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Sample records for containment pressure boundary

  1. Methods for assessing NPP containment pressure boundary integrity

    International Nuclear Information System (INIS)

    Naus, D.J.; Ellingwood, B.R.; Graves, H.L.

    2004-01-01

    Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion

  2. Fragility Modeling of Aging Containment Metallic Pressure Boundaries

    International Nuclear Information System (INIS)

    Cherry, J.L.; Ellingwood, B.R.

    1999-01-01

    The containment in a nuclear power plant (NPP) provides a barrier against the release of radioactivity in the event of an accident. Corrosion that has been observed in some steel containments and liners of reinforced concrete containments has raised questions about their ability to perform this function. The performance of corroded containments during events at or beyond the design basis is impacted by numerous sources of uncertainty. A fragility model of the containment provides a relatively simple depiction of the impact of uncertainties on structural performance and a basis for decision-making in the presence of uncertainty. Moreover, it is a necessary ingredient of any time-dependent structural reliability analysis. A nonlinear finite element analysis of containment response furnishes the necessary platform to perform numerical experiments to determine containment fragility. A statistically-based sampling plan minimizes the finite element computations required to develop the fragility curve. The -percentile (or other fractile) then gives a statistically based indication of the lower bound on containment capacity, and can be used as a screening tool to determine whether more refined further analysis or tests to support service life evaluations are warranted

  3. A survey of repair practices for nuclear power plant containment metallic pressure boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Oland, C.B.; Naus, D.J. [Oak Ridge National Lab., TN (United States)

    1998-05-01

    The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition.

  4. A survey of repair practices for nuclear power plant containment metallic pressure boundaries

    International Nuclear Information System (INIS)

    Oland, C.B.; Naus, D.J.

    1998-05-01

    The Nuclear Regulatory Commission has initiated a program at the Oak Ridge National Laboratory to provide assistance in their assessment of the effects of potential degradation on the structural integrity and leaktightness of metal containment vessels and steel liners of concrete containments in nuclear power plants. One of the program objectives is to identify repair practices for restoring metallic containment pressure boundary components that have been damaged or degraded in service. This report presents issues associated with inservice condition assessments and continued service evaluations and identifies the rules and requirements for the repair and replacement of nonconforming containment pressure boundary components by welding or metal removal. Discussion topics include base and welding materials, welding procedure and performance qualifications, inspection techniques, testing methods, acceptance criteria, and documentation requirements necessary for making acceptable repairs and replacements so that the plant can be returned to a safe operating condition

  5. Aging of the containment pressure boundary in light-water reactor plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.R.

    1997-01-01

    Research is being conducted by the Oak Ridge National Laboratory to address aging of the containment pressure boundary in light-water reactor plants. The objectives of this work are to (1) identify the significant factors related to occurrence of corrosion, efficacy of inspection, and structural capacity reduction of steel containments and liners of concrete containments, and to make recommendations on use of risk models in regulatory decisions; (2) provide NRC reviewers a means of establishing current structural capacity margins for steel containments, and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by NRC reviewers in assessing the seriousness of reported incidences of containment degradation. In meeting these objectives research is being conducted in two primary task areas - pressure boundary condition assessment and root-cause resolution practices, and reliability-based condition assessments. Under the first task area a degradation assessment methodology was developed for use in characterizing the in-service condition of metal and concrete containment pressure boundary components and quantifying the amount of damage that is present. An assessment of available destructive and nondestructive techniques for examining steel containments and liners is ongoing. Under the second task area quantitative structural reliability analysis methods are being developed for application to degraded metallic pressure boundaries to provide assurances that they will be able to withstand future extreme loads during the desired service period with a level of reliability that is sufficient for public safety. To date, mathematical models that describe time-dependent changes in steel due to aggressive environmental factors have been identified, and statistical data supporting their use in time-dependent reliability analysis have been summarized

  6. Nuclear power plant containment metallic pressure boundary materials and plans for collecting and presenting their properties

    International Nuclear Information System (INIS)

    Oland, C.B.

    1995-04-01

    A program is being conducted at the Oak Ridge National Laboratory (ORNL to assist the Nuclear Regulatory Commission (NRC)) in their assessment of the effects of degradation (primarily corrosion) on the structural capacity and leaktight integrity of metal containments and steel liners of reinforced concrete structures in nuclear power plants. One of the program objectives is to characterize and quantify manifestations of corrosion on the properties of steels used to construct containment pressure boundary components. This report describes a plan for use in collecting and presenting data and information on ferrous alloys permitted for use in construction of pressure retaining components in concrete and metal containments. Discussions about various degradation mechanisms that could potentially affect the mechanical properties of these materials are also included. Conclusions and recommendations presented in this report will be used to guide the collection of data and information that will be used to prepare a material properties data base for containment steels

  7. Reactor pressure boundary materials

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Chi, S. H.; Lee, B. S.

    2002-04-01

    With a long-term operation of nuclear power plants, the component materials are degraded under severe reactor conditions such as neutron irradiation, high temperature, high pressure and corrosive environment. It is necessary to establish the reliable and practical technologies for improving and developing the component materials and for evaluating the mechanical properties. Especially, it is very important to investigate the technologies for reactor pressure boundary materials such as reactor vessel and pipings in accordance with their critical roles. Therefore, this study was focused on developing and advancing the microstructural/micro-mechanical evaluation technologies, and on evaluating the neutron irradiation characteristics and radiation effects analysis technology of the reactor pressure boundary materials, and also on establishing a basis of nuclear material property database

  8. Advanced Pressure Boundary Materials

    Energy Technology Data Exchange (ETDEWEB)

    Santella, Michael L [ORNL; Shingledecker, John P [ORNL

    2007-01-01

    Increasing the operating temperatures of fossil power plants is fundamental to improving thermal efficiencies and reducing undesirable emissions such as CO{sub 2}. One group of alloys with the potential to satisfy the conditions required of higher operating temperatures is the advanced ferritic steels such as ASTM Grade 91, 9Cr-2W, and 12Cr-2W. These are Cr-Mo steels containing 9-12 wt% Cr that have martensitic microstructures. Research aimed at increasing the operating temperature limits of the 9-12 wt% Cr steels and optimizing them for specific power plant applications has been actively pursued since the 1970's. As with all of the high strength martensitic steels, specifying upper temperature limits for tempering the alloys and heat treating weldments is a critical issue. To support this aspect of development, thermodynamic analysis was used to estimate how this critical temperature, the A{sub 1} in steel terminology, varies with alloy composition. The results from the thermodynamic analysis were presented to the Strength of Weldments subgroup of the ASME Boiler & Pressure Vessel Code and are being considered in establishing maximum postweld heat treatment temperatures. Experiments are also being planned to verify predictions. This is part of a CRADA project being done with Alstom Power, Inc.

  9. Pressure effect on grain boundary diffusion

    International Nuclear Information System (INIS)

    Smirnova, E.S.; Chuvil'deev, V.N.

    1997-01-01

    The influence of hydrostatic pressure on grain boundary diffusion and grain boundary migration in metallic materials is theoretically investigated. The model is suggested that permits describing changes in activation energy of grain boundary self-diffusion and diffusion permeability of grain boundaries under hydrostatic pressure. The model is based on the ideas about island-type structure of grain boundaries as well as linear relationship of variations in grain boundary free volume to hydrostatic pressure value. Comparison of theoretical data with experimental ones for a number of metals and alloys (α-Zr, Sn-Ge, Cu-In with Co, In, Al as diffusing elements) shows a qualitative agreement

  10. Continuous containment monitoring with containment pressure fluctuation

    International Nuclear Information System (INIS)

    Dick, J.E.

    1996-01-01

    The monitoring of the integrity of containments particularly but not exclusively for nuclear plants is dealt with in this invention. While this application is primarily concerned with containment monitoring in the context of the single unit design, it is expected that the concepts presented will be universally applicable to any containment design, including containments for non-nuclear applications such as biological laboratories. The nuclear industry has long been interested in a means of monitoring containment integrity on a continuous basis, that is, while the reactor is operating normally. 12 refs., 2 figs

  11. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  12. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  13. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  14. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  15. Pressure suppression facility for reactor container

    International Nuclear Information System (INIS)

    Fujii, Tadashi; Fukui, Toru; Kataoka, Yoshiyuki; Tominaga, Kenji.

    1993-01-01

    In a nuclear reactor comprising heat transfer surfaces from a pressure suppression pool at the inside to the outer circumferential pool at the outside, a means for supplying water from a water supply source at the outside of the container to the pools is disposed. Then, a heat transfer means is disposed between the pressure suppression chamber and the water cooling pool. The water supply means comprises a pressurization means for applying pressure to water of the water supply source and a water supply channel. Water is supplied into the pressure suppression pool and the outer circumferential pool to elevate the water level and extend the region of heat contact with the water cooling heat transfer means. In addition, since dynamic pressure is applied to the feedwater, for example, by pressurizing the water surface of the water supply source, water can be supplied without using dynamic equipments such as pumps. Then, since water-cooling heat transfer surface can be extended after occurrence of accident, enlargement of a reactor container and worsening of earthquake proofness can be avoided as much as possible, to improve function for suppressing the pressure in the container. Further, since water-cooling heat transfer region can be extended, the arrangement of the water source and the place to which water is supplied is made optional without considering the relative height therebetween, to improve earthquake proofness. (N.H.)

  16. Pressure releasing device for reactor container

    International Nuclear Information System (INIS)

    Takeda, Mika.

    1994-01-01

    In the present invention, dose rate to public caused by radioactive rare gases can be decreased. That is, a reactor container contains a reactor pressure vessel incorporating a reactor core. There are disposed a pressure releasing system for releasing the pressure in the reactor pressure vessel to the outside, and a burning device for burning gases released from the pressure releasing system. An exhaustion pipe is disposed to the pressure releasing system. A burning device is disposed to the exhaustion pipe. It is effective to dispose a ventilation port at a portion of the exhaustion pipe upstream of the burning device. In addition, the burning device may preferably be disposed in a multi-stage in the axial direction of the exhaustion pipe. With such procedures, hydrogen in gases discharged along with the release of the pressure in the container is burned. Buoyancy is caused to the exhaustion gases by heat energy upon burning. Since the exhaustion gases can reach a higher level by the buoyancy, the dose rate due to the rare gases can be reduced. (I.S.)

  17. Analysis of specific factors causing RCS pressure boundary cracking

    International Nuclear Information System (INIS)

    Song, Taek-Ho; Jeong, Il-Seok

    2007-01-01

    As nuclear power plants become aged, pressure boundary integrity has become so important issue in domestic and foreign nuclear industry that many related research projects are on-going. KEPRI is going to embark a new research project for managing and preventing these kinds of cracks in nuclear power plants (NPPs). Many nuclear power plants experienced pressure boundary stress corrosion cracking (SCC) and shut downed because of it. In USA, V.C. Summer plant experienced reactor coolant pipe SCC near reactor outlet nozzle and Davis Vesse plant experienced reactor head crack around penetration pipe which is used to control rod drive mechanism. In this paper, RCS pressure boundary cracking cases and corrosion potential have been studied to find out what are the specific factors that have affected crack initiations in the reactor coolant pressure boundaries

  18. Role of the vertical pressure gradient in wave boundary layers

    DEFF Research Database (Denmark)

    Jensen, Karsten Lindegård; Sumer, B. Mutlu; Vittori, Giovanna

    2014-01-01

    By direct numerical simulation (DNS) of the flow in an oscillatory boundary layer, it is possible to obtain the pressure field. From the latter, the vertical pressure gradient is determined. Turbulent spots are detected by a criterion involving the vertical pressure gradient. The vertical pressure...... gradient is also treated as any other turbulence quantity like velocity fluctuations and statistical properties of the vertical pressure gradient are calculated from the DNS data. The presence of a vertical pressure gradient in the near bed region has significant implications for sediment transport....

