrogen dissolved in water in different elements of the first loop, it is necessary to know the limiting solubility of hydrogen in water at different temperatures and pressures, and also to have the corresponding theoretical dependences. The experimental data on the solubility of hydrogen in water are nonsystematic and do not cover the parameter ranges of modern nuclear power plants (P = 10-30 MPa, T = 260-370C). Therefore, the aim of the present work is to establish a well-founded method of calculating the limiting solubility of hydrogen in water and, on this basis, to compile tables of the limiting solubility of hydrogen in water at pressures 0.1-50 MPa and temperatures 0-370C

  16. 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.

  17. Total dissolved gas, barometric pressure, and water temperature data, lower Columbia River, Oregon and Washington, 1996

    Science.gov (United States)

    Tanner, Dwight Q.; Harrison, Howard E.; McKenzie, Stuart W.

    1996-01-01

    Increased levels of total dissolved gas pressure can cause gas-bubble trauma in fish downstream from dams on the Columbia River. In cooperation with the U.S. Army Corps of Engineers, the U.S. Geological Survey collected data on total dissolved gas pressure, barometric pressure, water temperature, and dissolved oxygen pressure at 11 stations on the lower Columbia River from the John Day forebay (river mile 215.6) to Wauna Mill (river mile 41.9) from March to September 1996. Methods of data collection, review, and processing are described in this report. Summaries of daily minimum, maximum, and mean hourly values are presented for total dissolved gas pressure, barometric pressure, and water temperature. Hourly values for these parameters are presented graphically. Dissolved oxygen data are not presented in this report because the quality-control data show that the data have poor precision and high bias. Suggested changes to monitoring procedures for future studies include (1) improved calibration procedures for total dissolved gas and dissolved oxygen to better define accuracy at elevated levels of supersaturation and (2) equipping dissolved oxygen sensors with stirrers because river velocities at the shoreline monitoring stations probably cannot maintain an adequate flow of water across the membrane surface of the dissolved oxygen sensor.

  18. [Variation of water DOC during the process of pre-pressure and coagulation sedimentation treatment].

    Science.gov (United States)

    Chen, Wen-Jing; Cong, Hai-Bing; Xu, Ya-Jun; Wang, Wei; Jiang, Xin-Yue; Liu, Yu-Jiao

    2014-07-01

    The aim of the study was to explore whether the pre-pressure and coagulation sedimentation process would result in algal cell disruption, leading to increased dissolved organic carbon (DOC) in water, based on which, the pressure application mode would be optimized and safe and efficient pre-pressure algae removal process would be obtained. The changes in DOC during the process of pre-pressure and preoxidation treatment, the distribution of molecular weight in water as well as the removal efficiency of algae, turbidity and DOC after coagulation and sedimentation were investigated. The results showed that the DOC in water did not increase but decreased, and the molecular weight decreased after treated with 0.5-0.8 MPa pressure. While KMnO4 and NaClO pre-oxidation both increased the DOC, in the meanwhile, the distribution of molecular weight showed no obvious change. After the pre-pressure coagulation and sedimentation process, the removal rate of algae was 96.23% and that of DOC was 29. 11%, which was by 10% - 30% higher than the rate of pre-oxidation coagulation and sedimentation process.

  19. Innovative Control concepts for German pressurized water reactors

    International Nuclear Information System (INIS)

    Brzozowski, Raphael; Kuhn, Andreas

    2010-01-01

    Controlling reactor power without any manual support is becoming more and more important. The READIG project (READIG = Reactor Instrumentation and Digital Control) power control system installed in unit 2 of the Philippsburg nuclear power station (KKP 2) requires no manual intervention except for specific strategy criteria settings. It was even possible to eliminate the power distribution set points. With minor adaptations, this concept can be applied in other PWR plants as well. KKP 2 is a PWR plant with particularly sophisticated core charges; as a consequence, the I and C systems were adapted accordingly. The increase in integral reactor power and the low-leakage core charges are the main reasons for lower limiting margins, especially in peak limiting. The standard control concept was supplemented in such a way that a more precise fine control concept for power distribution in the full-load regime is achieved. The READIG project fully utilizes the possibilities offered by digital TXS Technology, which is why use is also made of physical parameterization. The new power distribution control concept has these advantages: - Operation at small peak-/DNB-reactor output limitation margins. - Stable control without manual intervention also in load cycles and in the frequency control mode. - Simplified operation due to omission of the power distribution set point. - Reduction to zero of the frequency of L-bank steps at constant power with superimposed frequency control mode. - Reduction to zero of the frequency of D-bank steps at constant power with superimposed frequency control mode. - Lower quantities of demineralized water to be fed at constant power with superimposed frequency control mode (±1%). (orig.)

  20. Phase transition and water incorporation into Eu2Sn2O7 pyrochlore at high pressure

    Science.gov (United States)

    Zhang, F. X.; Lang, M.; Ewing, R. C.

    2016-04-01

    Structural changes of europium stannate pyrochlore, Eu2Sn2O7, have been investigated at high pressures with in situ Raman spectroscopy, photoluminescence (PL), and synchrotron X-ray diffraction (XRD) techniques. The XRD measurements suggest that a pressure-induced phase transition starts at 34.4 GPa. The PL spectrum from Eu3+ cations also suggests a phase transition above 36 GPa. XRD analysis shows that the unit cell of the cubic phase deviates from the equation of state at pressures above 23.8 GPa. This is due to the incorporation of water from the pressure medium in the structure at high pressures, which is confirmed by optical spectroscopy measurements.

  1. Pore water pressure response to small and large openings in argillaceous rocks

    International Nuclear Information System (INIS)

    Garitte, B.; Gens, A.; Vaunat, J.; Armand, G.; Conil, N.

    2012-01-01

    Document available in extended abstract form only. In the last decade an important amount of piezometers have been installed in the Bure Underground Rock Laboratory (URL) in the vicinity of ongoing works involving gallery excavations and drilling of boreholes and alveoles both in the major and minor stress directions. Relatively far field piezometers (placed one to four diameters from the excavation wall) showed a qualitatively consistent response at different scales. Here, we investigate whether the pore water pressure response around openings of different scales may be up-scaled. An attempt is made to find a common set of parameters that explains quantitatively the rock response at the different scales. The mechanisms underlying the pore water pressure response around an underground opening are twofold. The first class of mechanisms is usually associated with nearly undrained behaviour and the related pore water pressure changes are induced by the stress redistribution triggered by the creation of the tunnel opening causing a reorientation of the principal stresses and influenced by the initial stress anisotropy. These pore water pressure changes are closely linked to the mechanical constitutive law of the rock and to the damage zone around the opening. The second class of mechanisms is related to the drainage of excess pore water pressure relative to a state governed by the atmospheric water pressure condition prescribed at gallery wall and the water flow law, usually Darcy's. Strong anisotropy effects on the hydraulic response of Callovo-Oxfordian Clay can be observed with reference to Figure 1 that shows the pore pressure response to the drilling of a 150 mm-diameter borehole performed to install a heater for the TER thermal experiment. The borehole is aligned with the major horizontal principal stress. Therefore, in principle, the stress state should be approximately isotropic in a cross section of the borehole. As a matter of fact, however, a degree of

  2. Pressure of drinking water network on the Meyrin site to be boosted

    CERN Multimedia

    2004-01-01

    In the framework of the refurbishment of CERN's drinking water supply system, the final part of the network on the Meyrin site is to be connected to the pumping station operated by Services Industriels de Genève, bringing about a significant increase in the network pressure of up to 5 bar. This means that from January 2005 onwards, the water pressure in buildings will be increased from 2 - 4 bar to 7 - 9 bar. The TS Department will be checking and upgrading the drinking water supply equipment in toilets and washrooms. All users with devices connected to the water supply system are kindly requested to check that these are compatible with the new pressure levels. More information on the buildings affected, the new pressure levels and the dates on which the changes will come into effect can be found at: https://edms.cern.ch/document/525717/1 Should any equipment under your responsibility be incompatible with the future pressure levels, please contact the Technical Control Room on 72201.

  3. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  4. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  5. 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

  6. Effect of sulfur dioxide partial pressure on the reaction of iodine, sulfur dioxide and water

    International Nuclear Information System (INIS)

    Nakajima, Hayato; Imai, Yoshiyuki; Kasahara, Seiji; Kubo, Shinji; Onuki, Kaoru

    2007-01-01

    Effect of sulfur dioxide partial pressure on the reaction of iodine, sulfur dioxide and water, which is a unit reaction in the IS process for thermochemical hydrogen production, was studied experimentally at 323 K under iodine saturation. Quasi-equilibrium state was observed in the presence of sulfur dioxide gas at constant pressure. The composition of the poly-hydriodic acid solution formed was discussed assuming an ideal desulfurization by the reverse reaction of the Bunsen reaction. The value of HI/(HI+H 2 O) of the desulfurized solution was large at high sulfur dioxide pressure and reached the maximum of 15.7 ± 0.3 mol%. (author)

  7. Water Permeability of Pervious Concrete Is Dependent on the Applied Pressure and Testing Methods

    Directory of Open Access Journals (Sweden)

    Yinghong Qin

    2015-01-01

    Full Text Available Falling head method (FHM and constant head method (CHM are, respectively, used to test the water permeability of permeable concrete, using different water heads on the testing samples. The results indicate the apparent permeability of pervious concrete decreasing with the applied water head. The results also demonstrate the permeability measured from the FHM is lower than that from the CHM. The fundamental difference between the CHM and FHM is examined from the theory of fluid flowing through porous media. The testing results suggest that the water permeability of permeable concrete should be reported with the applied pressure and the associated testing method.

  8. The analysis of energy efficiency in water electrolysis under high temperature and high pressure

    Science.gov (United States)

    Hourng, L. W.; Tsai, T. T.; Lin, M. Y.

    2017-11-01

    This paper aims to analyze the energy efficiency of water electrolysis under high pressure and high temperature conditions. The effects of temperature and pressure on four different kinds of reaction mechanisms, namely, reversible voltage, activation polarization, ohmic polarization, and concentration polarization, are investigated in details. Results show that the ohmic and concentration over-potentials are increased as temperature is increased, however, the reversible and activation over-potentials are decreased as temperature is increased. Therefore, the net efficiency is enhanced as temperature is increased. The efficiency of water electrolysis at 350°C/100 bars is increased about 17%, compared with that at 80°C/1bar.

  9. Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

    International Nuclear Information System (INIS)

    Schumann, S.

    1984-01-01

    The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.) [de

  10. Franco-German nuclear cooperation: from the 'common product' to the first European pressurized water reactor

    International Nuclear Information System (INIS)

    Vignon, D.

    1999-01-01

    It has now been 10 years since Framatome and Siemens decided to collaborate on the design and sales of an advanced nuclear power plant (NPP) model based on pressurized water reactor (PWR) technology. Originally called the 'common product', this model was renamed the European pressurized water reactor when Electricite de France (EDF) and the German electric utilities joined this cooperative development effort in 1992. Since the beginning, this cooperation has been formalized in the framework of an agreement that led to the founding of a joint and equally owned subsidiary, Nucler Power International (NPI), which is reponsible for leading the development of the new model and later handling its export sales

  11. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

    Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

  12. 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

  13. Fatigue crack growth behavior of pressure vessel steels and submerged arc weldments in a high-temperature pressurized water environment

    International Nuclear Information System (INIS)

    Liaw, P.K.; Logsdon, W.A.; Begley, J.A.

    1989-01-01

    The fatigue crack growth rate (FCGR) properties of SA508 Cl 2a and SA533 Gr A Cl 2 pressure vessel steels and the corresponding automatic submerged arc weldments were developed in a high-temperature pressurized water (HPW) environment at 288 degrees C (550 degrees F) and 7.2 MPa (1044 psi) at load ratios of 0.20 and 0.50. The properties were generally conservative compared to American Society of Mechanical Engineers Section XI water environment reference curve. The growth rate of fatigue cracks in the base materials, however, was faster in the HPW environment than in a 288 degrees C (550 degrees F) base line air environment. The growth rate of fatigue cracks in the two submerged arc weldments was also accelerated in the HPW environment but to a lesser degree than that demonstrated by the base materials. In the air environment, fatigue striations were observed, independent of material and load ratio, while in the HPW environment, some intergranular facets were present. The greater environmental effect on crack growth rates displayed by the base materials compared the weldments attributed to a different sulfide composition and morphology

  14. Development of a higher capacity, lower pressure drop steam/water separator with reduced primary-to-secondary spacing

    International Nuclear Information System (INIS)

    Pruster, W.P.; Kidwell, J.H.; Eaton, A.M.; Wall, J.R.