  19. Vortex statistics for turbulence in a container with rigid boundaries

    DEFF Research Database (Denmark)

    Clercx, H.J.H.; Nielsen, A.H.

    2000-01-01

    The evolution of vortex statistics for decaying two-dimensional turbulence in a square container with rigid no-slip walls is compared with a few available experimental results and with the scaling theory of two-dimensional turbulent decay as proposed by Carnevale et al. Power-law exponents......, computed from an ensemble average of several numerical runs, coincide with some experimentally obtained values, but not with data obtained from numerical simulations of decaying two-dimensional turbulence with periodic boundary conditions....

  20. Aging of elastomers in CANDU pressure boundary service

    International Nuclear Information System (INIS)

    VanBerlo, C.; Leidner, J.

    1987-09-01

    This report describes the properties and aging of elastomers, and examines the performance of major elastomeric components in CANDU pressure boundary service. The components examined are vacuum building roof seals, pressure relief duct seals, airlock door seals, fuelling machine hoses, and cable penetrations. For each of these components, the design requirements, technical specifications and component testing procedures are compared with applicable standards. Information on actual and recommended monitoring and maintenance methods is presented. Operational and environmental stressors are identified. Component failure modes, causes and frequencies are described, as well as the remedial action taken. Many different elastomers are used in CANDU plants, for many different applications. Standards and manufacturers' recommendations are not consistent and may vary from one component to another. Accordingly, the monitoring, maintenance and replacement practices tend to vary from one application to another, and may also be different at different stations. Recommendations are given in this report for improved monitoring and maintenance, in an attempt to provide more consistency in approach. A summary of some experiences with elastomers from non-Canadian sources is contained in the last section. 125 refs

  1. A parametric study of adverse pressure gradient turbulent boundary layers

    International Nuclear Information System (INIS)

    Monty, J.P.; Harun, Z.; Marusic, I.

    2011-01-01

    There are many open questions regarding the behaviour of turbulent boundary layers subjected to pressure gradients and this is confounded by the large parameter space that may affect these flows. While there have been many valuable investigations conducted within this parameter space, there are still insufficient data to attempt to reduce this parameter space. Here, we consider a parametric study of adverse pressure gradient turbulent boundary layers where we restrict our attention to the pressure gradient parameter, β, the Reynolds number and the acceleration parameter, K. The statistics analyzed are limited to the streamwise fluctuating velocity. The data show that the mean velocity profile in strong pressure gradient boundary layers does not conform to the classical logarithmic law. Moreover, there appears to be no measurable logarithmic region in these cases. It is also found that the large-scale motions scaling with outer variables are energised by the pressure gradient. These increasingly strong large-scale motions are found to be the dominant contributor to the increase in turbulence intensity (scaled with friction velocity) with increasing pressure gradient across the boundary layer.

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

    Directory of Open Access Journals (Sweden)

    Donald M. McEligot

    2014-07-01

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

  3. Grain boundary cavity growth under applied stress and internal pressure

    International Nuclear Information System (INIS)

    Mancuso, J.F.

    1977-08-01

    The growth of grain boundary cavities under applied stress and internal gas pressure was investigated. Methane gas filled cavities were produced by the C + 4H reversible CH4 reaction in the grain boundaries of type 270 nickel by hydrogen charging in an autoclave at 500 0 C with a hydrogen pressure of either 3.4 or 14.5 MPa. Intergranular fracture of nickel was achieved at a charging temperature of 300 0 C and 10.3 MPa hydrogen pressure. Cavities on the grain boundaries were observed in the scanning electron microscope after fracture. Photomicrographs of the cavities were produced in stereo pairs which were analyzed so as to correct for perspective distortion and also to determine the orientational dependence of cavity growth under an applied tensile stress

  4. Failure internal pressure of spherical steel containments

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, G.

    1985-01-01

    An application of the British CEGB's R6 Failure Assessment Approach to the determination of failure internal pressure of nuclear power plant spherical steel containments is presented. The presence of hypothetical cracks both in the base metal and in the welding material of the containment, with geometrical idealizations according to the ASME Boiler and Pressure Vessel Code (Section XI), was taken into account in order to analyze the sensitivity of the failure assessment with the values of the material fracture properties. Calculations of the elastoplastic collapse load have been performed by means of the Finite Element System SAMCEF. The clean axisymmetric shell (neglecting the influence of nozzles and minor irregularities) and two major penetrations (personnel and emergency locks) have been taken separately into account. Large-strain elastoplastic behaviour of the material was considered in the Code, using lower bounds of true stress-true strain relations obtained by testing a collection of tensile specimens. Assuming the presence of cracks in non-perturbed regions, the reserve factor for test pressure and the failure internal pressure have been determined as a function of the flaw depth. (orig.)

  5. Pressure Fluctuations Induced by a Hypersonic Turbulent Boundary Layer

    Science.gov (United States)

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

    2016-01-01

    Direct numerical simulations (DNS) are used to examine the pressure fluctuations generated by a spatially-developed Mach 5.86 turbulent boundary layer. The unsteady pressure field is analyzed at multiple wall-normal locations, including those at the wall, within the boundary layer (including inner layer, the log layer, and the outer layer), and in the free stream. The statistical and structural variations of pressure fluctuations as a function of wall-normal distance are highlighted. Computational predictions for mean velocity pro les and surface pressure spectrum are in good agreement with experimental measurements, providing a first ever comparison of this type at hypersonic Mach numbers. The simulation shows that the dominant frequency of boundary-layer-induced pressure fluctuations shifts to lower frequencies as the location of interest moves away from the wall. The pressure wave propagates with a speed nearly equal to the local mean velocity within the boundary layer (except in the immediate vicinity of the wall) while the propagation speed deviates from the Taylor's hypothesis in the free stream. Compared with the surface pressure fluctuations, which are primarily vortical, the acoustic pressure fluctuations in the free stream exhibit a significantly lower dominant frequency, a greater spatial extent, and a smaller bulk propagation speed. The freestream pressure structures are found to have similar Lagrangian time and spatial scales as the acoustic sources near the wall. As the Mach number increases, the freestream acoustic fluctuations exhibit increased radiation intensity, enhanced energy content at high frequencies, shallower orientation of wave fronts with respect to the flow direction, and larger propagation velocity.

  6. Cavity pressure history of contained nuclear explosions

    Energy Technology Data Exchange (ETDEWEB)

    Chapin, C E [Lawrence Radiation Laboratory, University of California, Livermore, CA (United States)

    1970-05-01

    Knowledge of pressure in cavities created by contained nuclear explosions is useful for estimating the possibility of venting radioactive debris to the atmosphere. Measurements of cavity pressure, or temperature, would be helpful in evaluating the correctness of present code predictions of underground explosions. In instrumenting and interpreting such measurements it is necessary to have good theoretical estimates of cavity pressures. In this paper cavity pressure is estimated at the time when cavity growth is complete. Its subsequent decrease due to heat loss from the cavity to the surrounding media is also predicted. The starting pressure (the pressure at the end of cavity growth) is obtained by adiabatic expansion to the final cavity size of the vaporized rock gas sphere created by the explosion. Estimates of cavity size can be obtained by stress propagation computer codes, such as SOC and TENSOR. However, such estimates require considerable time and effort. In this paper, cavity size is estimated using a scheme involving simple hand calculations. The prediction is complicated by uncertainties in the knowledge of silica water system chemistry and a lack of information concerning possible blowoff of wall material during cavity growth. If wall material blows off, it can significantly change the water content in the cavity, compared to the water content in the ambient media. After cavity growth is complete, the pressure will change because of heat loss to the surrounding media. Heat transfer by convection, radiation and conduction is considered, and its effect on the pressure is calculated. Analysis of cavity heat transfer is made difficult by the complex nature of processes which occur at the wall where melting, vaporization and condensation of the gaseous rock can all occur. Furthermore, the melted wall material could be removed by flowing or dripping to the cavity floor. It could also be removed by expansion of the steam contained in the melt (blowoff) and by

  7. 16 CFR 1500.130 - Self-pressurized containers: labeling.

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Self-pressurized containers: labeling. 1500... § 1500.130 Self-pressurized containers: labeling. (a) Self-pressurized containers that fail to bear a...: warning—contents under pressure Do not puncture or incinerate container. Do not expose to heat or store at...

  8. Containment wells to form hydraulic barriers along site boundaries

    International Nuclear Information System (INIS)

    Vo, D.; Ramamurthy, A.S.; Qu, J.; Zhao, X.P.

    2008-01-01

    In the field, aquifer remediation methods include pump and treat procedures based on hydraulic control systems. They are used to reduce the level of residual contamination present in the soil and soil pores of aquifers. Often, physical barriers are erected along the boundaries of the target (aquifer) site to reduce the leakage of the released soil contaminant to the surrounding regions. Physical barriers are expensive to build and dismantle. Alternatively, based on simple hydraulic principles, containment wells or image wells injecting clear water can be designed and built to provide hydraulic barriers along the contaminated site boundaries. For brevity, only one pattern of containment well system that is very effective is presented in detail. The study briefly reports about the method of erecting a hydraulic barrier around a contaminated region based on the simple hydraulic principle of images. During the clean-up period, hydraulic barriers can considerably reduce the leakage of the released contaminant from the target site to surrounding pristine regions. Containment wells facilitate the formation of hydraulic barriers. Hence, they control the movement of contaminants away from the site that is being remedied. However, these wells come into play, only when the pumping operation for cleaning up the site is active. After operation, they can be filled with soil to permit the natural ground water movement. They can also be used as monitoring wells

  9. Development of pressure boundaries leak detection technology for nuclear reactor

    International Nuclear Information System (INIS)

    Zhang Yao; Zhang Dafa; Chen Dengke; Zhang Liming

    2008-01-01

    The leak detection for the pressure boundaries is an important safeguard in nuclear reactor operation. In the paper, the status and the characters on the development of the pressure boundaries leak detection technology for the nuclear reactor were reviewed, especially, and the advance of the radiation leak detection technology and the acoustic emission leak detection technology were analyzed. The new advance trend of the leak detection technology was primarily explored. According to the analysis results, it is point out that the advancing target of the leak detection technology is to enhance its response speed, sensitivity, and reliability, and to provide effective information for operator and decision-maker. The realization of the global leak detection and the whole life cycle health monitoring for the nuclear boundaries is a significant advancing tendency of the leak detection technology. (authors)

  10. LES of the adverse-pressure gradient turbulent boundary layer

    International Nuclear Information System (INIS)

    Inoue, M.; Pullin, D.I.; Harun, Z.; Marusic, I.

    2013-01-01

    Highlights: • The adverse-pressure gradient turbulent boundary layer at high Re is studied. • Wall-model LES works well for nonequilibrium turbulent boundary layer. • Relationship of skin-friction to Re and Clauser pressure parameter is explored. • Self-similarity is observed in the velocity statistics over a wide range of Re. -- Abstract: We describe large-eddy simulations (LES) of the flat-plate turbulent boundary layer in the presence of an adverse pressure gradient. The stretched-vortex subgrid-scale model is used in the domain of the flow coupled to a wall model that explicitly accounts for the presence of a finite pressure gradient. The LES are designed to match recent experiments conducted at the University of Melbourne wind tunnel where a plate section with zero pressure gradient is followed by section with constant adverse pressure gradient. First, LES are described at Reynolds numbers based on the local free-stream velocity and the local momentum thickness in the range 6560–13,900 chosen to match the experimental conditions. This is followed by a discussion of further LES at Reynolds numbers at approximately 10 times and 100 times these values, which are well out of range of present day direct numerical simulation and wall-resolved LES. For the lower Reynolds number runs, mean velocity profiles, one-point turbulent statistics of the velocity fluctuations, skin friction and the Clauser and acceleration parameters along the streamwise, adverse pressure-gradient domain are compared to the experimental measurements. For the full range of LES, the relationship of the skin-friction coefficient, in the form of the ratio of the local free-stream velocity to the local friction velocity, to both Reynolds number and the Clauser parameter is explored. At large Reynolds numbers, a region of collapse is found that is well described by a simple log-like empirical relationship over two orders of magnitude. This is expected to be useful for constant adverse-pressure

  11. Acoustic Emission for on-line reactor pressure boundary monitoring

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.

    1985-01-01

    The program objective is to develop AE for continuous surveillance to assess flaw growth in reactor pressure boundaries. Technology in the laboratory is being evaluated on structures. Results have demonstrated basic feasibility of the program objective. AE monitoring a long term fatigue test of a pressure vessel demonstrated an instrument system, and the ability to detect unexpected as well as well as known fatigue cracks. Monitoring a nuclear reactor system shows that the coolant flow noise problem is manageable and AE can be detected under simulated operating conditions

  12. CONTEMPT, LWR Containment Pressure and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Cheng, Teh-Chin; Wheat, L.L.; Mings, W.J.

    1991-01-01

    1 - Description of problem or function: CONTEMPT-LT was developed to predict the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. CONTEMPT-LT calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided for fan cooler and cooling spray engineered safety systems. One to four compartments can be modeled, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. The user determines the compartments to be used, specifies input mass and energy additions, defines heat structure and leakage systems, and prescribes the time advancement and output control. CONTEMPT-LT/28-H (NESC0433/08) includes also models for hydrogen combustion. 2 - Method of solution: The initial conditions of the containment atmosphere are calculated from input values, and the initial temperature distributions through the containment structures are determined from the steady-state solution of the heat conduction equations. A time advancement proceeds as follows. The input water and energy rates are evaluated at the midpoint of a time interval and added to the containment system. Pressure suppression, spray system effects, and fan cooler effects are calculated using conditions at the beginning of a time-step. Leakage and heat losses or gains, extrapolated from the last time-step, are added to the containment system. Containment volume pressure and temperature are estimated by solving the mass, volume, and energy balance equations. Using these results as boundary conditions, the heat conduction equations

  13. Electromagnetic stress at the boundary: Photon pressure or tension?

    Science.gov (United States)

    Wang, Shubo; Ng, Jack; Xiao, Meng; Chan, Che Ting

    2016-03-01

    It is well known that incident photons carrying momentum ℏk exert a positive photon pressure. But if light is impinging from a negative refractive medium in which ℏk is directed toward the source of radiation, should light exert a photon "tension" instead of a photon pressure? Using an ab initio method that takes the underlying microstructure of a material into account, we find that when an electromagnetic wave propagates from one material into another, the electromagnetic stress at the boundary is, in fact, indeterminate if only the macroscopic parameters are specified. Light can either pull or push the boundary, depending not only on the macroscopic parameters but also on the microscopic lattice structure of the polarizable units that constitute the medium. Within the context of an effective-medium approach, the lattice effect is attributed to electrostriction and magnetostriction, which can be accounted for by the Helmholtz stress tensor if we use the macroscopic fields to calculate the boundary optical stress.

  14. Pressure-induced transition in the grain boundary of diamond

    Science.gov (United States)

    Chen, J.; Tang, L.; Ma, C.; Fan, D.; Yang, B.; Chu, Q.; Yang, W.

    2017-12-01

    Equation of state of diamond powder with different average grain sizes was investigated using in situ synchrotron x-ray diffraction and a diamond anvil cell (DAC). Comparison of compression curves was made for two samples with average grain size of 50nm and 100nm. The two specimens were pre-pressed into pellets and loaded in the sample pressure chamber of the DAC separately to minimized differences of possible systematic errors for the two samples. Neon gas was used as pressure medium and ruby spheres as pressure calibrant. Experiments were conducted at room temperature and high pressures up to 50 GPa. Fitting the compression data in the full pressure range into the third order Birch-Murnaghan equation of state yields bulk modulus (K) and its pressure derivative (K') of 392 GPa and 5.3 for 50nm sample and 398GPa and 4.5 for 100nm sample respectively. Using a simplified core-shell grain model, this result indicates that the grain boundary has an effective bulk modulus of 54 GPa. This value is similar to that observed for carbon nanotube[1] validating the recent theoretical diamond surface modeling[2]. Differential analysis of the compression cures demonstrates clear relative compressibility change at the pressure about 20 GPa. When fit the compression data below and above this pressure separately, the effect of grain size on bulk modulus reverses in the pressure range above 20 GPa. This observation indicates a possible transition of grain boundary structure, likely from sp2 hybridization at the surface[2] towards sp3like orbital structure which behaves alike the inner crystal. [1] Jie Tang, Lu-Chang Qin, Taizo Sasaki, Masako Yudasaka, Akiyuki Matsushita, and Sumio Iijima, Compressibility and Polygonization of Single-Walled Carbon Nanotubes under Hydrostatic Pressure, Physical Review Letters, 85(9), 1187-1198, 2000. [2] Shaohua Lu, Yanchao Wang, Hanyu Liu, Mao-sheng Miao, and Yanming Ma, Self-assembled ultrathin nanotubes on diamond (100) surface, Nature

  15. 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

  16. 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.

  17. Transient pressure and productivity analysis in carbonate geothermal reservoirs with changing external boundary flux

    Directory of Open Access Journals (Sweden)

    Wang Dongying

    2017-01-01

    Full Text Available In this paper, a triple-medium flow model for carbonate geothermal reservoirs with an exponential external boundary flux is established. The pressure solution under constant production conditions in Laplace space is solved. The geothermal wellbore pressure change considering wellbore storage and skin factor is obtained by Stehfest numerical inversion. The well test interpretation charts and Fetkovich production decline chart for carbonate geothermal reservoirs are proposed for the first time. The proposed Fetkovich production decline curves are applied to analyze the production decline behavior. The results indicate that in carbonate geothermal reservoirs with exponential external boundary flux, the pressure derivative curve contains a triple dip, which represents the interporosity flow between the vugs or matrix and fracture system and the invading flow of the external boundary flux. The interporosity flow of carbonate geothermal reservoirs and changing external boundary flux can both slow down the extent of production decline and the same variation tendency is observed in the Fetkovich production decline curve.

  18. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  19. Pressure suppression device for a reactor container

    International Nuclear Information System (INIS)

    Shimizu, Toshiaki

    1982-01-01

    Purpose: To prevent damages in drain pipes or the likes upon the water level increase due to blowing of incompressible gases. Constitution: An exhaust pipe for guiding escaping steams is connected to a main steam releaf valve. The exhaust pipe is guided into pressure-suppression-chamber water through the inside of a dry-well and by way of a vent pipe, a vent header and a drain pipe or a downcomer. Since the exhaust pipe is not exposed to the water surface inside the pressure suppression chamber, even if steams blow out into the dry-well by the rapture of pipeways or the likes to rapidly increase the water level, the water surface does not hit on the exhaust pipe, whereby the damages for the exhaust pipe and support members can be prevented to improve the reliability. (Seki, T.)

  20. Containment for small pressurized water reactors

    International Nuclear Information System (INIS)

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

    1977-01-01

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

  1. Pressurized Water Reactor containment in Russia

    International Nuclear Information System (INIS)

    Taymouri, Majid.

    1993-01-01

    One of the most important systems of nuclear power plants from an economical point of view and view point of safety is containment; Therefore, the containments designed in Russia were studied in the first chapter. Russian general rules and requirements of structure of accident localization system were illustrated. Methods of accident localization system rooms tested for tightness and strength are presented in chapter three. Russian specialists have been working hard to ensure the safety culture in building structures and operational procedures and the have successfully implemented these objectives in new nuclear power plant designs and rules

  2. Ultimate pressure capacity of CANDU 6 containment structures

    International Nuclear Information System (INIS)

    Radulescu, J.P.; Pradolin, L.; Mamet, J.C.

    1997-01-01

    This paper summarizes the analytical work carried out and the results obtained when determining the ultimate pressure capacity (UPC) of the containment structures of CANDU 6 nuclear power plants. The purpose of the analysis work was to demonstrate that such containment structures are capable of meeting design requirements under the most severe accident conditions. For this concrete vessel subjected to internal pressure, the UPC was defined as the pressure causing through cracking in the concrete. The present paper deals with the overall behaviour of the containment. The presence of openings, penetrations and the ultimate pressure of the airlocks were considered separately. (author)

  3. Under pressure: Climate change, upwelling and eastern boundary upwelling ecosystems

    Directory of Open Access Journals (Sweden)

    Marisol eGarcía-Reyes

    2015-12-01

    Full Text Available The IPCC AR5 provided an overview of the likely effects of climate change on Eastern Boundary Upwelling Systems (EBUS, stimulating increased interest in research examining the issue. We use these recent studies to develop a new synthesis describing climate change impacts on EBUS. We find that model and observational data suggest coastal upwelling-favorable winds in poleward portions of EBUS have intensified and will continue to do so in the future. Although evidence is weak in data that are presently available, future projections show that this pattern might be driven by changes in the positioning of the oceanic high-pressure systems rather than by deepening of the continental low-pressure systems, as previously proposed. There is low confidence regarding the future effects of climate change on coastal temperatures and biogeochemistry due to uncertainty in the countervailing responses to increasing upwelling and coastal warming, the latter of which could increase thermal stratification and render upwelling less effective in lifting nutrient-rich deep waters into the photic zone. Although predictions of ecosystem responses are uncertain, EBUS experience considerable natural variability and may be inherently resilient. However, multi-trophic level, end-to-end (i.e., winds to whales studies are needed to resolve the resilience of EBUS to climate change, especially their response to long-term trends or extremes that exceed pre-industrial ranges.

  4. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  5. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  6. Experimental Investigation of Separated and Transitional Boundary Layers Under Low-Pressure Turbine Airfoil Conditions

    Science.gov (United States)

    Hultgren, Lennart S.; Volino, Ralph J.