    1985-01-01

    The goal of this development effort was to double the steam flow capacity of an existing module steam/water separator design without significantly increasing the pressure drop while simultaneously minimizing the vertical distance (spacing) between the primary and secondary separation stages. The development work included extensive air/water and steam/water testing. The steam/water tests were performed at a common pressure of 300 psia (2.1 MPa) with comparable water and steam flows

  15. Durable Suit Bladder with Improved Water Permeability for Pressure and Environment Suits

    Science.gov (United States)

    Bue, Grant C.; Kuznetz, Larry; Orndoff, Evelyne; Tang, Henry; Aitchison, Lindsay; Ross, Amy

    2009-01-01

    Water vapor permeability is shown to be useful in rejecting heat and managing moisture accumulation in launch-and-entry pressure suits. Currently this is accomplished through a porous Gortex layer in the Advanced Crew and Escape Suit (ACES) and in the baseline design of the Constellation Suit System Element (CSSE) Suit 1. Non-porous dense monolithic membranes (DMM) that are available offer potential improvements for water vapor permeability with reduced gas leak. Accordingly, three different pressure bladder materials were investigated for water vapor permeability and oxygen leak: ElasthaneTM 80A (thermoplastic polyether urethane) provided from stock polymer material and two custom thermoplastic polyether urethanes. Water vapor, carbon dioxide and oxygen permeability of the DMM's was measured in a 0.13 mm thick stand-alone layer, a 0.08 mm and 0.05 mm thick layer each bonded to two different nylon and polyester woven reinforcing materials. Additional water vapor permeability and mechanical compression measurements were made with the reinforced 0.05 mm thick layers, further bonded with a polyester wicking and overlaid with moistened polyester fleece thermal underwear .This simulated the pressure from a supine crew person. The 0.05 mm thick nylon reinforced sample with polyester wicking layer was further mechanically tested for wear and abrasion. Concepts for incorporating these materials in launch/entry and Extravehicular Activity pressure suits are presented.

  16. Water permeability evaluation of hollow cylindrical reinforced concrete structure by means of long-term water penetration test with pressure

    International Nuclear Information System (INIS)

    Fujiwara, Ai; Miura, Norihiko; Konishi, Kazuhiro; Tsuji, Yukikazu

    2005-01-01

    In order to evaluate initial permeability of large concrete structure, hollow cylindrical reinforced concrete structure, having 6 m in outer diameter, 6 m in height, 1 m in thickness, had been tested by means of 0.25 MPa of outside water pressure. As the results, although surface cracking and partial separation of joint had been observed at the inner side, no water permeation through concrete could be happened even after 5.5 years. After this test, concrete core specimen showed less water penetration within the depth of concrete cover of reinforcement. Thus it was verified that this concrete structure had very high water-tightness, and that the initial average water permeability was estimated to be about 1.6 x 10 -12 m/s. (author)

  17. Intelligent Pressure Management to Reduce Leakage in Urban Water Supply Networks, A Case Study of Sarafrazan District, Mashhad

    OpenAIRE

    Mohammad Soltani Asl; Mahmoud Faghfour Maghrebi

    2009-01-01

    Water losses are inevitable in urban water distribution systems. The two approaches adopted nowadays to combat this problem include management of hydraulic parameters such as pressure and leakage detection in the network. Intellitgent pressure management is a suitable technique for controlling leakage and reducing damages due to high operating pressures in a network. This paper aims to investigate the effects of pressure reduction on leakage. The EPANET 2.10 software is used to simulate the w...

  18. Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

    International Nuclear Information System (INIS)

    Li, D D; Jiang, J; Zhao, Z; Yi, W S; Lan, G

    2013-01-01

    We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system

  19. Research on the water hammer protection of the long distance water supply project with the combined action of the air vessel and over-pressure relief valve

    Science.gov (United States)

    Li, D. D.; Jiang, J.; Zhao, Z.; Yi, W. S.; Lan, G.

    2013-12-01

    We take a concrete pumping station as an example in this paper. Through the calculation of water hammer protection with a specific pumping station water supply project, and the analysis of the principle, mathematical models and boundary conditions of air vessel and over-pressure relief valve we show that the air vessel can protect the water conveyance system and reduce the transient pressure damage due to various causes. Over-pressure relief valve can effectively reduce the water hammer because the water column re-bridge suddenly stops the pump and prevents pipeline burst. The paper indicates that the combination set of air vessel and over-pressure relief valve can greatly reduce the quantity of the air valve and can eliminate the water hammer phenomenon in the pipeline system due to the vaporization and water column separation and re-bridge. The conclusion could provide a reference for the water hammer protection of long-distance water supply system.

  20. The influence of chemicals on water quality in a high pressure separation rig

    Energy Technology Data Exchange (ETDEWEB)

    Johnsen, Einar E.; Hemmingsen, Paal V.; Mediaas, Heidi; Svarstad, May Britt E.; Westvik, Arild

    2006-03-15

    In the research laboratory of Statoil at Rotvoll, Trondheim, a high pressure experimental rig used for separation and foaming studies has been developed. There have been several studies to ensure that the high pressure separation rig produces reliable and consistent results with regard to the water-in-oil and oil-in-water contents. The results are consistent with available field data and, just as important, consistent when changing variables like temperature, pressure drop and water cut. The results are also consistent when changing hydrodynamic variables like flow velocity and mixing point (using different choke valves) and when using oil with and without gas saturation. At equal experimental conditions, the high pressure separation rig is able to differentiate between separation characteristics of oil and water from different fields and from different wells at the same field. The high pressure separation and foam rig can be used from -10 deg C to 175 deg C and at pressures up to 200 bar. Crude oil and water are studied under relevant process conditions with respect to temperature, pressure, shear, water cut and separation time. In the present work the influence of chemicals on the oil and water quality has been studied. Chemicals have been mixed into the oil and/or water beforehand or added in situ (on-stream; simulated well stream). The amount of oil in the water after a given residence time in the separation cell has been measured. The results from the high pressure rig show that some demulsifiers, with their primary purpose of giving less water in oil, also have influence on the water quality. Improvement of water quality has been observed as well as no effect or aggravation. The experimental results have been compared to results from bottle tests at the field. The results from the bottle tests and from the laboratory are not corresponding, and only a full-scale field test can tell which of them are the correct results, if any. (Experience from corresponding

  1. Pressure Dependence of the Liquid-Liquid Phase Transition of Nanopore Water Doped Slightly with Hydroxylamine, and a Phase Behavior Predicted for Pure Water

    Science.gov (United States)

    Nagoe, Atsushi; Iwaki, Shinji; Oguni, Masaharu; Tôzaki, Ken-ichi

    2014-09-01

    Phase transition behaviors of confined pure water and confined water doped with a small amount of hydroxylamine (HA) with a mole fraction of xHA = 0.03 were examined by high-pressure differential thermal analyses at 0.1, 50, 100, and 150 MPa; the average diameters of silica pores used were 2.0 and 2.5 nm. A liquid-liquid phase transition (LLPT) of the confined HA-doped water was clearly observed and its pressurization effect could be evaluated, unlike in the experiments on undoped water. It was found that pressurization causes the transition temperature (Ttrs) to linearly decrease, indicating that the low-temperature phase has a lower density than the high-temperature one. Transition enthalpy (ΔtrsH) decreased steeply with increasing pressure. Considering the linear decrease in Ttrs with increasing pressure, the steep decrease in ΔtrsH indicates that the LLPT effect of the HA-doped water attenuates with pressure. We present a new scenario of the phase behavior concerning the LLPT of pure water based on the analogy from the behavior of slightly HA-doped water, where a liquid-liquid critical point (LLCP) and a coexistence line are located in a negative-pressure regime but not in a positive-pressure one. It is reasonably understood that doping a small amount of HA into water results in negative chemical pressurization and causes the LLPT to occur even at ambient pressure.

  2. Comparison between intraocular pressure spikes with water loading and postural change.

    Science.gov (United States)

    Chong, Calum Wk; Wang, Sarah B; Jain, Neeranjali S; Bank, Cassandra S; Singh, Ravjit; Bank, Allan; Francis, Ian C; Agar, Ashish

    2016-12-01

    To compare the agreement between peak intraocular pressures measured through the water drinking test and the supine test, in patients with primary open angle glaucoma. Consecutive, prospective, blinded. Twenty-one patients from the Glaucoma Unit, Prince of Wales Hospital, Sydney, Australia. For the supine test, intraocular pressure was recorded immediately after the patient had lain down and at 20 and 40 min. At the second evaluation, intraocular pressure was measured in each patient after drinking 10 mL/kg body weight of water for the water drinking test. Again, all patients had their intraocular pressure measured at 20 and 40 min (t = 20 and t = 40, respectively). Patients were excluded from the study if they had pre-existing cardiac, renal or pulmonary complications or had concurrent ocular disease or an anatomical abnormality (including angle recession, peripheral anterior synechiae and developmental anomalies of the angle) that may have influenced intraocular pressure. Bland-Altman analysis. Bland-Altman analysis indicated an overall excellent agreement in terms of mean difference between methods (1.0, -1.0 and -0.90 mmHg, at 0, 20 and 40 min, respectively). Further, with the exception of t = 40, all measured time points had 95% confidence intervals within 6.5 mmHg of their mean difference on the Bland-Altman plot. There was close agreement between the intraocular pressure values of the supine test and water drinking test. However, as the water drinking test may be uncomfortable and potentially hazardous, there is potential that the supine test may be a safer and more comfortable alternative. © 2016 Royal Australian and New Zealand College of Ophthalmologists.

  3. Characteristics of meter-scale surface electrical discharge propagating along water surface at atmospheric pressure

    Czech Academy of Sciences Publication Activity Database

    Hoffer, Petr; Sugiyama, Y.; Hosseini, S.H.R.; Akiyama, H.; Lukeš, Petr; Akiyama, M.

    2016-01-01

    Roč. 49, č. 41 (2016), č. článku 415202. ISSN 0022-3727 Institutional support: RVO:61389021 Keywords : water surface * spectroscopy * high-speed photography * pulsed plasma discharge * Atmospheric - pressure plasmas * electric discharges * liquids * water Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.588, year: 2016 http://iopscience.iop.org/article/10.1088/0022-3727/49/41/415202

  4. The European Pressurized Water Reactor. A safe and competitive solution for future energy needs

    International Nuclear Information System (INIS)

    Leverenz, R.; Gerhard, L.; Goebel, A.

    2004-01-01

    The European Pressurized Water Reactor - the EPR - is a PWR in the 1600 MW class. Its design is based on experience feedback from several thousand reactors x years of light water reactor operation worldwide, primarily those incorporating the most recent technologies: the French N4 and the German KONVOI reactors. It is an evolutionary design that ensures continuity in the mastery of PWR technology, minimizing the risk for the customer. (author)

  5. Characteristics of meter-scale surface electrical discharge propagating along water surface at atmospheric pressure

    Czech Academy of Sciences Publication Activity Database

    Hoffer, Petr; Sugiyama, Y.; Hosseini, S.H.R.; Akiyama, H.; Lukeš, Petr; Akiyama, M.

    2016-01-01

    Roč. 49, č. 41 (2016), č. článku 415202. ISSN 0022-3727 Institutional support: RVO:61389021 Keywords : water surface * spectroscopy * high-speed photography * pulsed plasma discharge * Atmospheric-pressure plasmas * electric discharges * liquids * water Subject RIV: BL - Plasma and Gas Discharge Physics Impact factor: 2.588, year: 2016 http://iopscience.iop.org/article/10.1088/0022-3727/49/41/415202

  6. Normal coordinate treatment of liquid water and calculation of vapor pressure isotope effects

    International Nuclear Information System (INIS)

    Gellai, B.; Van Hook, W.A.

    1983-01-01

    A vibrational analysis of liquid water is reported, assuming a completely hydrogen-bonded network with continuously varying strengths of the hydrogen bonds. Frequency distribution calculations are made for intramolecular stretching and bending modes and for the intramolecular frequency region. The calculated distributions are compared with the experimental spectroscopic ones. As another test, vapor pressure isotope effects are calculated from the theoretical distributions for some isotopic water molecules. Results are compared with those of other authors obtained from a mixture model. (author)

  7. Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

    International Nuclear Information System (INIS)

    Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

    2002-01-01

    A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

  8. Effect of water phase transition on dynamic ruptures with thermal pressurization: Numerical simulations with changes in physical properties of water

    Science.gov (United States)

    Urata, Yumi; Kuge, Keiko; Kase, Yuko

    2015-02-01

    Phase transitions of pore water have never been considered in dynamic rupture simulations with thermal pressurization (TP), although they may control TP. From numerical simulations of dynamic rupture propagation including TP, in the absence of any water phase transition process, we predict that frictional heating and TP are likely to change liquid pore water into supercritical water for a strike-slip fault under depth-dependent stress. This phase transition causes changes of a few orders of magnitude in viscosity, compressibility, and thermal expansion among physical properties of water, thus affecting the diffusion of pore pressure. Accordingly, we perform numerical simulations of dynamic ruptures with TP, considering physical properties that vary with the pressure and temperature of pore water on a fault. To observe the effects of the phase transition, we assume uniform initial stress and no fault-normal variations in fluid density and viscosity. The results suggest that the varying physical properties decrease the total slip in cases with high stress at depth and small shear zone thickness. When fault-normal variations in fluid density and viscosity are included in the diffusion equation, they activate TP much earlier than the phase transition. As a consequence, the total slip becomes greater than that in the case with constant physical properties, eradicating the phase transition effect. Varying physical properties do not affect the rupture velocity, irrespective of the fault-normal variations. Thus, the phase transition of pore water has little effect on dynamic ruptures. Fault-normal variations in fluid density and viscosity may play a more significant role.