    2002-01-01

    Modern low-pressure turbine airfoils are subject to increasingly stronger pressure gradients as designers impose higher loading in an effort to improve efficiency and to reduce part count. The adverse pressure gradients on the suction side of these airfoils can lead to boundary-layer separation, particularly under cruise conditions. Separation bubbles, notably those which fail to reattach, can result in a significant degradation of engine efficiency. Accurate prediction of separation and reattachment is hence crucial to improved turbine design. This requires an improved understanding of the transition flow physics. Transition may begin before or after separation, depending on the Reynolds number and other flow conditions, has a strong influence on subsequent reattachment, and may even eliminate separation. Further complicating the problem are the high free-stream turbulence levels in a real engine environment, the strong pressure gradients along the airfoils, the curvature of the airfoils, and the unsteadiness associated with wake passing from upstream stages. Because of the complicated flow situation, transition in these devices can take many paths that can coexist, vary in importance, and possibly also interact, at different locations and instances in time. The present work was carried out in an attempt to systematically sort out some of these issues. Detailed velocity measurements were made along a flat plate subject to the same nominal dimensionless pressure gradient as the suction side of a modern low-pressure turbine airfoil ('Pak-B'). The Reynolds number based on wetted plate length and nominal exit velocity, Re, was varied from 50;000 to 300; 000, covering cruise to takeoff conditions. Low, 0.2%, and high, 7%, inlet free-stream turbulence intensities were set using passive grids. These turbulence levels correspond to about 0.2% and 2.5% turbulence intensity in the test section when normalized with the exit velocity. The Reynolds number and free

  7. Ultimate internal pressure capacity of concrete containment structures

    International Nuclear Information System (INIS)

    Krishnaswamy, C.N.; Namperumal, R.; Al-Dabbagh, A.

    1983-01-01

    Lesson learned from the accident at Three-Mile Island nuclear plant has necessitated the computation of the ultimate internal pressure capacity of containment structures as a licensing requirement in the U.S. In general, a containment structure is designed to be essentially elastic under design accident pressure. However, as the containment pressure builds up beyond the design value due to a more severe postulated accident, the containment response turns nonlinear as it sequentially passes through cracking of concrete, yielding of linear plate, yielding of rebar, and yielding of post-tensioning tendon (if the containment concrete is prestressed). This paper reports on the determination of the ultimate internal pressure capacity and nonlinear behavior of typical reinforced and prestressed concrete BWR containments. The probable modes of failure, the criteria for ultimate pressure capacity, and the most critical sections are described. Simple equations to hand-calculate the ultimate pressure capacity and the nonlinear behavior at membrane sections of the containment shell are presented. A nonlinear finite element analysis performed to determine the nonlinear behavior of the entire shell including nonmembrane sections is briefly discribed. The analysis model consisted of laminated axisymmetric shell finite elements with nonlinear stress-strain properties for each material. Results presented for typical BWR concrete containments include nonlinear response plots of internal pressure versus containment deflection and strains in the liner, rebar, and post-tensioning tendons at the most stressed section in the shell. Leak-tightness of the containment liner and the effect of thermal loads on the ultimate capacity are discussed. (orig.)

  8. Grain boundary precipitation strengthening mechanism in W containing advanced creep resistant ferritic steels

    Energy Technology Data Exchange (ETDEWEB)

    Shibata, T.; Hasegawa, Y. [Tohoku Univ., Sendai (Japan)

    2010-07-01

    Grain boundary precipitation strengthening is expected to be a decisive factor in developing ferritic creep resistant steels. This study examined the grain boundary precipitation strengthening mechanism extracting the effect of the tempered martensitic microstructure and precipitates on the high angle grain boundary in M{sub 23}C4{sub 6} type carbide and the Fe{sub 2}W type Laves phase effect of the creep deformation fixing the grain boundary according to transmission electron microscope (TEM) observation. A creep test was carried out at high temperature in order to evaluate the high angle boundary strengthening effect simulating the long-term creep deformation microstructure by the lath structure disappearance. The correlation of the creep rupture time and the grain boundary shielding ratio were found to be independent of precipitate type. The creep deformation model represents block boundary shielding by precipitates as the decisive factor for W containing ferritic creep resistant steels. (orig.)

  9. 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

  10. Pressure release in containments of nuclear power stations

    International Nuclear Information System (INIS)

    Pauli, W.; Pellaud, B.; Saitoh, A.

    1992-01-01

    In France, Germany, Sweden and Switzerland, the licensing authorities have decided to equip nuclear reactor containments with a filter venting system to ensure survival of the containment after postulated severe nuclear accidents. This is a curious paradox. For years, the established wisdom was unambiguously 'Keep the containment tight. It's the ultimate barrier.' Three Mile Island seemed to prove the point. Yet, an old mechanical engineer's rule is 'Every pressure vessel must have a safety valve.' Filtered containment venting attempts to reconcile these two conflicting objectives by allowing a filtered pressure relief after an accident, in order to prevent containment failure due to overpressure, while keeping the release within acceptable limits. Achieving this dual objective is a matter of proper timing, i.e. pressure relief, not too early, not too late. (author)

  11. Designing high pressure containers for research- principles and applications

    International Nuclear Information System (INIS)

    Anandkumar, V.

    1997-01-01

    The high pressure scientist looks for a well engineered pressure apparatus for high pressure experiments for 1 kbar (0.1 GPa) and above. Often, a variety of difficulties including the choice of materials, design configuration, optimum utilisation of the strength of materials used in the design, are encountered. This article is intended to help the high pressure scientist to select the design approach for pressure retaining container. The limitations imposed by the strength of available materials and engineering standards in building high pressure containers are discussed. Engineering solutions to overcome these limitations with optimal utilisation of the strength of the materials are also discussed. Novel methods to boost up the pressure retaining capacity like multilayered design and autofrettaging are compared along with their relative advantages and disadvantages. Special methods by which it is possible to attain pressures which are several times the yield strength of the materials of construction are presented. In this aspects such as the basis of the codes and their relevance in the design of high pressure equipment will also be described. Discussions are centered around the methods to tackle situations where experimental constraints dictate requirements of pressures higher than those permitted by design codes. Safety features are also discussed. (author)

  12. Pressure test behaviour of embalse nuclear power plant containment structure

    International Nuclear Information System (INIS)

    Bruschi, S.; Marinelli, C.

    1984-01-01

    It's described the structural behaviour of the containment structure during the pressure test of the Embalse plant (CANDU type, 600MW), made of prestressed concrete with an epoxi liner. Displacement, strain, temperature, and pressure measurements of the containment structure of the Embalse Nuclear Power Plant are presented. The instrumentation set up and measurement specifications are described for all variables of interest before, during and after the pressure test. The analytical models to simulate the heat transfer due to sun heating and air convenction and to predict the associated thermal strains and displacements are presented. (E.G.) [pt

  13. Excess-pressure suppression device in a reactor container

    International Nuclear Information System (INIS)

    Nishio, Masahide

    1985-01-01

    Purpose: To reliably decrease the radioactivity of radioactive gases when they are released externally. Constitution: The exit of a gas exhaust pipe for discharging gases in a reactor container, on generation of an excess pressure in the reactor container upon loss of coolant accident, is adapted to be always fluided in the cooling tank. Then, the exhaust gases discharged in the cooling tank is realeased to the atmosphere. In this way, the excess pressure in the reactor container can be prevented previously and the radioactivity of the gases released externally is significantly reduced by the scrubbing effect. (Kamimura, M.)

  14. Pressure Indication of 3013 Inner Containers Using Digital Radiography

    International Nuclear Information System (INIS)

    HENSEL, SJ

    2004-01-01

    Plutonium bearing materials packaged for long term storage per the Department of Energy Standard 3013 (DOE-STD-3013) are required to be examined periodically in a non-destructive manner (i.e. without compromising the storage containers) for pressure buildup. Radiography is the preferred technology for performing the examinations. The concept is to measure and record the container lid position. As a can pressurizes the lid will deflect outward and thus provide an indication of the internal pressure. A radiograph generated within 30 days of creation of each storage container serves as the baseline from which future surveillance examinations will be compared. A problem with measuring the lid position was discovered during testing of a digital radiography system. The solution was to provide a distinct feature upon the lower surface of the container lid from which the digital radiography system could easily track the lid position

  15. Storage of hydrogen in advanced high pressure container. Appendices

    International Nuclear Information System (INIS)

    Bentzen, J.J.; Lystrup, A.

    2005-07-01

    The objective of the project has been to study barriers for a production of advanced high pressure containers especially suitable for hydrogen, in order to create a basis for a container production in Denmark. The project has primarily focused on future Danish need for hydrogen storage in the MWh area. One task has been to examine requirement specifications for pressure tanks that can be expected in connection with these stores. Six potential storage needs have been identified: (1) Buffer in connection with start-up/regulation on the power grid. (2) Hydrogen and oxygen production. (3) Buffer store in connection with VEnzin vision. (4) Storage tanks on hydrogen filling stations. (5) Hydrogen for the transport sector from 1 TWh surplus power. (6) Tanker transport of hydrogen. Requirements for pressure containers for the above mentioned use have been examined. The connection between stored energy amount, pressure and volume compared to liquid hydrogen and oil has been stated in tables. As starting point for production technological considerations and economic calculations of various container concepts, an estimation of laminate thickness in glass-fibre reinforced containers with different diameters and design print has been made, for a 'pure' fibre composite container and a metal/fibre composite container respectively. (BA)

  16. Analytical studies on optimization of containment design pressure

    International Nuclear Information System (INIS)

    Haware, S.K.; Ghosh, A.K.; Kushwaha, H.S.

    2005-01-01

    The containment of the proposed Advanced Heavy Water Reactor (AHWR) is divided into two main volumes viz. V1 and V2 interconnected by vent system via suppression pool. The arrangement is such that the volume V2 surrounds the volume V1 (see Fig.1). Blow Out Panels (BOPs), installed on volume V1 are designed to rupture at a differential pressure of 50 kPa. The containment was analysed using the in-house developed code CONTRAN, for three different scenario considered viz. (i) Loss of Coolant Accident (LOCA) involving double ended break in the downcomer pipe, (ii) LOCA involving double ended break in the reactor inlet header and (iii) Main Steam Line Break (MSLB) Accident. It was revealed that the accident involving the double-ended break of reactor inlet header results in the maximum value of the containment peak pressure. Results of the analyses indicated that the size of the BOP has bearing on the containment peak pressure. Therefore, five cases were analysed, varying the size of BOP from 0 to 10 m 2 , in order to quantify the influence of the size of BOP on the containment peak pressure. The blowdown mass and energy discharge data calculated using the code RELAP5/MOD3.2 was used in the analysis. It was observed that the vents are cleared in around 0.41 seconds into the accident. The containment peak pressures obtained in various cases are presented in Fig.2. The containment peak pressure varies with the size of BOP and passes through minima for a BOP size of around 5 m 2 . There are two flow processes, competing with each other viz. the steam-air mixture passage through the vent system via suppression pool and direct passage of steam air mixture through BOP bypassing the suppression pool. Though the energy suppression efficiency of the suppression pool decreases with increasing size of BOP, the pressure suppression efficiency was found to be maximum at around 5 m 2 size of BOP. The containment peak pressure passing through minima indicates that there is a scope for

  17. Nonstationary pressure build up in full-pressure containments after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1977-01-01

    The time histories of pressure, temperature and pressure difference during the pressure build up phase of a loss-of-coolant accident (LOCA) in the primary system in full-pressure containments of water cooled nuclear power reactors are treated. These are important for the design of such containments. The experiments within the German research program RS 50 ''Druckverteilung im Containment'' offered, for the first time, the opportunity to observe experimentally fluid-dynamic processes in a multiple divided full-pressure containment, and to test at the same time, computer codes which serve to describe the physical processes during the LOCA. The comparison of the results calculated by the computer codes ZOCO VI and DDIFF with the experimental results showed apparent deviations by special arrangements of the compartments and the vent flow paths of a model containment for the calculation of time dependent pressure-, temperature- and pressure difference-histories. The deviations lead to the development of the analytical model and computer code COFLOW. This new model was primarily designed to deal with the fluid-dynamic processes in the beginning phase of the blowdown as maximal pressure differences appear. Furthermore, it can be used to determine the maximum containment pressure, as well as for long term calculations. The analytical model and computer code COFLOW shows a better correlation between theory and experiment than previous codes

  18. 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.

  19. Application of pressure-sensitive paint in shock-boundary layer interaction experiments

    OpenAIRE

    Seivwright, Douglas L.

    1996-01-01

    Approved for public release; distribution is unlimited A new type of pressure transducer, pressure-sensitive paint, was used to obtain pressure distributions associated with shock-boundary layer interaction. Based on the principle of photoluminescence and the process of oxygen quenching, pressure-sensitive paint provides a continous mapping of a pressure field over a surface of interest. The data measurement and acquisition system developed for use with the photoluminescence sensor was eva...

  20. Risk analysis of in-service pressure piping containing defects

    International Nuclear Information System (INIS)

    Lin, Y.C.; Xie, Y.J.; Wang, X.H.; Luo, H.

    2004-01-01

    The reliability of pressure piping containing defects is important in engineering. The failure probability of pressure piping containing defects may be used as a guide to the most economic deployment of resources on maintenance, inspection and repair. This paper presents a probabilistic assessment methodology for in-service pressure piping containing defects, which is especially designed for programming. It is based on three assessment codes, BS 7910, R6 and SAPV-99, considering uncertainties in operating loadings, flaw sizes, material fracture toughness and flow stress. A general sampling computation method of stress intensity factor (SIF), in the form of the relationship between SIF and axial force and bending moment and torsion, is adopted. This relationship has been successfully used in developing software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), to assess planar and non-planar flaws. A numerical example is presented to illustrate the application of SAPP-2003 for calculating the failure probabilities of separate defects and for the assessed pressure piping

  1. Modified Pressure-Correction Projection Methods: Open Boundary and Variable Time Stepping

    KAUST Repository

    Bonito, Andrea

    2014-10-31

    © Springer International Publishing Switzerland 2015. In this paper, we design and study two modifications of the first order standard pressure increment projection scheme for the Stokes system. The first scheme improves the existing schemes in the case of open boundary condition by modifying the pressure increment boundary condition, thereby minimizing the pressure boundary layer and recovering the optimal first order decay. The second scheme allows for variable time stepping. It turns out that the straightforward modification to variable time stepping leads to unstable schemes. The proposed scheme is not only stable but also exhibits the optimal first order decay. Numerical computations illustrating the theoretical estimates are provided for both new schemes.

  2. Modified Pressure-Correction Projection Methods: Open Boundary and Variable Time Stepping

    KAUST Repository

    Bonito, Andrea; Guermond, Jean-Luc; Lee, Sanghyun

    2014-01-01

    © Springer International Publishing Switzerland 2015. In this paper, we design and study two modifications of the first order standard pressure increment projection scheme for the Stokes system. The first scheme improves the existing schemes in the case of open boundary condition by modifying the pressure increment boundary condition, thereby minimizing the pressure boundary layer and recovering the optimal first order decay. The second scheme allows for variable time stepping. It turns out that the straightforward modification to variable time stepping leads to unstable schemes. The proposed scheme is not only stable but also exhibits the optimal first order decay. Numerical computations illustrating the theoretical estimates are provided for both new schemes.

  3. BWR Mark III pressure suppression containment response to hydrogen deflagration

    International Nuclear Information System (INIS)

    Fuls, G.M.; Gunter, A.D.

    1982-01-01

    The CLASIX-3 computer program has been used to evaluate the temperature and pressure response of the BWR Mark III Suppression Containment System to hydrogen deflagration resulting from a degraded core condition. The CLASIX-3 computer program is an extension of the CLASIX program which was originally developed to analyze ice condenser containments. A brief description is given of the modifications made to CLASIX to increase its flexibility and versatility to include the capability of analyzing the Mark III Containment. Analytical results are presented for the two base case transients. The two base cases are the stuck open steam relief valve and the small break LOCA, both of which are assumed to lead to a degraded core condition and the release of hydrogen to the containment. Results include pressure and temperature response, gas concentrations and suppression pool response

  4. Containment pressure analysis model using CONTEMPT-LT

    International Nuclear Information System (INIS)

    Gupta, R.N.

    1975-09-01

    An analytical model for evaluating the reactor containment pressure transient following a loss-of-coolant accident (LOCA) is presented. The model uses the CONTEMPT-LT computer program developed by Aerojet Nuclear Company. The sample problem studied is the containment response following the most severe postulated LOCA at the Yankee Rowe Nuclear Power Station. The results show good agreement with the response predicted by Westinghouse Electric Corporation. (auth)

  5. 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)

  6. Formation of a Boundary-Free Dust Cluster in a Low-Pressure Gas-Discharge Plasma

    International Nuclear Information System (INIS)

    Usachev, A. D.; Zobnin, A. V.; Petrov, O. F.; Fortov, V. E.; Annaratone, B. M.; Thoma, M. H.; Hoefner, H.; Kretschmer, M.; Fink, M.; Morfill, G. E.

    2009-01-01

    An attraction between negatively charged micron-sized plastic particles was observed in the bulk of a low-pressure gas-discharge plasma under microgravity conditions. This attraction had led to the formation of a boundary-free dust cluster, containing one big central particle with a radius of about 6 μm and about 30 1 μm-sized particles situated on a sphere with a radius of 190 μm and with the big particle in the center. The stability of this boundary-free dust cluster was possible due to its confinement by the plasma flux on the central dust particle

  7. NEK containment integrated leak rate test at full pressure

    International Nuclear Information System (INIS)

    Skaler, F.; Planinc, V.; Gregoric, D.; Cicvaric, D.

    1999-01-01

    NPP Krsko is a Pressure Water Reactor (PWR) Plant which has four barriers to prevent release of radioactive fission products. These four barriers are following: Fuel itself, Fuel Clad, Reactor Coolant System and Containment Building. Containment is the last barrier which can prevent release of fission product when other barriers have been already broken. To find out the real condition of containment vessel and to prove its ability of withstanding increased parameters during accident we have to perform Containment Integrated Leak Rate Test at least three times in every ten years of operation. CILRT 1999 in NPP Krsko was completely performed following regulation of 10CFR50 App. J Option A and ANSI/ANS 56.8-1987. The main goal of CILRT is to prove that the leakage of containment pathways and wall structures are within limits prescribed in Technical Specifications by pressurization of containment building above peak accident pressure Pa and measuring the mass changes of air using Ideal Gas Law.(author)

  8. Evaluation of high-pressure containment buildings for LMFBR's

    International Nuclear Information System (INIS)

    Armstrong, G.R.

    1981-01-01

    A study was conducted on the use of High Pressure LMFBR Containment Buildings for 1000 MW(e) LMFBRs. Two principal aspects were investigated: accident consequence mitigation and cost. Two types of hypothetical accidents were analyzed to establish consequence mitigation: melt-through and energetic expulsion. Three Containment Building (CB) design pressures were investigated: 69 kPa (10 psig), 207 kPa (30 psig), and 414 kPa (60 psig). Four types of design structures were analyzed to establish cost: steel, steel with confinement building, reinforced concrete, and prestressed/post-tensioned concrete. Results show that: it is within reason that a high pressure containment for a 1000 MW(e) reactor can be fabricated that will retain its integrity during postulated severe hypothetical accidents, if available measures are taken to reduce or prevent hydrogen production and the cost differential between basic high (414 kPa) and low (69 kPa) pressure containments is $10 x 10 6 or less

  9. Vent clearing analysis of a Mark III pressure suppression containment

    International Nuclear Information System (INIS)

    Quintana, R.

    1979-01-01

    An analysis of the vent clearing transient in a Mark III pressure suppression containment after a hypothetical LOCA is carried out. A two-dimensional numerical model solving the transient fluid dynamic equations is used. The geometry of the pressure suppression pool is represented and the pressure and velocity fields in the pool are obtained from the moment the LOCA occurs until the first vent in the drywell wall clears. The results are compared to those obtained with the one-diemensional model used for containment design, with special interest on two-dimensional effects. Some conclusions concerning the effect of the water discharged into the suppression pool through the vents on submerged structures are obtained. Future improvements to the model are suggested. (orig.)

  10. Approximation for maximum pressure calculation in containment of PWR reactors

    International Nuclear Information System (INIS)

    Souza, A.L. de

    1989-01-01

    A correlation was developed to estimate the maximum pressure of dry containment of PWR following a Loss-of-Coolant Accident - LOCA. The expression proposed is a function of the total energy released to the containment by the primary circuit, of the free volume of the containment building and of the total surface are of the heat-conducting structures. The results show good agreement with those present in Final Safety Analysis Report - FSAR of several PWR's plants. The errors are in the order of ± 12%. (author) [pt

  11. Instability waves and transition in adverse-pressure-gradient boundary layers

    Science.gov (United States)

    Bose, Rikhi; Zaki, Tamer A.; Durbin, Paul A.

    2018-05-01

    Transition to turbulence in incompressible adverse-pressure-gradient (APG) boundary layers is investigated by direct numerical simulations. Purely two-dimensional instability waves develop on the inflectional base velocity profile. When the boundary layer is perturbed by isotropic turbulence from the free stream, streamwise elongated streaks form and may interact with the instability waves. Subsequent mechanisms that trigger transition depend on the intensity of the free-stream disturbances. All evidence from the present simulations suggest that the growth rate of instability waves is sufficiently high to couple with the streaks. Under very low levels of free-stream turbulence (˜0.1 % ), transition onset is highly sensitive to the inlet disturbance spectrum and is accelerated if the spectrum contains frequency-wave-number combinations that are commensurate with the instability waves. Transition onset and completion in this regime is characterized by formation and breakdown of Λ vortices, but they are more sporadic than in natural transition. Beneath free-stream turbulence with higher intensity (1-2 % ), bypass transition mechanisms are dominant, but instability waves are still the most dominant disturbances in wall-normal and spanwise perturbation spectra. Most of the breakdowns were by disturbances with critical layers close to the wall, corresponding to inner modes. On the other hand, the propensity of an outer mode to occur increases with the free-stream turbulence level. Higher intensity free-stream disturbances induce strong streaks that favorably distort the boundary layer and suppress the growth of instability waves. But the upward displacement of high amplitude streaks brings them to the outer edge of the boundary layer and exposes them to ambient turbulence. Consequently, high-amplitude streaks exhibit an outer-mode secondary instability.

  12. Initial boundary-value problem for the spherically symmetric Einstein equations with fluids with tangential pressure.

    Science.gov (United States)

    Brito, Irene; Mena, Filipe C

    2017-08-01

    We prove that, for a given spherically symmetric fluid distribution with tangential pressure on an initial space-like hypersurface with a time-like boundary, there exists a unique, local in time solution to the Einstein equations in a neighbourhood of the boundary. As an application, we consider a particular elastic fluid interior matched to a vacuum exterior.

  13. Temperature and stress distribution in pressure vessel by the boundary element method

    International Nuclear Information System (INIS)

    Alujevic, A.; Apostolovic, D.

    1990-01-01

    The aim of this paper is to demonstrate the applicability of boundary element method for the solution of temperatures and thermal stresses in the body of reactor pressure vessel of the NPP Krsko . In addition to the theory of boundary elements for thermo-elastic continua (2D, 3D) results are given of a numerically evaluated meridional cross-section. (author)

  14. Pressurized solid oxide fuel cell integral air accumular containment

    Science.gov (United States)

    Gillett, James E.; Zafred, Paolo R.; Basel, Richard A.