  9. 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.

  10. Effects of High Hydrostatic Pressure on Water Absorption of Adzuki Beans

    Science.gov (United States)

    Ueno, Shigeaki; Shigematsu, Toru; Karo, Mineko; Hayashi, Mayumi; Fujii, Tomoyuki

    2015-01-01

    The effect of high hydrostatic pressure (HHP) treatment on dried soybean, adzuki bean, and kintoki kidney bean, which are low-moisture-content cellular biological materials, was investigated from the viewpoint of water absorption. The samples were vacuum-packed with distilled water and pressurized at 200 MPa and 25 °C for 10 min. After the HHP treatment, time courses of the moisture contents of the samples were measured, and the dimensionless moisture contents were estimated. Water absorption in the case of soybean could be fitted well by a simple water diffusion model. High pressures were found to have negligible effects on water absorption into the cotyledon of soybean and kintoki kidney bean. A non-linear least square method based on the Weibull equation was applied for the adzuki beans, and the effective water diffusion coefficient was found to increase significantly from 8.6 × 10−13 to 6.7 × 10−10 m2/s after HHP treatment. Approximately 30% of the testa of the adzuki bean was damaged upon HHP treatment, which was comparable to the surface area of the testa in the partially peeled adzuki bean sample. Thus, HHP was confirmed to promote mass transfer to the cotyledon of legumes with a tight testa. PMID:28231195

  11. Effects of High Hydrostatic Pressure on Water Absorption of Adzuki Beans

    Directory of Open Access Journals (Sweden)

    Shigeaki Ueno

    2015-05-01

    Full Text Available The effect of high hydrostatic pressure (HHP treatment on dried soybean, adzuki bean, and kintoki kidney bean, which are low-moisture-content cellular biological materials, was investigated from the viewpoint of water absorption. The samples were vacuum-packed with distilled water and pressurized at 200 MPa and 25 °C for 10 min. After the HHP treatment, time courses of the moisture contents of the samples were measured, and the dimensionless moisture contents were estimated. Water absorption in the case of soybean could be fitted well by a simple water diffusion model. High pressures were found to have negligible effects on water absorption into the cotyledon of soybean and kintoki kidney bean. A non-linear least square method based on the Weibull equation was applied for the adzuki beans, and the effective water diffusion coefficient was found to increase significantly from 8.6 × 10−13 to 6.7 × 10−10 m2/s after HHP treatment. Approximately 30% of the testa of the adzuki bean was damaged upon HHP treatment, which was comparable to the surface area of the testa in the partially peeled adzuki bean sample. Thus, HHP was confirmed to promote mass transfer to the cotyledon of legumes with a tight testa.

  12. Negative Pressures and the First Water Siphon Taller than 10.33 Meters.

    Directory of Open Access Journals (Sweden)

    Francisco Vera

    Full Text Available A siphon is a device that is used to drain a container, with water rising inside a hose in the form of an inverted U and then going down towards a discharge point placed below the initial water level. The siphon is the first of a number of inventions of the ancients documented about 2.000 years ago by Hero of Alexandria in his treatise Pneumatics, and although the explanation given by Hero was essentially correct, there is nowadays a controversy about the underlying mechanism that explains the working of this device. Discussions concerning the physics of a siphon usually refer to concepts like absolute negative pressures, the strength of liquid's cohesion and the possibility of a siphon working in vacuum or in the presence of bubbles. Torricelli understood the working principle of the barometer and the impossibility of pumping water out of wells deeper than 10.33 m. Following Torricelli's ideas it would also not be possible to build a siphon that drives pure water to ascend higher than 10.33 m. In this work, we report the first siphon that drives water (with surfactant to ascend higher than the Torricellian limit. Motivated by the rising of sap in trees, we built a 15.4 m siphon that shows that absolute negative pressures are not prohibited, that cohesion plays an important role in transmitting forces through a fluid, and that surfactants can help to the transport of water in a metastable regime of negative pressures.

  13. 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

  14. Pressure effect on the conformational equilibrium of [Leu]{sup 5}-enkephalin in water

    Energy Technology Data Exchange (ETDEWEB)

    Shimizu, A [Department of Environmental Engineering for Symbiosis, Soka University, 1-326 Tangi-cho, Hachioji, Tokyo, 192-8577 (Japan); Takekiyo, T; Yoshimura, Y [Department of Applied Chemistry, National Defence Academy, 1-10-20 Hashirimizu, Yokosuka, Kanagawa, 239-8686 (Japan); Kato, M; Taniguchi, Y, E-mail: shimizu@soka.ac.j, E-mail: take214@nda.ac.j [Department of Applied Chemistry, Ritsumeikan University, 1-1-1, Nojihigashi, Kusatsu, Shiga, 525-8577 (Japan)

    2010-03-01

    The conformational stability of [Leu]{sup 5}-enkephalin,Tyr-Gly-Gly-Phe-Leu, in water have been investigated under high pressure by FTIR spectroscopy. Three peaks at 1638, 1650, and 1680 cm{sup -1} were determined by second derivative FTIR spectra in the amide I' region of [Leu]{sup 5}-enkephalin. The peaks at 1637 and 1680 cm{sup -1} are assigned to the {beta}-strand and turn structures, respectively. These peaks mean that [Leu]{sup 5}-enkephalin takes a {beta}-hairpin-like structure in water. Moreover, the absorbance at 1638 cm{sup -1} increases with increasing pressure, and this change shows a sigmoidal curve. Thus, we concluded that [Leu]{sup 5}-enkephalin has the {beta}-hairpin-like and disordered structures in water. From the FTIR profile at high pressures, the {beta}-hairpin-like structure of [Leu]{sup 5}-enkephalin is stabilized by a high pressures. Our result shows that the folded structures such as {alpha}-helix and {beta}-hairpin structures of short peptide such as [Leu]{sup 5}-enkephalin are stabilized at high pressures.

  15. Water cycle and its management for plant habitats at reduced pressures

    Science.gov (United States)

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

    2004-01-01

    Experimental and mathematical models were developed for describing and testing temperature and humidity parameters for plant production in bioregenerative life support systems. A factor was included for analyzing systems operating at low (10-101.3 kPa) pressure to reduce gas leakage and structural mass (e.g., inflatable greenhouses for space application). The expected close relationship between temperature and relative humidity was observed, along with the importance of heat exchanger coil temperature and air circulation rate. The presence of plants in closed habitats results in increased water flux through the system. Changes in pressure affect gas diffusion rates and surface boundary layers, and change convective transfer capabilities and water evaporation rates. A consistent observation from studies with plants at reduced pressures is increased evapotranspiration rates, even at constant vapor pressure deficits. This suggests that plant water status is a critical factor for managing low-pressure production systems. The approach suggested should help space mission planners design artificial environments in closed habitats.

  16. Use of inexpensive pressure transducers for measuring water levels in wells

    Science.gov (United States)

    Keeland, B.D.; Dowd, J.F.; Hardegree, W.S.

    1997-01-01

    Frequent measurement of below ground water levels at multiple locations is an important component of many wetland ecosystem studies. These measurements, however, are usually time consuming, labor intensive, and expensive. This paper describes a water-level sensor that is inexpensive and easy to construct. The sensor is placed below the expected low water level in a shallow well and, when connected to a datalogger, uses a pressure transducer to detect groundwater or surface water elevations. Details of pressure transducer theory, sensor construction, calibration, and examples of field installations are presented. Although the transducers must be individually calibrated, the sensors have a linear response to changing water levels (r2 ??? .999). Measurement errors resulting from temperature fluctuations are shown to be about 4 cm over a 35??C temperature range, but are minimal when the sensors are installed in groundwater wells where temperatures are less variable. Greater accuracy may be obtained by incorporating water temperature data into the initial calibration (0.14 cm error over a 35??C temperature range). Examples of the utility of these sensors in studies of groundwater/surface water interactions and the effects of water level fluctuations on tree growth are provided. ?? 1997 Kluwer Academic Publishers.

  17. Fissures in rock under water pressure, implications on stability : 3 unusual cases

    Energy Technology Data Exchange (ETDEWEB)

    Helwig, P.C. [Helwig Hydrotechnique Ltd., St. John' s, NL (Canada)

    2006-07-01

    The presence of water in rock joints has important implications on the stability of rock foundations. Appropriate analyses are needed to assess the stability of dam foundations, abutments and rock walls. This paper presented 3 case studies in which the freezing of seepage flows in rock joints and transient pressure in rock walls were investigated: (1) an assessment of the effects of freezing water in rock joints at the Paradise River arch dam in Newfoundland; (2) stability of rock walls in the unlined power tunnel of the Cat Arm hydroelectric development in Newfoundland due to transient pressures; and (3) assessing the influence of fluctuating water pressures in a stilling basin excavated in rock. After an investigation of the Paradise River canyon walls, a drainage system comprised of peripheral drain holes was drilled into the foundation and walls at regular intervals to intercept seepage flows and to relieve uplift water pressures. However, no special treatment was found for the potential freezing of water in the joints of the dam walls and foundation. The Cat Arm tunnel was used to study the depth at which significant transient pressures can be used to assess rock stability. Rock properties, typical fracture apertures and spacing were assumed and joint deformability was taken into account. An axisymmetric solution was obtained by considering the continuity and flow through an annular element of the rock wall. A finite difference method was used to solve the resulting nonlinear differential equation. In the final case study, blast-damaged rock was undermining the toe of a spillway. A cut-off wall was constructed as a series of drilled, cast-in-place concrete caisson piles. Criteria for the design included extending the cut-off wall to a depth beyond the effects of fluctuating surface pressures. Depth was assessed by considering the transient behaviour of water penetrating a sub-vertical joint subject exposed to fluctuating pressures. Results of the calculations

  18. Formation of genotoxic compounds by medium pressure ultra violet treatment of nitrate rich water

    NARCIS (Netherlands)

    Martijn, A.J.; Boersma, M.G.; Vervoort, Jacques; Rietjens, I.; Kruithof, J.C.

    2014-01-01

    Genotoxic compounds were produced by full-scale medium pressure (MP) ultraviolet hydrogen peroxide (UV/H2O2) treatment of nitrate-rich pretreated surface water. It was hypothesized that this formation was caused by the reaction of nitrate photolysis intermediates with natural organic matter (NOM).

  19. Anode pattern formation in atmospheric pressure air glow discharges with water anode

    NARCIS (Netherlands)

    Verreycken, T.; Bruggeman, P.J.; Leys, C.

    2009-01-01

    Pattern formation in the anode layer at a water electrode in atmospheric pressure glow discharges in air is studied. With increasing current a sequence of different anode spot structures occurs from a constricted homogeneous spot in the case of small currents to a pattern consisting of small

  20. An investigation of rock fall and pore water pressure using LIDAR in Highway 63 rock cuts.

    Science.gov (United States)

    2014-07-01

    The purpose of this research work is compare LIDAR scanning measurements of rock fall with the natural changes in groundwater level to determining the effect of water pressures (levels) on rock fall. To collect the information of rock cut volume chan...

  1. Pore-water pressures associated with clogging of soil pipes: Numerical analysis of laboratory experiments

    Science.gov (United States)

    Clogging of soil pipes due to excessive internal erosion has been hypothesized to cause extreme erosion events such as landslides, debris flows, and gullies, but confirmation of this phenomenon has been lacking. Laboratory and field measurements have failed to measure pore water pressures within pip...

  2. Pressurized hot water extraction of proteins from Sambucus nigra L. branches

    Czech Academy of Sciences Publication Activity Database

    Šalplachta, Jiří; Hohnová, Barbora

    2017-01-01

    Roč. 108, DEC (2017), s. 312-315 ISSN 0926-6690 Grant - others:GA AV ČR(CZ) R200311521 Institutional support: RVO:68081715 Keywords : elderberry * pressurized hot water extraction * proteins Subject RIV: CB - Analytical Chemistry, Separation OBOR OECD: Analytical chemistry Impact factor: 3.181, year: 2016

  3. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... comments were received. A companion guide, DG-1277, ``Initial Test Program of Emergency Core Cooling... NUCLEAR REGULATORY COMMISSION [NRC-2011-0129] Preoperational Testing of Emergency Core Cooling... (RG), 1.79, ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors...