    2004-02-10

    A fuel cell generator apparatus contains at least one fuel cell subassembly module in a module housing, where the housing is surrounded by a pressure vessel such that there is an air accumulator space, where the apparatus is associated with an air compressor of a turbine/generator/air compressor system, where pressurized air from the compressor passes into the space and occupies the space and then flows to the fuel cells in the subassembly module, where the air accumulation space provides an accumulator to control any unreacted fuel gas that might flow from the module.

  15. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

    Hensel, S.

    2012-01-01

    Transportation and storage of plutonium oxide is typically done using a convenience container to hold the oxide powder which is then placed inside a containment vessel. Intermediate containers which act as uncredited confinement barriers may also be used. The containment vessel is subject to an internal pressure due to several sources including; (1) plutonium oxide provides a heat source which raises the temperature of the gas space, (2) helium generation due to alpha decay of the plutonium, (3) hydrogen generation due to radiolysis of the water which has been adsorbed onto the plutonium oxide, and (4) degradation of plastic bags which may be used to bag out the convenience can from a glove box. The contributions of these sources are evaluated in a reasonably conservative manner.

  16. A Sharp-Interface Immersed Boundary Method with Improved Mass Conservation and Reduced Spurious Pressure Oscillations.

    Science.gov (United States)

    Seo, Jung Hee; Mittal, Rajat

    2011-08-10

    A method for reducing the spurious pressure oscillations observed when simulating moving boundary flow problems with sharp-interface immersed boundary methods (IBMs) is proposed. By first identifying the primary cause of these oscillations to be the violation of the geometric conservation law near the immersed boundary, we adopt a cut-cell based approach to strictly enforce geometric conservation. In order to limit the complexity associated with the cut-cell method, the cut-cell based discretization is limited only to the pressure Poisson and velocity correction equations in the fractional-step method and the small-cell problem tackled by introducing a virtual cell-merging technique. The method is shown to retain all the desirable properties of the original finite-difference based IBM while at the same time, reducing pressure oscillations for moving boundaries by roughly an order of magnitude.

  17. Leakage of pressurized gases through unlined concrete containment structures

    International Nuclear Information System (INIS)

    Rizkalla, S.H.; Simmonds, S.H.

    1983-01-01

    Eight reinforced concrete specimens were fabricated and subjected to tensile membrane forces and air pressure to study the air leakage characteristics in cracked reinforced concrete members. A mathematical expression for the rate of pressurized air flowing through an idealized crack is presented. The mathematical expression is refined by using the experimental data to describe the air flow rate through any given crack pattern. Graphical charts are also presented for the calculation of the air leakage rate through concrete cracks. The concept of equivalent crack width for a given crack pattern is introduced. The mathematical expression and graphical charts are modified to include this equivalent crack width concept. The proposed technique is applicable for the prediction of the leakage from concrete containment structures or any similar structures due to high internal pressure sufficient to initiate cracking. (orig.)

  18. A Psychodynamic Perspective of Workplace Bullying: Containment, Boundaries and a Futile Search for Recognition

    Science.gov (United States)

    White, Sheila

    2004-01-01

    This paper presents a psychodynamic perspective of workplace bullying. It focuses on two related psychoanalytical concepts, containment and boundaries. The life cycle theory of bullying builds on these concepts and describes in-depth the evolving relationship between a bully and a victim. The search for recognition by the bully and victim proves…

  19. Pressure estimation from single-snapshot tomographic PIV in a turbulent boundary layer

    NARCIS (Netherlands)

    Schneiders, J.F.G.; Pröbsting, S.; Dwight, R.P.; Van Oudheusden, B.W.; Scarano, F.

    2016-01-01

    A method is proposed to determine the instantaneous pressure field from a single tomographic PIV velocity snapshot and is applied to a flat-plate turbulent boundary layer. The main concept behind the single-snapshot pressure evaluation method is to approximate the flow acceleration using the

  20. Loads on EPR containment after RPV failure at high pressure

    International Nuclear Information System (INIS)

    Jacobs, G.

    1995-01-01

    As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

  1. Wall-pressure fluctuations beneath a spatially evolving turbulent boundary layer

    Science.gov (United States)

    Mahesh, Krishnan; Kumar, Praveen

    2016-11-01

    Wall-pressure fluctuations beneath a turbulent boundary layer are important in applications dealing with structural deformation and acoustics. Simulations are performed for flat plate and axisymmetric, spatially evolving zero-pressure-gradient turbulent boundary layers at inflow Reynolds number of 1400 and 2200 based on momentum thickness. The simulations generate their own inflow using the recycle-rescale method. The results for mean velocity and second-order statistics show excellent agreement with the data available in literature. The spectral characteristics of wall-pressure fluctuations and their relation to flow structure will be discussed. This work is supported by ONR.

  2. Tuning of turbulent boundary layer anisotropy for improved surface pressure and trailing-edge noise modeling

    DEFF Research Database (Denmark)

    Bertagnolio, Franck; Fischer, Andreas; Zhu, Wei Jun

    2014-01-01

    The modeling of the surface pressure spectrum beneath a turbulent boundary layer is investigated, focusing on the case of airfoil flows and associated trailing edge noise prediction using the so-called TNO model. This type of flow is characterized by the presence of an adverse pressure gradient...... along the airfoil chord. It is shown that discrepancies between measurements and results from the TNO model increase as the pressure gradient increases. The original model is modified by introducing anisotropy in the definition of the turbulent vertical velocity spectrum across the boundary layer...

  3. Behaviour of concrete containment under over-pressure conditions

    International Nuclear Information System (INIS)

    Atchison, R.J.; Asmis, G.J.K.; Campbell, F.R.

    1979-01-01

    The Atomic Energy Control Board of Canada initiated June, 1975, a major study of the behaviour of concrete containment under over-pressure conditions. Although extensive theoretical and experimental work has been carried out for thick-walled Prestressed Concrete Reactor Vessels (PCRV's), there is a want of information on the non-linear response of thin-walled structures typical of the CANDU, 600 MW(e) cylindrical/spherical, post-tensioned containment shells. The purpose of this paper is to provide an overview of the total program, to present the reasons behind the research contract, and the specification and implementation of the work. The results of the theoretical and experimental work and their implications with respect to Canadian Concrete Containment practice are discussed. This study is unique, and, as far as is known, has no world-wide precedence. (orig.)

  4. Hydrodynamic pressure in a tank containing two liquids

    International Nuclear Information System (INIS)

    Tang, Yu.

    1992-01-01

    A study on the dynamic response of a tank containing two different liquids under seismic excitation is presented. Both analytical and numerical (FEM) methods are employed in the analysis. The results obtained by the two methods are in good agreement. The response functions examined include the hydrodynamic pressure, base shear and base moments. A simple approach that can be used to estimate the fundamental natural frequency of the tank-liquid system containing two liquids is proposed. This simple approach is an extension of the method used for estimating the frequency of a tank-liquid system containing only one liquid. This study shows that the dynamic response of a tank filled with two liquids is quite different from that of an identical tank filled with only one liquid

  5. Process-based investigation of cross-boundary environmental pressure from urban household consumption

    International Nuclear Information System (INIS)

    Yang, Dewei; Lin, Yanjie; Gao, Lijie; Sun, Yanwei; Wang, Run; Zhang, Guoqin

    2013-01-01

    Sustainability research at the city scale is increasingly focusing on urban household consumption in the context of global climate change. We use a complementary emergy accounting (EMA) and carbon footprint accounting (CFA) method to investigate the environmental pressure generated by household consumption in Xiamen, China. We distinguish between the resource extraction, consumption and disposal stages within an urban spatial conceptual framework, comprising the Urban Footprint Region (UFR) and Urban Sprawl Region (USR), and analyze five environmental footprint categories associated with cross-boundary household emergy and carbon flows. Cross-boundary activities, which link the USR with its UFR, contributed nearly 90% of total emergy and 70% of total GHG emissions in CFA. Transport fuel, building materials and food contribute most to environmental pressure in both EMA and CFA. The results indicate a significant cross-boundary resource burden and environmental footprint associated with household activities. The employed framework, method, and scope challenge the conventional spatial boundary of the urban system, and the results have important policy implications for urban sustainability and cross-boundary environmental management. - Highlights: ► We propose an urban spatial conceptual framework that includes USR and UFRs. ► A complementary EMA and CFA method is employed in urban household consumption system. ► Process-based cross-boundary environmental pressure of household consumption are evaluated. ► USR exerts pressure on its UFRs by extensive resource extraction and environmental emissions. ► We elucidate the USR–UFR environmental relationships and household energy policy

  6. Boundary element analysis of earthquake induced hydrodynamic pressures in a water reservoir

    International Nuclear Information System (INIS)

    Jablonski, A.M.

    1988-11-01

    The seismic analysis of concrete gravity and arch dams is affected by the hydrodynamic pressures in the water reservoir. Boundary element method (BEM) formulations are derived for the hydrodynamic pressures arising in a gravity dam-reservoir-foundation system, treating both 2- and 3-dimensional cases. The formulations are based on the respective mathematical models which are governed by two- and three-dimensional Helmholtz equations with appropriate boundary conditions. For infinite reservoirs, loss of energy due to pressure waves moving away toward infinity strongly influence response. Since it is not possible to discretize an infinite extent, the radiation damping due to outgoing waves is accounted for by incorporating special boundary conditions at the far end, and in a similar manner the loss of energy due to absorption of waves by a flexible bottom of reservoir and banks can be accounted for by a special condition along the boundaries. Numerical results are obtained and compared with available classical solutions and convergence of numerical results with the size and number of boundary elements is studied. It is concluded that the direct boundary element method is an effective tool for the evaluation of the hydrodynamic pressures in finite and infinite dam-reservoir-foundation systems subjected to harmonic-type motion, and can easily be extended to any type of random motion with fast Fourier transform techniques. 82 refs., 65 figs., 25 tabs

  7. Assessment of Mechanisms Impacting N-Nitrosodimethylamine Fate Within the North Boundary Containment System, Rocky Mountain Arsenal

    National Research Council Canada - National Science Library

    Gunnison, Douglas

    1997-01-01

    .... Chemical analyses by both RMA and Shell Chemical have detected N-nitrosodimethylamine (NDMA), also known as dimethylnitrosamine, within the groundwater around the North Boundary Containment System...

  8. A Rotational Pressure-Correction Scheme for Incompressible Two-Phase Flows with Open Boundaries

    Science.gov (United States)

    Dong, S.; Wang, X.

    2016-01-01

    Two-phase outflows refer to situations where the interface formed between two immiscible incompressible fluids passes through open portions of the domain boundary. We present several new forms of open boundary conditions for two-phase outflow simulations within the phase field framework, as well as a rotational pressure correction based algorithm for numerically treating these open boundary conditions. Our algorithm gives rise to linear algebraic systems for the velocity and the pressure that involve only constant and time-independent coefficient matrices after discretization, despite the variable density and variable viscosity of the two-phase mixture. By comparing simulation results with theory and the experimental data, we show that the method produces physically accurate results. We also present numerical experiments to demonstrate the long-term stability of the method in situations where large density contrast, large viscosity contrast, and backflows occur at the two-phase open boundaries. PMID:27163909

  9. 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.

  10. A probability model for the failure of pressure containing parts

    International Nuclear Information System (INIS)

    Thomas, H.M.