  4. Adaptation of high pressure water jets with abrasives for nuclear installations dismantling

    International Nuclear Information System (INIS)

    Rouviere, R.; Pinault, M.; Gasc, B.; Guiadeur, R.; Pilot, M.

    1989-01-01

    This report presents the work realized for adjust the cutting technology with high pressure water jet with abrasives for nuclear installation dismantling. It has necessited the conception and the adjustement of a remote tool and the realization of cutting tests with waste produce analysis. This technic can be ameliorated with better viewing systems and better fog suction systems

  5. 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

  6. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  7. On the effect of pressure and carrier gas on homogeneous water nucleation

    Czech Academy of Sciences Publication Activity Database

    Fransen, M. A. L. J.; Hrubý, Jan; Smeulders, D. M. J.; Dongen, M. E. H.

    2015-01-01

    Roč. 142, č. 16 (2015), 164307_1-164307_7 ISSN 0021-9606 R&D Projects: GA ČR(CZ) GAP101/11/1593 Institutional support: RVO:61388998 Keywords : homogeneous nucleation * water * pressure effect Subject RIV: BJ - Thermodynamics Impact factor: 2.894, year: 2015 http://dx.doi.org/10.1063/1.4919249

  8. 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)

  9. An improved film evaporation correlation for saline water at sub-atmospheric pressures

    KAUST Repository

    Shahzada, Muhammad Wakil

    2011-10-03

    This paper presents an investigation of heat transfer correlation in a falling-film evaporator working with saline water at sub-atmospheric pressures. The experiments are conducted at different salinity levels ranging from 15000 to 90000 ppm, and the pressures were maintained between 0.92 to 2.81 kPa (corresponds to saturation temperatures of 5.9 – 23 0C). The effect of salinity, saturation pressures and chilled water temperatures on the heat transfer coefficient are accounted in the modified film evaporation correlations. The results are fitted to the Han & Fletcher\\'s and Chun & Seban\\'s falling-film correlations which are used in desalination industry. We modify the said correlations by adding salinity and saturation temperature corrections with respective indices to give a better agreement to our measured data.

  10. An improved film evaporation correlation for saline water at sub-atmospheric pressures

    KAUST Repository

    Shahzada, Muhammad Wakil; Ng, Kim Choon; Thu, Kyaw; Myat, Aung; Gee, Chun Won

    2011-01-01

    This paper presents an investigation of heat transfer correlation in a falling-film evaporator working with saline water at sub-atmospheric pressures. The experiments are conducted at different salinity levels ranging from 15000 to 90000 ppm, and the pressures were maintained between 0.92 to 2.81 kPa (corresponds to saturation temperatures of 5.9 – 23 0C). The effect of salinity, saturation pressures and chilled water temperatures on the heat transfer coefficient are accounted in the modified film evaporation correlations. The results are fitted to the Han & Fletcher's and Chun & Seban's falling-film correlations which are used in desalination industry. We modify the said correlations by adding salinity and saturation temperature corrections with respective indices to give a better agreement to our measured data.

  11. Adiabatic pressure dependence of the 2.7 and 1.9 micron water vapor bands

    Science.gov (United States)

    Mathai, C. V.; Walls, W. L.; Broersma, S.

    1977-01-01

    An acoustic excitation technique is used to determine the adiabatic pressure derivative of the spectral absorptance of the 2.7 and 1.9 micron water vapor bands, and the 3.5 micron HCl band. The dependence of this derivative on thermodynamic parameters such as temperature, concentration, and pressure is evaluated. A cross-flow water vapor system is used to measure spectral absorptance. Taking F as the ratio of nonrigid to rotor line strengths, it is found that an F factor correction is needed for the 2.7 micron band. The F factor for the 1.9 micron band is also determined. In the wings of each band a wavelength can be found where the concentration dependence is predominant. Farther out in the wings a local maximum occurs for the temperature derivative. It is suggested that the pressure derivative is significant in the core of the band.

  12. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  13. Cavitation inception in nozzle-plate and wire mesh pressure droppers in water and sodium

    International Nuclear Information System (INIS)

    Collinson, A.E.

    1976-01-01

    Cavitation tests on multi-hole nozzle plates and wire meshes approximately 100mm diameter in water at 20 deg C and sodium at 300 deg C are described. These pressure dropping elements were mounted in recirculating loops where cavitation was induced by gradually lowering the back-ground pressure at constant flow. Cavitation was detected acoustically using wall mounted piezoelectric microphones, the signal being displayed on a ratemeter recording individual cavitation events. For nozzle plates, cavitation started intermittently as the pressure was lowered, the noise level suddenly increasing at a critical cavitation number sigma. For meshes the intermittent region was absent. Values of sigma for nozzles and meshes were similar in water and sodium for the conditions prevailing during the tests. It was apparent that cavitation took place on the axes of vortices both in the free stream and close to nozzle curved surfaces

  14. Sensitivity of total stress to changes in externally applied water pressure in KBS-3 buffer bentonite

    International Nuclear Information System (INIS)

    Harrington, J.F.; Birchall, D.J.

    2007-04-01

    In the current Swedish repository design concept, composite copper and steel canisters containing spent nuclear fuel will be placed in large diameter disposal boreholes drilled into the floor of the repository tunnels. The space around each canister will be filled with pre-compacted bentonite which over time will draw in the surrounding ground water and swell, closing up any construction joints. However, for the purposes of performance assessment, it is necessary to consider the effect of glacial loading of a future repository and its impact on the mechanical behaviour of the bentonite, in particular, the sensitivity of total stress to changes in porewater pressure (backpressure). Two experimental histories have been undertaken using a custom-designed constant volume and radial flow (CVRF) apparatus. In both tests backpressure was varied in a number of incremental and decremental cycles while total stress, porewater pressure and volumetric flow rate were continuously monitored. The swelling pressure of the buffer clay at dry densities of 1.8 Mg/m 3 and 1.61 Mg/m 3 was determined to be around 5.5 MPa and 7.2 MPa respectively. For initial ascending porewater pressure histories the average proportionality factor α ranged from 0.86 and 0.92. Data exhibited a general trend of increasing α with increasing backpressure. In test Mx80-11 this was supported by analysis of the water inflow data which indicated a reduction in system compressibility. Asymptotic values of porewater pressure within the clay are in good agreement with externally applied backpressure values. Inspection of data provides no evidence for the development of hydraulic thresholds within the clay, subject to the boundary conditions of this test geometry. Analysis of the stress data demonstrates significant hysteresis between ascending and descending porewater pressure histories. The amount of hysteresis appears to be linked to the magnitude of the backpressure applied to the specimen, suggesting some

  15. Sensitivity of total stress to changes in externally applied water pressure in KBS-3 buffer bentonite

    Energy Technology Data Exchange (ETDEWEB)

    Harrington, J.F.; Birchall, D.J. [British Geological Survey, Chemical and Biological Hazards Programme, Kingsley Dunham Centre (United Kingdom)

    2007-04-15

    In the current Swedish repository design concept, composite copper and steel canisters containing spent nuclear fuel will be placed in large diameter disposal boreholes drilled into the floor of the repository tunnels. The space around each canister will be filled with pre-compacted bentonite which over time will draw in the surrounding ground water and swell, closing up any construction joints. However, for the purposes of performance assessment, it is necessary to consider the effect of glacial loading of a future repository and its impact on the mechanical behaviour of the bentonite, in particular, the sensitivity of total stress to changes in porewater pressure (backpressure). Two experimental histories have been undertaken using a custom-designed constant volume and radial flow (CVRF) apparatus. In both tests backpressure was varied in a number of incremental and decremental cycles while total stress, porewater pressure and volumetric flow rate were continuously monitored. The swelling pressure of the buffer clay at dry densities of 1.8 Mg/m{sup 3} and 1.61 Mg/m{sup 3} was determined to be around 5.5 MPa and 7.2 MPa respectively. For initial ascending porewater pressure histories the average proportionality factor {alpha} ranged from 0.86 and 0.92. Data exhibited a general trend of increasing {alpha} with increasing backpressure. In test Mx80-11 this was supported by analysis of the water inflow data which indicated a reduction in system compressibility. Asymptotic values of porewater pressure within the clay are in good agreement with externally applied backpressure values. Inspection of data provides no evidence for the development of hydraulic thresholds within the clay, subject to the boundary conditions of this test geometry. Analysis of the stress data demonstrates significant hysteresis between ascending and descending porewater pressure histories. The amount of hysteresis appears to be linked to the magnitude of the backpressure applied to the specimen

  16. The effect of drinking water salinity on blood pressure in young adults of coastal Bangladesh.

    Science.gov (United States)

    Talukder, Mohammad Radwanur Rahman; Rutherford, Shannon; Phung, Dung; Islam, Mohammad Zahirul; Chu, Cordia

    2016-07-01

    More than 35 million people in coastal Bangladesh are vulnerable to increasing freshwater salinization. This will continue to affect more people and to a greater extent as climate change projections are realised in this area in the future. However the evidence for health effects of consuming high salinity water is limited. This research examined the association between drinking water salinity and blood pressure in young adults in coastal Bangladesh. We conducted a cross-sectional study during May-June 2014 in a rural coastal sub-district of Bangladesh. Data on blood pressure (BP) and salinity of potable water sources was collected from 253 participants aged 19-25 years. A linear regression method was used to examine the association between water salinity exposure categories and systolic BP (SBP) and diastolic BP (DBP) level. Sixty five percent of the study population were exposed to highly saline drinking water above the Bangladesh standard (600 mg/L and above). Multivariable linear regression analyses identified that compared to the low water salinity exposure category (water salinity category (>600 mg/L), had statistically significantly higher SBP (B 3.46, 95% CI 0.75, 6.17; p = 0.01) and DBP (B 2.77, 95% CI 0.31, 5.24; p = 0.03). Our research shows that elevated salinity in drinking water is associated with higher BP in young coastal populations. Blood pressure is an important risk factor of hypertension and cardiovascular diseases. Given the extent of salinization of freshwater in many low-lying countries including in Bangladesh, and the likely exacerbation related to climate change-induced sea level rise, implementation of preventative strategies through dietary interventions along with promotion of low saline drinking water must be a priority in these settings. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Determine spray droplets on water sensitive paper (WSP) for low pressure deflector nozzle using image J

    Science.gov (United States)

    Sies, M. F.; Madzlan, N. F.; Asmuin, N.; Sadikin, A.; Zakaria, H.

    2017-09-01

    In this study, determine of spray droplets size (SMD) using water sensitive paper (WSP) at low fluid pressure with deflector nozzle or tangential flow nozzle model Delavan AL75 and New Design Nozzle with two different type of swirl (ND2.5 A1.0 & ND2.5 B1.0). These three deflected flat sprays have used at different liquid mixing ratio. These liquid mixture ratios are pure water, 10% of lime juice + 90% of water (L10W90) and 30% of lime juice + 70% of water (L30W70). WSP is used to collect the spray droplets from nozzles. The operational liquid pressure of each nozzle is 3 bar, while air operational pressures are 3 bar and 6 bar. Then, the WSP were scanned using scanner then it was analyzed using ImageJ software. ImageJ can be used for determining the diameter of droplets size on the WSP. As the results from an experiment, the AL75 nozzle recorded the lowest Sauter mean diameter which is 193.69μm at 6 bar of pressurized air while ND2.5 A1.0 recorded the highest Sauter mean diameter which is 353.61µm at 3 bar of pressurized air. Summary from the experiment shows that the higher of droplet size is because of the lower air pressure (3 Bar). Then, increasing of liquid viscosity also increase the SMD. The orifice diameter for New Design nozzle (ND-2.5) is smaller than AL75, which are 2.5mm and 2.8mm respectively. The different nozzle design also gives effect the SMD. WSP is an alternative method to determine SMD for spray droplets with the low cost if compared to Phase Doppler Anemometry (PDA).

  18. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, K. [Kobe Univ. of Mercantile Marine (Japan); Shiotsu, M.; Sakurai, A. [Kyoto Univ. (Japan)

    1995-09-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.

  19. Toxic pressure of herbicides on microalgae in Dutch estuarine and coastal waters

    Science.gov (United States)

    Booij, Petra; Sjollema, Sascha B.; van der Geest, Harm G.; Leonards, Pim E. G.; Lamoree, Marja H.; de Voogt, W. Pim; Admiraal, Wim; Laane, Remi W. P. M.; Vethaak, A. Dick

    2015-08-01

    For several decades now, there has been an increase in the sources and types of chemicals in estuarine and coastal waters as a consequence of anthropogenic activities. This has led to considerable concern about the effects of these chemicals on the marine food chain. The fact is that estuarine and coastal waters are the most productive ecosystems with high primary production by microalgae. The toxic pressure of specific phytotoxic chemicals now poses a major threat to these ecosystems. In a previous study, six herbicides (atrazine, diuron, irgarol, isoproturon, terbutryn and terbutylazine) were identified as the main contaminants affecting photosynthesis in marine microalgae. The purpose of this study is to investigate the toxic pressure of these herbicides in the Dutch estuarine and coastal waters in relation to the effective photosystem II efficiency (ΦPSII) in microalgae. Temporal and spatial variations in the concentrations of these herbicides were analyzed based on monitoring data. Additionally, a field study was carried out in which chemical analysis of water was performed and also a toxicity assessment using the Pulse Amplitude Modulation (PAM) fluorometry assay that measures ΦPSII. The toxic pressure on ΦPSII in microalgae has decreased with 55-82% from 2003 to 2012, with the Western Scheldt estuary showing the highest toxic pressure. By combining toxicity data from the PAM assay with chemical analysis of herbicide concentrations, we have identified diuron and terbutylazine as the main contributors to the toxic pressure on microalgae. Although direct effects are not expected, the toxic pressure is close to the 10% effect level in the PAM assay. A compliance check with the current environmental legislation of the European Union revealed that the quality standards are not sufficient to protect marine microalgae.

  20. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    International Nuclear Information System (INIS)

    Fukuda, K.; Shiotsu, M.; Sakurai, A.

    1995-01-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q max , on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q o e t/T , with periods, τ, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q max . Two main mechanisms of q max exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q max for long period range belonging to the former mechanism becomes longer and the q max mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q max for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling

  1. Fundamental R and D program on water chemistry of supercritical pressure water under radiation field

    International Nuclear Information System (INIS)

    Katsumura, Yosuke; Kiuchi, Kiyoshi; Wada, Yoichi; Yotsuyanagi, Tadasu

    2003-01-01

    In a supercritical water-cooled reactor, property of water changes significantly around the critical point. It is expected that irradiation and change of water property will affect the chemistry and material corrosion. Deep understanding of interactions between supercritical water and materials under irradiation is important. However, comprehensive data on radiolysis, kinetics, corrosion and thermodynamics have not been obtained due to the severe experimental condition. To get such data by experiments and computer simulations, a national program funded by Ministry of Education, Culture, Sports, Science and Technology (MEXT) has been started since December 2002. (author)

  2. Application of pressure assisted forward osmosis for water purification and reuse of reverse osmosis concentrate from a water reclamation plant

    KAUST Repository

    Jamil, Shazad

    2016-07-26

    The use of forward osmosis (FO) is growing among the researchers for water desalination and wastewater treatment due to use of natural osmotic pressure of draw solute. In this study pressure assisted forward osmosis (PAFO) was used instead of FO to increase the water production rate. In this study a low concentration of draw solution (0.25 M KCl) was applied so that diluted KCl after PAFO operation can directly be used for fertigation. The performance of PAFO was investigated for the treatment of reverse osmosis concentrate (ROC) from a water reclamation plant. The water production in PAFO was increased by 9% and 29% at applied pressure of 2 and 4 bars, respectively, to feed side based on 90 h of experiments. Granular activated carbon (GAC) pretreatment and HCl softening were used to reduce organic fouling and scaling prior to application of PAFO. It reduced total organic carbon (TOC) and total inorganic carbon (TIC) by around 90% and 85%, respectively from untreated ROC. Subsequently, this led to an increase in permeate flux. In addition, GAC pretreatment adsorbed 12 out of 14 organic micropollutants tested from ROC to below detection limit. This application enabled to minimise the ROC volume with a sustainable operation and produced high quality and safe water for discharge or reuse. The draw solution (0.25 M KCl) used in this study was diluted to 0.14 M KCl, which is a suitable concentration (10 kg/m3) for fertigation, due to water transport from feed solution. © 2016 Elsevier B.V.

  3. Passive containment cooling system with drywell pressure regulation for boiling water reactor

    Science.gov (United States)

    Hill, P.R.

    1994-12-27

    A boiling water reactor is described having a regulating valve for placing the wetwell in flow communication with an intake duct of the passive containment cooling system. This subsystem can be adjusted to maintain the drywell pressure at (or slightly below or above) wetwell pressure after the initial reactor blowdown transient is over. This addition to the PCCS design has the benefit of eliminating or minimizing steam leakage from the drywell to the wetwell in the longer-term post-LOCA time period and also minimizes the temperature difference between drywell and wetwell. This in turn reduces the rate of long-term pressure buildup of the containment, thereby extending the time to reach the design pressure limit. 4 figures.

  4. Capillary pressure at irregularly shaped pore throats: Implications for water retention characteristics

    Science.gov (United States)

    Suh, Hyoung Suk; Kang, Dong Hun; Jang, Jaewon; Kim, Kwang Yeom; Yun, Tae Sup

    2017-12-01

    The random shapes of pore throats in geomaterials hinder accurate estimation of capillary pressure, and conventional pore network models that simply use the Young-Laplace equation assuming circular pore throats overestimate the capillary pressure. As a solution to this problem that does not complicate the pore network model or slow its implementation, we propose a new morphological analysis method to correlate the capillary pressure at an irregular pore channel with its cross-sectional geometry using lattice Boltzmann (LB) simulation and Mayer and Stowe-Princen theory. Geometry-based shape factors for pore throats are shown here to correlate strongly with the capillary pressure obtained by LB simulation. Water retention curves obtained by incorporating the morphological calibration into conventional pore network simulation and their correlative scheme agree well with experimental data. The suggested method is relevant to pore-scale processes such as geological CO2 sequestration, methane bubbling from wetlands, and enhanced carbon recovery.

  5. 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

  6. 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.

  7. Effect of renal venous pressure elevation on tubular sodium and water reabsorption in the dog kidney

    DEFF Research Database (Denmark)

    Abildgaard, U; Amtorp, O; Holstein-Rathlou, N H

    1988-01-01

    of [51Cr]EDTA was used as a measure of the rate of glomerular filtration (GFR). GFR, urinary excretion rates of sodium and water, and lithium clearance were used for assessing the absolute and fractional reabsorption rates of sodium and water in the proximal as well as in more distal segments......This study was performed in order to quantify the effects of renal venous pressure (RVP) elevation on absolute and fractional reabsorption rates of sodium and water in proximal and distal segments of the nephron in dog kidneys. Renal blood flow (RBF) was measured electromagnetically. Clearance...

  8. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  9. The effect of drinking water salinity on blood pressure in young adults of coastal Bangladesh

    International Nuclear Information System (INIS)

    Talukder, Mohammad Radwanur Rahman; Rutherford, Shannon; Phung, Dung; Islam, Mohammad Zahirul; Chu, Cordia

    2016-01-01

    More than 35 million people in coastal Bangladesh are vulnerable to increasing freshwater salinization. This will continue to affect more people and to a greater extent as climate change projections are realised in this area in the future. However the evidence for health effects of consuming high salinity water is limited. This research examined the association between drinking water salinity and blood pressure in young adults in coastal Bangladesh. We conducted a cross-sectional study during May-June 2014 in a rural coastal sub-district of Bangladesh. Data on blood pressure (BP) and salinity of potable water sources was collected from 253 participants aged 19–25 years. A linear regression method was used to examine the association between water salinity exposure categories and systolic BP (SBP) and diastolic BP (DBP) level. Sixty five percent of the study population were exposed to highly saline drinking water above the Bangladesh standard (600 mg/L and above). Multivariable linear regression analyses identified that compared to the low water salinity exposure category (<600 mg/L), those in the high water salinity category (>600 mg/L), had statistically significantly higher SBP (B 3.46, 95% CI 0.75, 6.17; p = 0.01) and DBP (B 2.77, 95% CI 0.31, 5.24; p = 0.03). Our research shows that elevated salinity in drinking water is associated with higher BP in young coastal populations. Blood pressure is an important risk factor of hypertension and cardiovascular diseases. Given the extent of salinization of freshwater in many low-lying countries including in Bangladesh, and the likely exacerbation related to climate change-induced sea level rise, implementation of preventative strategies through dietary interventions along with promotion of low saline drinking water must be a priority in these settings. - Highlights: • Freshwater salinization will affect more people and to a greater extent as climate projections are realised in low-lying regions of the world.

  10. The effect of mineral radon water applied in the form of full baths on blood pressure in patients with hypertension

    Directory of Open Access Journals (Sweden)

    Amila Kapetanović

    2013-04-01

    Full Text Available Introduction: Due to patients’ safety, increased blood pressure often restricts wider use of mineral water for therapeutic purposes in rehabilitation practice. The aim of this study was to examine the effect of radon mineral water applied in the form of full baths on blood pressure in people with hypertension.Methods: A total of 27 patients, average age 58.10 years with hypertension were included in the study. Balneotherapy was applied in the form of full baths with mineral radon water of neutral temperature. Values of systolic and diastolic blood pressure were measured before and after twenty minutes therapy on the first and fifth day of treatment.Results: On the first day of treatment there was no significant change in blood pressure after the application of full baths with mineral radon water of neutral temperature (systolic pressure t = 0.697, not significant; diastolic pressure t = 0.505, not significant. On the fi fth day of treatment there was no significant changes in blood pressure after the application of medical baths with mineral radon water of neutral temperature (systolic pressure t = 1.372, not significant; diastolic pressure t = 1.372, not significant.Conclusion: The significant increase of blood pressure in patients with mild and moderate hypertension is not expected when Fojnica water (radioactive mineral water is being used in the form of full baths of neutral temperature, which allows a broader application of this balneo procedure in rehabilitation practice.

  11. Water Flow Testing and Unsteady Pressure Analysis of a Two-Bladed Liquid Oxidizer Pump Inducer

    Science.gov (United States)

    Schwarz, Jordan B.; Mulder, Andrew; Zoladz, Thomas

    2011-01-01

    The unsteady fluid dynamic performance of a cavitating two-bladed oxidizer turbopump inducer was characterized through sub-scale water flow testing. While testing a novel inlet duct design that included a cavitation suppression groove, unusual high-frequency pressure oscillations were observed. With potential implications for inducer blade loads, these high-frequency components were analyzed extensively in order to understand their origins and impacts to blade loading. Water flow testing provides a technique to determine pump performance without the costs and hazards associated with handling cryogenic propellants. Water has a similar density and Reynolds number to liquid oxygen. In a 70%-scale water flow test, the inducer-only pump performance was evaluated. Over a range of flow rates, the pump inlet pressure was gradually reduced, causing the flow to cavitate near the pump inducer. A nominal, smooth inducer inlet was tested, followed by an inlet duct with a circumferential groove designed to suppress cavitation. A subsequent 52%-scale water flow test in another facility evaluated the combined inducer-impeller pump performance. With the nominal inlet design, the inducer showed traditional cavitation and surge characteristics. Significant bearing loads were created by large side loads on the inducer during synchronous cavitation. The grooved inlet successfully mitigated these loads by greatly reducing synchronous cavitation, however high-frequency pressure oscillations were observed over a range of frequencies. Analytical signal processing techniques showed these oscillations to be created by a rotating, multi-celled train of pressure pulses, and subsequent CFD analysis suggested that such pulses could be created by the interaction of rotating inducer blades with fluid trapped in a cavitation suppression groove. Despite their relatively low amplitude, these high-frequency pressure oscillations posed a design concern due to their sensitivity to flow conditions and

  12. Water vapor pressure over molten KH2PO4 and demonstration of water electrolysis at ∼300ºC

    DEFF Research Database (Denmark)

    Berg, Rolf W.; Nikiforov, Aleksey Valerievich; Petrushina, Irina

    2016-01-01

    A new potentially high-efficiency electrolyte for water electrolysis: molten monobasic potassium phosphate, KH2PO4 or KDP has been investigated at temperatures ∼275–325 °C. At these temperatures, KH2PO4 was found to dissociate into H2O gas in equilibrium with a melt mixture of KH2PO4—K2H2P2O7—KPO3...... of water and methane or hydrogen to be 0.40 ± 0.02 or 1.2 ± 0.03. At equilibrium the vapor pressure was much lower than the vapor pressure above liquid water at the same temperature. Electrolysis was realized by passing current through closed ampoules (vacuum sealed quartz glass electrolysis cells...... with platinum electrodes and the electrolyte melt). The formation of mixtures of hydrogen and oxygen gases as well as the water vapor was detected by Raman spectroscopy. In this way it was demonstrated that water is present in the new electrolyte: molten KH2PO4 can be split by electrolysis via the reaction 2H2O...