    1978-01-01

    The model provides a method of estimating the order of magnitude of the leakage failure probability of pressure containing parts. It is a fatigue based model which makes use of the statistics available for both specimens and vessels. Some novel concepts are introduced but essentially the model simply quantifies the obvious i.e. that failure probability increases with increases in stress levels, number of cycles, volume of material and volume of weld metal. A further model based on fracture mechanics estimates the catastrophic fraction of leakage failures. (author)

  11. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Yen-Shu, E-mail: yschen@iner.org.t [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China); Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon [Nuclear Engineering Division, Institute of Nuclear Energy Research, 1000, Wenhua Rd., Jiaan Village, Longtan Township, Taoyuan County 32546, Taiwan (China)

    2011-05-15

    Research highlights: The Chinshan Mark I containment pressure-temperature responses are analyzed. GOTHIC is used to calculate the containment responses under three pipe break events. This study is used to support the Chinshan Stretch Power Uprate (SPU) program. The calculated peak pressure and temperature are still below the design values. The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 {sup o}C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 {sup o}C). Additionally, the peak drywell temperature of 155.3 {sup o}C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 {sup o}C, which is below the pool temperature used for evaluating the

  12. Pressure and temperature analyses using GOTHIC for Mark I containment of the Chinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che; Lin, Yon-Pon

    2011-01-01

    Research highlights: → The Chinshan Mark I containment pressure-temperature responses are analyzed. → GOTHIC is used to calculate the containment responses under three pipe break events. → This study is used to support the Chinshan Stretch Power Uprate (SPU) program. → The calculated peak pressure and temperature are still below the design values. → The Chinshan containment integrity can be maintained under SPU condition. - Abstract: Chinshan Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/4 plant with original licensed thermal power (OLTP) of 1775 MWt for each unit. Recently, the Stretch Power Uprate (SPU) program for the Chinshan plant is being conducted to uprate the core thermal power to 1858 MWt (104.66% OLTP). In this study, the Chinshan Mark I containment pressure/temperature responses during LOCA at 105% OLTP (104.66% OLTP + 0.34% OLTP power uncertainty = 105% OLTP) are analyzed using the containment thermal-hydraulic program GOTHIC. Three kinds of LOCA (Loss of Coolant Accident) scenarios are investigated: Recirculation Line Break (RCLB), Main Steam Line Break (MSLB), and Feedwater Line Break (FWLB). In the short-term analyses, blowdown data generated by RELAP5 transient analyses are provided as boundary conditions to the GOTHIC containment model. The calculated peak drywell pressure and temperature in the RCLB event are 217.2 kPaG and 137.1 o C, respectively, which are close to the original FSAR results (219.2 kPaG and 138.4 o C). Additionally, the peak drywell temperature of 155.3 o C calculated by MSLB is presented in this study. To obtain the peak suppression pool temperature, a long-term RCLB analysis is performed using a simplified RPV (Reactor Pressure Vessel) volume to calculate blowdown flow rate. One RHR (Residual Heat Removal) heat exchanger is assumed to be inoperable for suppression pool cooling mode. The calculated peak suppression pool temperature is 93.2 o C, which is below the pool temperature used for evaluating the

  13. Prevention of bolting degradation or failure in pressure boundary and support applications

    International Nuclear Information System (INIS)

    Merrick, E.A.; Rivers, A.; Bickford, J.; Marston, T.U.

    1986-01-01

    A discussion is presented of bolting degradation or failure experience in pressure boundary and component support applications in US commercial nuclear plants and the industry program to prevent failures in the future. The focus turns to steps which plant owners can take today to guard against pressure boundary bolt failure or degradation for existing plants or units being constructed. 'Tools' or products which the plant owner can expect from current industry programs which will be available in the near future to aid in understanding and improving bolting practices are described. (author)

  14. Pressure tuning of the morphotropic phase boundary in piezoelectric lead zirconate titanate

    International Nuclear Information System (INIS)

    Rouquette, J.; Haines, J.; Bornand, V.; Pintard, M.; Papet, Ph.; Bousquet, C.; Konczewicz, L.; Gorelli, F. A.; Hull, S.

    2004-01-01

    Titanium-rich PZT solid solutions were studied under high pressure by neutron and x-ray diffraction, Raman spectroscopy and dielectric measurements. The results show that high pressure stabilizes the ferroelectric monoclinic phases, which are proposed to be responsible for the high piezoelectric properties characteristic of the morphotropic composition PbZr 0.52 Ti 0.48 O 3 . Pressure may thus be used to tune the morphotropic phase boundary in the composition-pressure plane to include a wide range of titanium-rich PZT compositions

  15. Ultimate internal pressure capacity of a reinforced concrete Mark III containment

    International Nuclear Information System (INIS)

    McGaughy, J.P. Jr.; Lin, F.T.; Sen, S.K.

    1983-01-01

    The static ultimate capacity of a Mark III BWR pressure suppression type containment has been investigated with a view to determine its capability to withstand the internal pressure associated with a postulated hydrogen burn. The reinforced concrete containment consists of a right circular cylinder covered by a hemispherical dome and supported on a flat circular foundation mat. A 1/4'' thick welded steel liner plate covers the inside surface of the containment shell. The cylinder is a 3.5 ft. thick shell with an inside radius of 62.0 feet. The thickness of the dome is 3.5 feet. Reinforcement in the shell is comprised of multi-layers of circumferential, meridional and diagonal rebars. Major containment penetrations consists of a circular equipment hatch and two personnel airlock assemblies. The containment ultimate capacity is determined by performing a non-linear analysis using the proprietary finite element computer code 'FINEL'. The code has the capability of modelling concrete cracking in tension and redistribution forces and moments to account for such phenomenon. For analysis purposes, the finite element model included the containment dome and the upper portion of the containment cylinder with appropriate boundary conditions applied at the model cut off region. This portion of the containment structure is selected because the segment of the cylinder that is included in the model has the least amount of hopp reinforcement, and when the general yield state is reached, the hoop reinforcement will be the limiting element. The containment structure has been treated as an axisymmetric shell using axisymmetric quadrilateral finite elements in the radial plane to model the liner plate and concrete. The reinforcing steel have been idealized by finite elements with unidirectional stiffness. (orig./RW)

  16. DNS of transcritical turbulent boundary layers at supercritical pressures under abrupt variations in thermodynamic properties

    Science.gov (United States)

    Kawai, Soshi

    2014-11-01

    In this talk, we first propose a numerical strategy that is robust and high-order accurate for enabling to simulate transcritical flows at supercritical pressures under abrupt variations in thermodynamic properties due to the real fluid effects. The method is based on introducing artificial density diffusion in a physically-consistent manner in order to capture the steep variation of thermodynamic properties in transcritical conditions robustly, while solving a pressure evolution equation to achieve pressure equilibrium at the transcritical interfaces. We then discuss the direct numerical simulation (DNS) of transcritical heated turbulent boundary layers on a zero-pressure-gradient flat plate at supercritical pressures. To the best of my knowledge, the present DNS is the first DNS of zero-pressure-gradient flat-plate transcritical turbulent boundary layer. The turbulent kinetic budget indicates that the compressibility effects (especially, pressure-dilatation correlation) are not negligible at the transcritical conditions even if the flow is subsonic. The unique and interesting interactions between the real fluid effects and wall turbulence, and their turbulence statistics, which have never been seen in the ideal-fluid turbulent boundary layers, are also discussed. This work was supported in part by Japan Society for the Promotion of Science (JSPS) Grant-in-Aid for Young Scientists (A) KAKENHI 26709066 and the JAXA International Top Young Fellowship Program.

  17. Acoustic emission and estimation of flaw significance in reactor pressure boundaries

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.

    1982-01-01

    The work discussed is intended to establish the feasibility of using acoustic emission (AE) to detect and evaluate growing flaws in nuclear reactor pressure boundaries. Basic AE identification and interpretation methods have grown out of Phase 1. Phases 2 and 3 to test and demonstrate developed methodology on a vessel test and on a reactor are in progress

  18. Local characteristics of the nocturnal boundary layer in response to external pressure forcing

    NARCIS (Netherlands)

    van der Linden, S.J.A.; Baas, P.; van Hooft, J.A.; van Hooijdonk, I.G.S.; Bosveld, F.C.; van de Wiel, B.J.H.

    2017-01-01

    Geostrophic wind speed data, derived from pressure observations, are used in combination with tower measurements to investigate the nocturnal stable boundary layer at Cabauw (The Netherlands). Since the geostrophic wind speed is not directly influenced by local nocturnal stability, it may be

  19. Hypersonic Wind-Tunnel Measurements of Boundary-Layer Pressure Fluctuations

    Science.gov (United States)

    2009-08-01

    Fluctuation Cone The Pressure-Fluctuation Cone was used for all wind-tunnel tests (Figure 3.7). The model is a 7◦ half-angle stainless - steel cone. It...analysis as a medium for fault detection: A review. Journal of Tribology , 130, January 2008. [80] L. M. Mack. Boundary layer linear stability theory. In

  20. A preliminary investigation of boundary-layer transition along a flat plate with adverse pressure gradient

    Science.gov (United States)

    Von Doenhoff, Albert E

    1938-01-01

    Boundary-layer surveys were made throughout the transition region along a smooth flat plate placed in an airstream of practically zero turbulence and with an adverse pressure gradient. The boundary-layer Reynolds number at the laminar separation point was varied from 1,800 to 2,600. The test data, when considered in the light of certain theoretical deductions, indicated that transition probably began with separation of the laminar boundary layer. The extent of the transition region, defined as the distance from a calculated laminar separation point to the position of the first fully developed turbulent boundary-layer profile, could be expressed as a constant Reynolds number run of approximately 70,000. Some speculations are presented concerning the application of the foregoing concepts, after certain assumptions have been made, to the problem of the connection between transition on the upper surface of an airfoil at high angles of attack and the maximum lift.

  1. Numerical study of ambient pressure for laser-induced bubble near a rigid boundary

    Science.gov (United States)

    Li, BeiBei; Zhang, HongChao; Han, Bing; Lu, Jian

    2012-07-01

    The dynamics of the laser-induced bubble at different ambient pressures was numerically studied by Finite Volume Method (FVM). The velocity of the bubble wall, the liquid jet velocity at collapse, and the pressure of the water hammer while the liquid jet impacting onto the boundary are found to increase nonlinearly with increasing ambient pressure. The collapse time and the formation time of the liquid jet are found to decrease nonlinearly with increasing ambient pressure. The ratios of the jet formation time to the collapse time, and the displacement of the bubble center to the maximal radius while the jet formation stay invariant when ambient pressure changes. These ratios are independent of ambient pressure.

  2. Effect of copper precipitates on the toughness of low alloy steels for pressure boundary components

    International Nuclear Information System (INIS)

    Foehl, J.; Willer, D.; Katerbau, K.H.

    2004-01-01

    The ferritic bainitic steel 15NiCuMoNb5 (WB 36)is widely used for pressure boundary components. Due to the high copper content which leads to precipitation hardening high strength and toughness are characteristic for this type of steel. However, in the initial state, there is still a high amount of dissolved copper in an oversaturated state which makes the steel susceptible to thermal ageing. Ageing and annealing experiments were performed, and the change in microstructure was investigated by small angle neutron scattering (SANS), measurements of the residual electric resistance and hardness measurements. A correlation between micro structural changes and changes in mechanical properties could be established. It could clearly be shown that significant effects on strength and toughness have to be considered when the size of the copper rich precipitates vary in the range from 1.2 to 2.2 nm in radius. The changes in microstructure affect both, the Carpy impact transition temperature and the fracture toughness qualitatively and quantitatively in a similar way. The investigations have contributed to a better understanding of precipitation hardening by copper not only for this type of steel but also for copper containing steels and weld subjected to neutron irradiation. (orig.)