  13. Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

    Directory of Open Access Journals (Sweden)

    J. Mao

    2018-03-01

    Full Text Available The strategy denoted as in-vessel retention (IVR is widely used in severe accident (SA management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

  14. Robust aqua material. A pressure-resistant self-assembled membrane for water purification

    International Nuclear Information System (INIS)

    Cohen, Erez; Weissman, Haim; Rybtchinski, Boris; Shimoni, Eyal; Kaplan-Ashiri, Ifat; Werle, Kai; Wohlleben, Wendel

    2017-01-01

    ''Aqua materials'' that contain water as their major component and are as robust as conventional plastics are highly desirable. Yet, the ability of such systems to withstand harsh conditions, for example, high pressures typical of industrial applications has not been demonstrated. We show that a hydrogel-like membrane self-assembled from an aromatic amphiphile and colloidal Nafion is capable of purifying water from organic molecules, including pharmaceuticals, and heavy metals in a very wide range of concentrations. Remarkably, the membrane can sustain high pressures, retaining its function. The robustness and functionality of the water-based self-assembled array advances the idea that aqua materials can be very strong and suitable for demanding industrial applications. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  15. Robust aqua material. A pressure-resistant self-assembled membrane for water purification

    Energy Technology Data Exchange (ETDEWEB)

    Cohen, Erez; Weissman, Haim; Rybtchinski, Boris [Department of Organic Chemistry, Weizmann Institute of Science, 234 Herzl Street, Rehovot, 7610001 (Israel); Shimoni, Eyal; Kaplan-Ashiri, Ifat [Department of Chemical Research Support, Weizmann Institute of Science, 234 Herzl Street, Rehovot, 7610001 (Israel); Werle, Kai; Wohlleben, Wendel [Department of Material Physics, Materials and Systems Research, BASF SE, 67056, Ludwigshafen (Germany)

    2017-02-13

    ''Aqua materials'' that contain water as their major component and are as robust as conventional plastics are highly desirable. Yet, the ability of such systems to withstand harsh conditions, for example, high pressures typical of industrial applications has not been demonstrated. We show that a hydrogel-like membrane self-assembled from an aromatic amphiphile and colloidal Nafion is capable of purifying water from organic molecules, including pharmaceuticals, and heavy metals in a very wide range of concentrations. Remarkably, the membrane can sustain high pressures, retaining its function. The robustness and functionality of the water-based self-assembled array advances the idea that aqua materials can be very strong and suitable for demanding industrial applications. (copyright 2017 Wiley-VCH Verlag GmbH and Co. KGaA, Weinheim)

  16. Effect of supplementation of water-soluble vitamins on oxidative stress and blood pressure in prehypertensives.

    Science.gov (United States)

    Talikoti, Prashanth; Bobby, Zachariah; Hamide, Abdoul

    2015-01-01

    The objective of the study was to evaluate the effect of water-soluble vitamins on oxidative stress and blood pressure in prehypertensives. Sixty prehypertensives were recruited and randomized into 2 groups of 30 each. One group received water-soluble vitamins and the other placebo for 4 months. Further increase in blood pressure was not observed in the vitamin group which increased significantly in the placebo group at the end of 4 months. Malonedialdehyde and protein carbonylation were reduced during the course of treatment with vitamins whereas in the placebo group there was an increase in the level of malondialdehyde. In conclusion, supplementation of water-soluble vitamins in prehypertension reduces oxidative stress and its progression to hypertension.

  17. Thermodynamic data of water on smectite surface and those applications to swelling pressure of compacted bentonite

    International Nuclear Information System (INIS)

    Sato, H.

    2009-01-01

    Swelling pressure was discussed focusing on the thermodynamic properties of water on smectite (montmorillonite) which is the major clay mineral constituent of the bentonite buffer. The thermodynamic data of the water on the smectite surface were obtained as a function of water content and temperature in a range of dry density 0.6-0.9 Mg/m 3 . Purified Na-smectite of which all interlayer cations were exchanged with Na+ ions, was used. The activity (a H 2 O ) and the relative partial molar Gibbs free energy (ΔG H 2 O ) of the water were obtained at 25 C. Both a H 2 O and ΔG H 2 O decreased with a decrease of water content, and similar results were obtained to data reported for montmorillonite (Kunipia-F bentonite). Since the specific surface area of smectite is about 800 m 2 /g, water up to approximately 2 water layers from smectite surface is thermodynamically evaluated to be bound. Swelling pressure versus smectite partial density was calculated based on ΔG H 2 O and compared to data experimentally obtained for various kinds of bentonites. The calculated results were in good agreement with the measured data over the range of smectite partial density between 1.0 and 2.0 Mg/m 3 . (author)

  18. Pressure surge free fire water systems increase safety on offshore oil- and gas drilling platforms

    International Nuclear Information System (INIS)

    Carlsen, Randi

    2001-01-01

    The article describes a new fire water system for use on the oil- and gas drilling platforms that is characterized by improved start-up time and reduced energy consumption. Deluge valves are commonly used in fast large-flow fire water systems all over the world. During the test of a new fire water system on a platform a few years ago, a pipe near the living quarter suddenly ruptured due to an unexpected pressure surge thought to be impossible. It was caused by a weakness of the deluge valve. A better valve was needed and the 'UniqValve' was designed and manufactured. The UniqValve operates in cooperation with the fire pumps during start-up as it 'reads' the pressure variations of the water flow and corrects the water flow to the fire areas in less than a tenth of a second. The valve is now integrated in a modular system. The fire water unit is mounted in a container, which reduces cost and simplifies the placement of the fire water installation

  19. Environmental response nanosilica for reducing the pressure of water injection in ultra-low permeability reservoirs

    Science.gov (United States)

    Liu, Peisong; Niu, Liyong; Li, Xiaohong; Zhang, Zhijun

    2017-12-01

    The super-hydrophobic silica nanoparticles are applied to alter the wettability of rock surface from water-wet to oil-wet. The aim of this is to reduce injection pressure so as to enhance water injection efficiency in low permeability reservoirs. Therefore, a new type of environmentally responsive nanosilica (denote as ERS) is modified with organic compound containing hydrophobic groups and "pinning" groups by covalent bond and then covered with a layer of hydrophilic organic compound by chemical adsorption to achieve excellent water dispersibility. Resultant ERS is homogeneously dispersed in water with a size of about 4-8 nm like a micro-emulsion system and can be easily injected into the macro or nano channels of ultra-low permeability reservoirs. The hydrophobic nanosilica core can be released from the aqueous delivery system owing to its strong dependence on the environmental variation from normal condition to injection wells (such as pH and salinity). Then the exposed silica nanoparticles form a thin layer on the surface of narrow pore throat, leading to the wettability from water-wet to oil-wet. More importantly, the two rock cores with different permeability were surface treated with ERS dispersion with a concentration of 2 g/L, exhibit great reduce of water injection pressure by 57.4 and 39.6%, respectively, which shows great potential for exploitation of crude oil from ultra-low permeability reservoirs during water flooding. [Figure not available: see fulltext.

  20. Weak interactions between water and clathrate-forming gases at low pressures

    Energy Technology Data Exchange (ETDEWEB)

    Thurmer, Konrad; Yuan, Chunqing; Kimmel, Gregory A.; Kay, Bruce D.; Smith, R. Scott

    2015-11-01

    Using scanning probe microscopy and temperature programed desorption we examined the interaction between water and two common clathrate-forming gases, methane and isobutane, at low temperature and low pressure. Water co-deposited with up to 10-1 mbar methane or 10-5 mbar isobutane at 140 K onto a Pt(111) substrate yielded pure crystalline ice, i.e., the exposure to up to ~107 gas molecules for each deposited water molecule did not have any detectable effect on the growing films. Exposing metastable, less than 2 molecular layers thick, water films to 10-5 mbar methane does not alter their morphology, suggesting that the presence of the Pt(111) surface is not a strong driver for hydrate formation. This weak water-gas interaction at low pressures is supported by our thermal desorption measurements from amorphous solid water and crystalline ice where 1 ML of methane desorbs near ~43 K and isobutane desorbs near ~100 K. Similar desorption temperatures were observed for desorption from amorphous solid water.

  1. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  2. The self-similar turbulent flow of low-pressure water vapor

    Science.gov (United States)

    Konyukhov, V. K.; Stepanov, E. V.; Borisov, S. K.

    2018-05-01

    We studied turbulent flows of water vapor in a pipe connecting two closed vessels of equal volume. The vessel that served as a source of water vapor was filled with adsorbent in the form of corundum ceramic balls. These ceramic balls were used to obtain specific conditions to lower the vapor pressure in the source vessel that had been observed earlier. A second vessel, which served as a receiver, was empty of either air or vapor before each vapor sampling. The rate of the pressure increase in the receiver vessel was measured in a series of six samplings performed with high precision. The pressure reduction rate in the source vessel was found to be three times lower than the pressure growth rate in the receiver vessel. We found that the pressure growth rates in all of the adjacent pairs of samples could be arranged in a combination that appeared to be identical for all pairs, and this revealed the existence of a rather interesting and peculiar self-similarity law for the sampling processes under consideration.

  3. ANNEALING OF POLYCRYSTALLINE THIN FILM SILICON SOLAR CELLS IN WATER VAPOUR AT SUB-ATMOSPHERIC PRESSURES

    Directory of Open Access Journals (Sweden)

    Peter Pikna

    2014-10-01

    Full Text Available Thin film polycrystalline silicon (poly-Si solar cells were annealed in water vapour at pressures below atmospheric pressure. PN junction of the sample was contacted by measuring probes directly in the pressure chamber filled with steam during passivation. Suns-VOC method and a Lock-in detector were used to monitor an effect of water vapour to VOC of the solar cell during whole passivation process (in-situ. Tested temperature of the sample (55°C – 110°C was constant during the procedure. Open-circuit voltage of a solar cell at these temperatures is lower than at room temperature. Nevertheless, voltage response of the solar cell to the light flash used during Suns-VOC measurements was good observable. Temperature dependences for multicrystalline wafer-based and polycrystalline thin film solar cells were measured and compared. While no significant improvement of thin film poly-Si solar cell parameters by annealing in water vapour at under-atmospheric pressures was observed up to now, in-situ observation proved required sensitivity to changing VOC at elevated temperatures during the process.

  4. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  5. 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.

  6. Variation of Pressure with Depth of Water: Working with High-Tech and Low-Cost Materials

    Science.gov (United States)

    Ornek, Funda; Zziwa, Byansi Jude; Taganahan, Teresita D.

    2013-01-01

    When you dive underwater, you feel the pressure on your ears and, as you dive deeper, more pressure is felt. This article presents an activity that teachers might find useful for demonstrating the relationship between water depth and pressure. (Contains 5 figures and 1 table.)

  7. Real-time adjustment of pressure to demand in water distribution systems: Parameter-less P-controller algorithm

    CSIR Research Space (South Africa)

    Page, Philip R

    2016-08-01

    Full Text Available Remote real-time control is currently the most advanced form of pressure management. Here the parameters describing pressure control valves (or pumps) are changed in real-time in such a way to provide the most optimal pressure in the water...

  8. Measurement of pressure distributions and velocity fields of water jet intake flow

    International Nuclear Information System (INIS)

    Jeong, Eun Ho; Yoon, Sang Youl; Kwon, Seong Hoon; Chun, Ho Hwan; Kim, Mun Chan; Kim, Kyung Chun

    2002-01-01

    Waterjet propulsion system can avoid cavitation problem which is being arised conventional propeller propulsion system. The main issue of designing waterjet system is the boundary layer separation at ramp and lib of water inlet. The flow characteristics are highly depended on Jet to Velocity Ratio(JVR) as well as the intake geometry. The present study is conducted in a wind tunnel to provide accurate pressure destribution at the inlet wall and velocity field of the inlet and exit planes. Particle image velocimetry technique is used to obtain detail velocity fields. Pressure distributions and velocity field are discussed with accelerating and deaccelerating flow zones and the effect of JVR

  9. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    Science.gov (United States)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  10. In-reactor oxidation of zircaloy-4 under low water vapor pressures

    International Nuclear Information System (INIS)

    Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

    2015-01-01

    Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

  11. RELAP5 analyses of overcooling transients in a pressurized water reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.; Ogden, D.M.; Stitt, B.D.; Waterman, M.E.

    1983-01-01

    In support of the Pressurized Thermal Shock Integration Study sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.5 computer code. These analyses were performed for Oconee Plants 1 and 3, which are pressurized water reactors of Babcock and Wilcox lowered-loop design. Results of the RELAP5 analyses are presented, including a comparison with plant data. The capabilities and limitations of the RELAP5/MOD1.5 computer code in analyzing integral plant transients are examined. These analyses require detailed thermal-hydraulic and control system computer models

  12. RELAP5 thermal-hydraulic analyses of overcooling sequences in a pressurized water reactor

    International Nuclear Information System (INIS)

    Bolander, M.A.; Fletcher, C.D.; Davis, C.B.; Kullberg, C.M.; Stitt, B.D.; Waterman, M.E.; Burtt, J.D.

    1984-01-01

    In support of the Pressurized Thermal Shock Integration Study, sponsored by the United States Nuclear Regulatory Commission, the Idaho National Engineering Laboratory has performed analyses of overcooling transients using the RELAP5/MOD1.6 and MOD2.0 computer codes. These analyses were performed for the H.B. Robinson Unit 2 pressurized water reactor, which is a Westinghouse 3-loop design plant. Results of the RELAP5 analyses are presented. The capabilities of the RELAP5 computer code as a tool for analyzing integral plant transients requiring a detailed plant model, including complex trip logic and major control systems, are examined

  13. Sensor response monitoring in pressurized water reactors using time series modeling

    International Nuclear Information System (INIS)

    Upadhyaya, B.R.; Kerlin, T.W.

    1978-01-01

    Random data analysis in nuclear power reactors for purposes of process surveillance, pattern recognition and monitoring of temperature, pressure, flow and neutron sensors has gained increasing attention in view of their potential for helping to ensure safe plant operation. In this study, application of autoregressive moving-average (ARMA) time series modeling for monitoring temperature sensor response characteristrics is presented. The ARMA model is used to estimate the step and ramp response of the sensors and the related time constant and ramp delay time. The ARMA parameters are estimated by a two-stage algorithm in the spectral domain. Results of sensor testing for an operating pressurized water reactor are presented. 16 refs

  14. Instrumenting a pressure suppression experiment for a Mark I boiling water reactor: another measurements engineering challenge

    International Nuclear Information System (INIS)

    Shay, W.M.; Brough, W.G.; Miller, T.B.

    1978-01-01

    A 1 / 5 -scale test facility of a pressure-suppression system from a Mark I boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy, dynamic loading data during a hypothetical loss-of-coolant accident. A total of 27 air tests have been completed with an average of 175 transducers recorded for each test. An end-to-end calibration of the total measurement system was run to establish accuracy of the data. The instrumentation verified the analysis of the dynamic loading of the pressure-suppression system

  15. Modeling of hydrogen stratification in a pressurized water reactor containment with the contain computer code

    International Nuclear Information System (INIS)

    Kljenak, I.; Skerlavaj, A.; Parzer, I.

    1999-01-01

    Hydrogen distribution during a severe accident in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN computer code. The accidents is initiated by a large-break loss-of-coolant accident which is nit successfully mitigated by the action of the emergency core cooling system. Cases with and without successful actuation of spray systems and fan coolers were considered. The simulations predicted hydrogen stratification within the containment main compartment with intensive hydrogen mixing in the containment dome region. Pressure and temperature responses were analyzed as well.(author)

  16. Pressure impact of autoclave treatment on water sorption and pectin composition of flax cellulosic-fibres.

    Science.gov (United States)

    Alix, S; Colasse, L; Morvan, C; Lebrun, L; Marais, S

    2014-02-15

    The tensile properties of flax fibres might permit them to be used in composites as reinforcement in organic resin, as long as their mechanical properties are reproducible and their water sorption are reduced. In this study, to minimise the variability of mechanical properties, several samples of flax fibres were blended as a non-woven fabric. In order to reduce the water absorption of this non-woven technical fibres, an autoclave treatment was performed which was expected to remove the pectins and then to reduce the water sorption on their negative charges. The impact of autoclave pressure (0.5, 1 and 2 bars) on water sorption was investigated by using a gravimetric static equilibrium method. The Park model based on the three sorption modes: Langmuir, Henry's law and clustering, was successfully used to simulate the experimental sorption data. The lowest pressure treatments impacted only the Langmuir contribution while the 2 bar autoclave-treatment positively impacted the water resistance in the core of fibres by reducing Henry's absorption rate. This was shown to be related to the chemical modifications at the surface and in the core of fibres. A schematic model is presented relating the water sorption and the pectic composition of the fabric. Copyright © 2013 Elsevier Ltd. All rights reserved.

  17. Pressure management strategies for water loss reduction in large-scale water piping networks: a review

    CSIR Research Space (South Africa)

    Adedeji, K

    2017-06-01

    Full Text Available . Therefore, in a WDN with kth number of pipes, the probability of a pipe breakage in the network as a result of the system pressure variations is estimated as [17] 4.35 3.738 5 8 0.0021 21.4 Pr 1 10 k kD D k k k e D e D     (5) where Prk...

  18. Evaluation of water cooled supersonic temperature and pressure probes for application to 2000 F flows

    Science.gov (United States)

    Lagen, Nicholas T.; Seiner, John M.

    1990-01-01

    The development of water cooled supersonic probes used to study high temperature jet plumes is addressed. These probes are: total pressure, static pressure, and total temperature. The motivation for these experiments is the determination of high temperature supersonic jet mean flow properties. A 3.54 inch exit diameter water cooled nozzle was used in the tests. It is designed for exit Mach 2 at 2000 F exit total temperature. Tests were conducted using water cooled probes capable of operating in Mach 2 flow, up to 2000 F total temperature. Of the two designs tested, an annular cooling method was chosen as superior. Data at the jet exit planes, and along the jet centerline, were obtained for total temperatures of 900 F, 1500 F, and 2000 F, for each of the probes. The data obtained from the total and static pressure probes are consistent with prior low temperature results. However, the data obtained from the total temperature probe was affected by the water coolant. The total temperature probe was tested up to 2000 F with, and without, the cooling system turned on to better understand the heat transfer process at the thermocouple bead. The rate of heat transfer across the thermocouple bead was greater when the coolant was turned on than when the coolant was turned off. This accounted for the lower temperature measurement by the cooled probe. The velocity and Mach number at the exit plane and centerline locations were determined from the Rayleigh-Pitot tube formula.

  19. Improvement of corrosion resistance of vanadium alloys in high-temperature pressurized water

    International Nuclear Information System (INIS)

    Fujiwara, Mitsuhiro; Sakamoto, Toshiya; Satou, Manabu; Hasegawa, Akira; Abe, Katsunori; Kaiuchi, Kazuo; Furuya, Takemi

    2005-01-01

    Corrosion tests in pressurized and vaporized water were conducted for V-based high Cr and Ti alloys and V-4Cr-4Ti type alloys containing minor elements such as Si, Al and Y. Weight losses were observed for every alloy after corrosion tests in pressurized water. It was apparent that addition of Cr effectively reduced the weight change in pressurized water. The weight loss of V-4Cr-4Ti type alloys in corrosion tests in vaporized water was also reduced as Cr content increased. The V-20Cr-4Ti alloy had a slight weight gain, almost same as that of SUS316, which had the best corrosion properties in the tested alloys. The elongation of alloys with in excess of 10% Cr was reduced as Cr content increased. The elongations of the V-12Cr-4Ti and the V-15Cr-4Ti alloys were significantly reduced by corrosion and cleavage fracture was observed reflecting hydrogen embrittlement. The reduced elongations of the alloys of the alloys were recovered to the same level of as annealed conditions after hydrogen degassing. After corrosion, the V-15Cr-4Ti-0.5Y alloy still kept enough elongation, suggesting that the addition of Y is effective to reduce the hydrogen embrittlement. (author)

  20. The Effect of Pressure and Temperature on Mid-Infrared Sensing of Dissolved Hydrocarbons in Water.

    Science.gov (United States)

    Heath, Charles; Myers, Matthew; Pejcic, Bobby

    2017-12-19

    Attenuated total reflectance Fourier transform infrared (ATR-FTIR) spectroscopy using a polymer coated internal reflection element/waveguide is an established sensor platform for the detection of a range of organic and hydrocarbon molecules dissolved in water. The polymer coating serves two purposes: to concentrate hydrocarbons from the aqueous phase and to exclude water along with other interfering molecules from the surface of the internal reflection element. Crucial to reliable quantification and analytical performance is the calibration of the ATR-FTIR sensor which is commonly performed in water under mild ambient conditions (i.e., 25 °C and 1 atm). However, there is a pressing need to monitor environmental and industrial processes/events that may occur at high pressures and temperatures where this calibration approach is unsuitable. Using a ruggedized optical fiber probe with a diamond-based ATR, we have conducted mid-infrared sensor experiments to understand the influence of high pressure (up to 207 bar) and temperature (up to 80 °C) on the detection of toluene and naphthalene dissolved in water. Using a poly(isobutylene) film, we have shown that the IR spectroscopic response is relatively unaffected by changes in pressure; however, a diminished response was observed with increasing temperature. We reveal that changes in the refractive index of the polymer film with temperature have only a minor effect on sensitivity. A more plausible explanation for the observed significant change in sensor response with temperature is that the partitioning process is exothermic and becomes less favorable with increasing temperature. This Article shows that the sensitivity is relatively invariant to pressure; however, the thermal variations are significant and need to be considered when quantifying the concentration of hydrocarbons in water.

  1. Characterization of the TIP4P-Ew water model: vapor pressure and boiling point.

    Science.gov (United States)

    Horn, Hans W; Swope, William C; Pitera, Jed W

    2005-11-15

    The liquid-vapor-phase equilibrium properties of the previously developed TIP4P-Ew water model have been studied using thermodynamic integration free-energy simulation techniques in the temperature range of 274-400 K. We stress that free-energy results from simulations need to be corrected in order to be compared to the experiment. This is due to the fact that the thermodynamic end states accessible through simulations correspond to fictitious substances (classical rigid liquids and classical rigid ideal gases) while experiments operate on real substances (liquids and real gases, with quantum effects). After applying analytical corrections the vapor pressure curve obtained from simulated free-energy changes is in excellent agreement with the experimental vapor pressure curve. The boiling point of TIP4P-Ew water under ambient pressure is found to be at 370.3+/-1.9 K, about 7 K higher than the boiling point of TIP4P water (363.7+/-5.1 K; from simulations that employ finite range treatment of electrostatic and Lennard-Jones interactions). This is in contrast to the approximately +15 K by which the temperature of the density maximum and the melting temperature of TIP4P-Ew are shifted relative to TIP4P, indicating that the temperature range over which the liquid phase of TIP4P-Ew is stable is narrower than that of TIP4P and resembles more that of real water. The quality of the vapor pressure results highlights the success of TIP4P-Ew in describing the energetic and entropic aspects of intermolecular interactions in liquid water.

  2. Molecular Dynamics of Equilibrium and Pressure-Driven Transport Properties of Water through LTA-Type Zeolites

    KAUST Repository

    Turgman-Cohen, Salomon; Araque, Juan C.; Hoek, Eric M. V.; Escobedo, Fernando A.

    2013-01-01

    We consider an atomistic model to investigate the flux of water through thin Linde type A (LTA) zeolite membranes with differing surface chemistries. Using molecular dynamics, we have studied the flow of water under hydrostatic pressure through a fully hydrated LTA zeolite film (∼2.5 nm thick) capped with hydrophilic and hydrophobic moieties. Pressure drops in the 50-400 MPa range were applied across the membrane, and the flux of water was monitored for at least 15 ns of simulation time. For hydrophilic membranes, water molecules adsorb at the zeolite surface, creating a highly structured fluid layer. For hydrophobic membranes, a depletion of water molecules occurs near the water/zeolite interface. For both types of membranes, the water structure is independent of the pressure drop established in the system and the flux through the membranes is lower than that observed for the bulk zeolitic material; the latter allows an estimation of surface barrier effects to pressure-driven water transport. Mechanistically, it is observed that (i) bottlenecks form at the windows of the zeolite structure, preventing the free flow of water through the porous membrane, (ii) water molecules do not move through a cage in a single-file fashion but rather exhibit a broad range of residence times and pronounced mixing, and (iii) a periodic buildup of a pressure difference between inlet and outlet cages takes place which leads to the preferential flow of water molecules toward the low-pressure cages. © 2013 American Chemical Society.

  3. Molecular Dynamics of Equilibrium and Pressure-Driven Transport Properties of Water through LTA-Type Zeolites

    KAUST Repository

    Turgman-Cohen, Salomon

    2013-10-08

    We consider an atomistic model to investigate the flux of water through thin Linde type A (LTA) zeolite membranes with differing surface chemistries. Using molecular dynamics, we have studied the flow of water under hydrostatic pressure through a fully hydrated LTA zeolite film (∼2.5 nm thick) capped with hydrophilic and hydrophobic moieties. Pressure drops in the 50-400 MPa range were applied across the membrane, and the flux of water was monitored for at least 15 ns of simulation time. For hydrophilic membranes, water molecules adsorb at the zeolite surface, creating a highly structured fluid layer. For hydrophobic membranes, a depletion of water molecules occurs near the water/zeolite interface. For both types of membranes, the water structure is independent of the pressure drop established in the system and the flux through the membranes is lower than that observed for the bulk zeolitic material; the latter allows an estimation of surface barrier effects to pressure-driven water transport. Mechanistically, it is observed that (i) bottlenecks form at the windows of the zeolite structure, preventing the free flow of water through the porous membrane, (ii) water molecules do not move through a cage in a single-file fashion but rather exhibit a broad range of residence times and pronounced mixing, and (iii) a periodic buildup of a pressure difference between inlet and outlet cages takes place which leads to the preferential flow of water molecules toward the low-pressure cages. © 2013 American Chemical Society.

  4. Development of plant dynamic analysis code for integrated self-pressurized water reactor (ISPDYN), and comparative study of pressure control methods

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Yokomura, Takeyoshi; Nabeshima, Kunihiko; Shimazaki, Junya; Shinohara, Yoshikuni.

    1988-01-01

    This report describes the development of plant dynamic analysis code (ISPDYN) for integrated self-pressurized water reactor, and comparative study of pressure control methods with this code. ISPDYN is developed for integrated self-pressurized water reactor, one of the trial design by JAERI. In the transient responses, the calculated results by ISPDYN are in good agreement with the DRUCK calculations. In addition, this report presents some sensitivity studies for selected cases. Computing time of this code is very short so as about one fifth of real time. The comparative study of self-pressurized system with forced-pressurized system by this code, for rapid load decrease and increase cases, has provided useful informations. (author)

  5. Heat transfer in vertical pipe flow at supercritical pressures of water

    International Nuclear Information System (INIS)

    Loewenberg, M.F.

    2007-05-01

    A new reactor concept with light water at supercritical conditions is investigated in the framework of the European project ''High Performance Light Water Reactor'' (HPLWR). Characteristics of this reactor are the system pressure and the coolant outlet temperature above the critical point of water. Water is regarded as a single phase fluid under these conditions with a high energy density. This high energy density should be utilized in a technical application. Therefore in comparison with up to date nuclear power plants some constructive savings are possible. For instance, steam dryers or steam separators can be avoided in contrast to boiling water reactors. A thermal efficiency of about 44% can be accomplished at a system pressure of 25MPa through a water heat-up from 280 C to 510 C. To ensure this heat-up within the core reliable predictions of the heat transfer are necessary. Water as the working fluid changes its fluid properties dramatically during the heat up in the core. As such; the density in the core varies by the factor of seven. The motivation to develop a look-up table for heat transfer predications in supercritical water is due to the significant temperature dependence of the fluid properties of water. A systematic consolidation of experimental data was performed. Together with further developments of the methods to derive a look-up table made it possible to develop a look-up table for heat transfer in supercritical water in vertical flows. A look-up table predicts the heat transfer for different boundary conditions (e.g. pressure or heat flux) with tabulated data. The tabulated wall temperatures for fully developed turbulent flows can be utilized for different geometries by applying hydraulic diameters. With the developed look-up table the difficulty of choosing one of the many published correlations can be avoided. In general, the correlations have problems with strong fluid property variations. Strong property variations combined with high heat

  6. Material and water chemistry for a ferritic reactor coolant system in pressure water reactors

    International Nuclear Information System (INIS)

    Stieding, L.

    1979-04-01

    The use of unplated, low-alloy steels in a boric acid controlled PWR is not considered possible without changing the water conditions during the start-up and shut-down periods of the reactor. The significant pH reduction of the water due to boric acid during these periods most probably leads to damage of the magnetite protective layers followed by selective corrosion. As this highly important process has not been sufficiently evaluated with respect to our specific application problem, more detailed information will be necessary. KWU test facilities provide a means of performing such tests. In order to avoid corrosion attack during the above operating conditions, an inhibition of the water with 7 Li-borate is recommended which, however, will amount to approx. DM 60.000,-- per period of use. (orig.) [de

  7. Vapour pressures, densities, and viscosities of the (water + lithium bromide + potassium acetate) system and (water + lithium bromide + sodium lactate) system

    International Nuclear Information System (INIS)

    Lucas, Antonio de; Donate, Marina; Rodriguez, Juan F.

    2006-01-01

    Measurements of thermophysical properties (vapour pressure, density, and viscosity) of the (water + lithium bromide + potassium acetate) system LiBr:CH 3 COOK = 2:1 by mass ratio and the (water + lithium bromide + sodium lactate) system LiBr:CH 3 CH(OH)COONa = 2:1 by mass ratio were measured. The system, a possible new working fluid for absorption heat pump, consists of absorbent (LiBr + CH 3 COOK) or (LiBr + CH 3 CH(OH)COONa) and refrigerant H 2 O. The vapour pressures were measured in the ranges of temperature and absorbent concentration from T = (293.15 to 333.15) K and from mass fraction 0.20 to 0.50, densities and viscosities were measured from T = (293.15 to 323.15) K and from mass fraction 0.20 to 0.40. The experimental data were correlated with an Antoine-type equation. Densities and viscosities were measured in the same range of temperature and absorbent concentration as that of the vapour pressure. Regression equations for densities and viscosities were obtained with a minimum mean square error criterion

  8. Intelligent Pressure Management to Reduce Leakage in Urban Water Supply Networks, A Case Study of Sarafrazan District, Mashhad

    Directory of Open Access Journals (Sweden)

    Mohammad Soltani Asl

    2009-09-01

    Full Text Available Water losses are inevitable in urban water distribution systems. The two approaches adopted nowadays to combat this problem include management of hydraulic parameters such as pressure and leakage detection in the network. Intellitgent pressure management is a suitable technique for controlling leakage and reducing damages due to high operating pressures in a network. This paper aims to investigate the effects of pressure reduction on leakage. The EPANET 2.10 software is used to simulate the water distribution network in the Sarafrazan District,Mashhad, assuming leakage from network nodes. The results are then used to develop a pressure variation program based on the patterns obtained from the simulation, which is applied to the pressure reducing valve. The results show that pressure management can reduce nightly leakage by up to 35% while maintaining a more uniform pressure distribution. Implementation of the time-dependent pressure pattern by applying programmable pressure reducing valves in a real urban water distribution network is feasible and plays a key role in reducing water losses to leakage.

  9. Orthogonal optimization of a water hydraulic pilot-operated pressure-reducing valve

    Science.gov (United States)

    Mao, Xuyao; Wu, Chao; Li, Bin; Wu, Di

    2017-12-01

    In order to optimize the comprehensive characteristics of a water hydraulic pilot-operated pressure-reducing valve, numerical orthogonal experimental design was adopted. Six parameters of the valve, containing diameters of damping plugs, volume of spring chamber, half cone angle of main spool, half cone angle of pilot spool, mass of main spool and diameter of main spool, were selected as the orthogonal factors, and each factor has five different levels. An index of flowrate stability, pressure stability and pressure overstrike stability (iFPOS) was used to judge the merit of each orthogonal attempt. Embedded orthogonal process turned up and a final optimal combination of these parameters was obtained after totally 50 numerical orthogonal experiments. iFPOS could be low to a fairly low value which meant that the valve could have much better stabilities. During the optimization, it was also found the diameters of damping plugs and main spool played important roles in stability characteristics of the valve.

  10. Study of radiation damage of steels for light water pressure vessels at UJV

    International Nuclear Information System (INIS)

    Vacek, N.; Stoces, B.

    1980-01-01

    Preoperational determination of radiation resistance of pressure vessel steels is performed at accelerated neutron exposure in a test or materials research reactor. The results obtained at accelerated and operating exposure are not fully identical and surveillance bodies are therefore used manufactured from the pressure vessel material. Currently, the following steels are used for the manufacture of light water reactor pressure vessels: Mn-Mo-Ni (ASTM-A533-B, ASTM-A508), Cr-Mo-V (15Kh2M1FA). At UJV Rez, for irradiation Chanca-M probes imported from France are used featuring electric temperature control. Almost identical radiation embrittlement was measured for all three steels after irradiation with a neutron fluence of 3x10 23 n.m -2 at a temperature of 290 degC. (H.S.)

  11. Impact of climate change and anthropogenic pressure on the water resources of India: challenges in management

    Directory of Open Access Journals (Sweden)

    K. Shadananan Nair

    2016-10-01

    Full Text Available Freshwater resources of India are getting fast degraded and depleted from the changing climate and pressure of fast rising population. Changing intensity and seasonality of rainfall affect quantity and quality of water. Most of the rivers are polluted far above safety limits from the untreated domestic, industrial and agricultural effluents. Changes in the intensity, frequency and tracks of storms salinate coastal aquifers. Aquifers are also under the threat from rising sea level. Groundwater in urban limits and industrial zones are far beyond safety limits. Large-scale destruction of wetlands for industries and residential complexes has affected the quality of surface and groundwater resources in most parts of India. Measures to maintain food security and the new developments schemes such as river linking will further deteriorate the water resources. Falling water availability leads to serious health issues and various socio-economic issues. India needs urgent and appropriate adaptation strategies in the water sector.

  12. Impact of climate change and anthropogenic pressure on the water resources of India: challenges in management

    Science.gov (United States)

    Shadananan Nair, K.

    2016-10-01

    Freshwater resources of India are getting fast degraded and depleted from the changing climate and pressure of fast rising population. Changing intensity and seasonality of rainfall affect quantity and quality of water. Most of the rivers are polluted far above safety limits from the untreated domestic, industrial and agricultural effluents. Changes in the intensity, frequency and tracks of storms salinate coastal aquifers. Aquifers are also under the threat from rising sea level. Groundwater in urban limits and industrial zones are far beyond safety limits. Large-scale destruction of wetlands for industries and residential complexes has affected the quality of surface and groundwater resources in most parts of India. Measures to maintain food security and the new developments schemes such as river linking will further deteriorate the water resources. Falling water availability leads to serious health issues and various socio-economic issues. India needs urgent and appropriate adaptation strategies in the water sector.

  13. Experimental and numerical investigation of heat transfer from a narrow annulus to supercritical pressure water

    International Nuclear Information System (INIS)

    Wang, Han; Bi, Qincheng; Yang, Zhendong; Wang, Linchuan

    2015-01-01

    Highlights: • Heat transfer of supercritical water in a narrow annulus is investigated. • Effects of system parameters and flow direction on heat transfer are studied. • Deteriorated heat transfer is analyzed both experimentally and numerically. - Abstract: Heat transfer characteristics of supercritical pressure water in a narrow annulus with vertically upward and downward flows were investigated experimentally and numerically. The outer diameter of the inner heated rod is 8 mm with an effective heated length of 620 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 400–1000 kg/m 2 s and heat flux on the outer surface of the heated rod from 200 to 1000 kW/m 2 . The general heat transfer behaviors were discussed with respect to various mass fluxes and pressures. According to the experimental data, it was found that the effect of flow direction on heat transfer depends on the heat-flux to mass-flux ratio (q/G). Heat transfer is much improved in the downward flow compared to that of upward flow at high q/G ratios. At the pressure of 25 MPa, low-mass-flux deteriorated heat transfer occurred in the upward flow but not in the downward flow. At the same test parameters, however, heat transfer deterioration was observed at both of the two flow directions when the pressure was lowered to 23 MPa. The experimental results indicate that buoyancy plays an important role for this type of deterioration, but is not the only mechanism that leads to the heat transfer deterioration. Three turbulence models were assessed against the annulus test data, it was found that the SST k-ω model gives a satisfying prediction of heat transfer deterioration especially for the case of downward flow. The mechanisms for the low-mass-flow heat transfer deterioration were investigated from the viewpoints of buoyancy and property variations of the supercritical water

  14. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

  15. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  16. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  17. Common safety approach for future pressurized water reactors in France and Germany

    International Nuclear Information System (INIS)

    Queniart, D.; Gros, G.

    1994-01-01

    In France and Germany all major activities related to future pressurized water reactors are now proceeding in a coordinated way among the two countries. This holds for utilities and industry in the development of a joint PWR project, the ''European Pressurized Water Reactor (EPR)'' by Electricite de France (EDF), German utilities, Nuclear Power International (NPI), Framatome and Siemens as well as for the technical safety objectives for future evolutionary reactors on the basis of a common safety approach adopted by the safety authorities of both countries for plants to operate form the beginning of the next century. The proposed paper covers this common development of a safety approach and particular technical safety objectives. (authors). 5 refs. 1 fig

  18. Effect of pressurized hot water extraction on antioxidants from grape pomace before and after enological fermentation.

    Science.gov (United States)

    Vergara-Salinas, José R; Bulnes, Pedro; Zúñiga, María Carolina; Pérez-Jiménez, Jara; Torres, Josep Lluís; Mateos-Martín, María Luisa; Agosin, Eduardo; Pérez-Correa, José R

    2013-07-17

    Grape pomace was extracted with pressurized hot water at laboratory scale before and after fermentation to explore the effects of fermentation and extraction temperature (50-200 °C) and time (5 and 30 min) on total extracted antioxidant levels and activity and to determine the content and recovery efficiency of main grape polyphenols, anthocyanins, and tannins. Fermented pomace yielded more total antioxidants (TAs), antioxidant activity, and tannins, than unfermented pomace but fewer anthocyanins. Elevating the extraction temperature increased TA extraction and antioxidant activity. Maximum anthocyanin extraction yields were achieved at 100 °C and at 150 °C for tannins and tannin-anthocyanin adducts. Using higher temperatures and longer extraction times resulted in a sharp decrease of polyphenol extraction yield. Relevant proanthocyanidin amounts were extracted only at 50 and 100 °C. Finally, TA recovery and activity were not directly related to the main polyphenol content when performing pressurized hot water grape pomace extraction.

  19. 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

  20. 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