  3. Modelling of HTR Confinement Behaviour during Accidents Involving Breach of the Helium Pressure Boundary

    Directory of Open Access Journals (Sweden)

    Joan Fontanet

    2009-01-01

    Full Text Available Development of HTRs requires the performance of a thorough safety study, which includes accident analyses. Confinement building performance is a key element of the system since the behaviour of aerosol and attached fission products within the building is of an utmost relevance in terms of the potential source term to the environment. This paper explores the available simulation capabilities (ASTEC and CONTAIN codes and illustrates the performance of a postulated HTR vented confinement under prototypical accident conditions by a scoping study based on two accident sequences characterized by Helium Pressure Boundary breaches, a small and a large break. The results obtained indicate that both codes predict very similar thermal-hydraulic responses of the confinement both in magnitude and timing. As for the aerosol behaviour, both codes predict that most of the inventory coming into the confinement is eventually depleted on the walls and only about 1% of the aerosol dust is released to the environment. The crosscomparison of codes states that largest differences are in the intercompartmental flows and the in-compartment gas composition.

  4. Boundary layers affected by different pressure gradients investigated computationally by a zonal RANS-LES method

    International Nuclear Information System (INIS)

    Roidl, B.; Meinke, M.; Schröder, W.

    2014-01-01

    Highlights: • Reformulated synthetic turbulence generation method (RSTGM) is applied. • Zonal RANS-LES method is applied to boundary layers at pressure gradients. • Good agreement with the pure LES and other reference data is obtained. • The RSTGM is applicable to pressure gradient flows without modification. • RANS-to-LES boundary should be located where -1·10 6 6 is satisfied. -- Abstract: The reformulated synthetic turbulence generation (RSTG) method is used to compute by a fully coupled zonal RANS-LES approach turbulent non-zero-pressure gradient boundary layers. The quality of the RSTG method, which is based on the same shape functions and length scale distributions as in zero-pressure gradient flow, is discussed by comparing the zonal RANS-LES findings with pure LES, pure RANS, direct numerical simulation (DNS), and experimental data. For the favorable pressure gradient (FPG) simulation the RANS-to-LES transition occurs in the accelerated flow region and for the adverse pressure gradient (APG) case it is located in the decelerated flow region. The results of the time and spanwise averaged skin-friction distributions, velocity profiles, and Reynolds stress distributions of the zonal RANS-LES simulation show a satisfactory to good agreement with the pure LES, reference DNS, and experimental data. The quality of the findings shows that the rigorous formulation of the synthetic turbulence generation makes the RSTG method applicable without a priori knowledge of the flow properties but those determined by the RANS solution and without using additional control planes to regulate the shear stress budget to a wide range of Reynolds numbers and pressure gradients. The method is a promising approach to formulate embedded RANS-to-LES boundaries in flow regions where the Pohlhausen or acceleration parameter satisfies -1·10 -6 ⩽K⩽2·10 -6

  5. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    Energy Technology Data Exchange (ETDEWEB)

    Dai, Liang-Che, E-mail: lcdai@iner.gov.tw; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-12-15

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  6. Short-term pressure and temperature MSLB response analyses for large dry containment of the Maanshan nuclear power station

    International Nuclear Information System (INIS)

    Dai, Liang-Che; Chen, Yen-Shu; Yuann, Yng-Ruey

    2014-01-01

    Highlights: • The GOTHIC code is used for the PWR dry containment pressure and temperature analysis. • Boundary conditions are hot standby and 102% power main steam line break accidents. • Containment pressure and temperature responses of GOTHIC are similar with FSAR. • The capability of the developed model to perform licensing calculation is assessed. - Abstract: Units 1 and 2 of the Maanshan nuclear power station are the typical Westinghouse three-loop PWR (pressurized water reactor) with large dry containments. In this study, the containment analysis program GOTHIC is adopted for the dry containment pressure and temperature analysis. Free air space and sump of the PWR dry containment are individually modeled as control volumes. The containment spray system and fan cooler unit are also considered in the GOTHIC model. The blowdown mass and energy data of the main steam line break (hot standby condition and various reactor thermal power levels) are tabulated in the Maanshan Final Safety Analysis Report (FSAR) 6.2 which could be used as the boundary conditions for the containment model. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR results. In this study, hot standby and 102% reactor thermal power main steam line break accidents are selected. The calculated peak containment pressure is 323.50 kPag (46.92 psig) for hot standby MSLB, which is a little higher than the FSAR value of 311.92 kPag (45.24 psig). But it is still below the design value of 413.69 kPag (60 psig). The calculated peak vapor temperature inside the containment is 187.0 °C (368.59 F) for 102% reactor thermal power MSLB, which is lower than the FSAR result of 194.42 °C (381.95 F). The effects of the containment spray system and fan cooler units could be clearly observed in the GOTHIC analysis. The calculated containment pressure and temperature behaviors of the selected cases are in good agreement with the FSAR

  7. Problems and chances for probabilistic fracture mechanics in the analysis of steel pressure boundary reliability

    Energy Technology Data Exchange (ETDEWEB)

    Staat, M [Forschungszentrum Juelich GmbH (Germany). Inst. fuer Sicherheitsforschung und Reaktortechnik

    1996-12-01

    It is shown that the difficulty for probabilistic fracture mechanics (PFM) is the general problem of the high reliability of a small population. There is no way around the problem as yet. Therefore what PFM can contribute to the reliability of steel pressure boundaries is demonstrated with the example of a typical reactor pressure vessel and critically discussed. Although no method is distinguishable that could give exact failure probabilities, PFM has several additional chances. Upper limits for failure probability may be obtained together with trends for design and operating conditions. Further, PFM can identify the most sensitive parameters, improved control of which would increase reliability. Thus PFM should play a vital role in the analysis of steel pressure boundaries despite all shortcomings. (author). 19 refs, 7 figs, 1 tab.

  8. Problems and chances for probabilistic fracture mechanics in the analysis of steel pressure boundary reliability

    International Nuclear Information System (INIS)

    Staat, M.

    1996-01-01

    It is shown that the difficulty for probabilistic fracture mechanics (PFM) is the general problem of the high reliability of a small population. There is no way around the problem as yet. Therefore what PFM can contribute to the reliability of steel pressure boundaries is demonstrated with the example of a typical reactor pressure vessel and critically discussed. Although no method is distinguishable that could give exact failure probabilities, PFM has several additional chances. Upper limits for failure probability may be obtained together with trends for design and operating conditions. Further, PFM can identify the most sensitive parameters, improved control of which would increase reliability. Thus PFM should play a vital role in the analysis of steel pressure boundaries despite all shortcomings. (author). 19 refs, 7 figs, 1 tab

  9. Analysis of Numerical Simulation Database for Pressure Fluctuations Induced by High-Speed Turbulent Boundary Layers

    Science.gov (United States)

    Duan, Lian; Choudhari, Meelan M.

    2014-01-01

    Direct numerical simulations (DNS) of Mach 6 turbulent boundary layer with nominal freestream Mach number of 6 and Reynolds number of Re(sub T) approximately 460 are conducted at two wall temperatures (Tw/Tr = 0.25, 0.76) to investigate the generated pressure fluctuations and their dependence on wall temperature. Simulations indicate that the influence of wall temperature on pressure fluctuations is largely limited to the near-wall region, with the characteristics of wall-pressure fluctuations showing a strong temperature dependence. Wall temperature has little influence on the propagation speed of the freestream pressure signal. The freestream radiation intensity compares well between wall-temperature cases when normalized by the local wall shear; the propagation speed of the freestream pressure signal and the orientation of the radiation wave front show little dependence on the wall temperature.

  10. Pressure sensor to determine spatial pressure distributions on boundary layer flows

    Science.gov (United States)

    Sciammarella, Cesar A.; Piroozan, Parham; Corke, Thomas C.

    1997-03-01

    The determination of pressures along the surface of a wind tunnel proves difficult with methods that must introduce devices into the flow stream. This paper presents a sensor that is part of the wall. A special interferometric reflection moire technique is developed and used to produce signals that measures pressure both in static and dynamic settings. The sensor developed is an intelligent sensor that combines optics and electronics to analyze the pressure patterns. The sensor provides the input to a control system that is capable of modifying the shape of the wall and preserve the stability of the flow.

  11. Minimum containment pressure and its effect on ECCS performance of APR-1400

    International Nuclear Information System (INIS)

    Kim, In Goo; Bang, Young S.; Kim, Hho Jung

    2004-01-01

    The containment pressure has a strong effect on the late reheat behavior for a large break LOCA, associated with the DVI issue. The downcomer boiling, which occurs during the post-reflood phase, has a negative effect on core cooling for a LBLOCA. Because the downcomer boiling is enhanced as the containment pressure decreases, how to determine containment pressure is important to the evaluation of ECCS performance. In spite of its importance of containment pressure, there are few studies on the containment pressure and the interaction between RCS and containment thermal hydraulics. To have a better knowledge of the effect of containment pressure on APR-1400 ECCS performance, a parametric study for containment pressure has been carried out. Also, the interaction between RCS and containment behavior has been also investigated

  12. Direct numerical simulation of thermally-stratified turbulent boundary layer subjected to adverse pressure gradient

    International Nuclear Information System (INIS)

    Hattori, Hirofumi; Kono, Amane; Houra, Tomoya

    2016-01-01

    Highlights: • We study various thermally-stratified turbulent boundary layers having adverse pressure gradient (APG) by means of DNS. • The detailed turbulent statistics and structures in various thermally-stratified turbulent boundary layers having APG are discussed. • It is found that the friction coefficient and Stanton number decrease along the streamwise direction due to the effects of stable thermal stratification and APG, but those again increase due to the APG effect in the case of weak stable thermal stratification. • In the case of strong stable stratification with or without APG, the flow separation is observed in the downstream region. - Abstract: The objective of this study is to investigate and observe turbulent heat transfer structures and statistics in thermally-stratified turbulent boundary layers subjected to a non-equilibrium adverse pressure gradient (APG) by means of direct numerical simulation (DNS). DNSs are carried out under conditions of neutral, stable and unstable thermal stratifications with a non-equilibrium APG, in which DNS results reveal heat transfer characteristics of thermally-stratified non-equilibrium APG turbulent boundary layers. In cases of thermally-stratified turbulent boundary layers affected by APG, heat transfer performances increase in comparison with a turbulent boundary layer with neutral thermal stratification and zero pressure gradient (ZPG). Especially, it is found that the friction coefficient and Stanton number decrease along the streamwise direction due to the effects of stable thermal stratification and APG, but those again increase due to the APG effect in the case of weak stable thermal stratification (WSBL). Thus, the analysis for both the friction coefficient and Stanton number in the case of WSBL with/without APG is conducted using the FIK identity in order to investigate contributions from the transport equations, in which it is found that both Reynolds-shear-stress and the mean convection terms

  13. Non-adiabatic pressure loss boundary condition for modelling turbocharger turbine pulsating flow

    International Nuclear Information System (INIS)

    Chiong, M.S.; Rajoo, S.; Romagnoli, A.; Costall, A.W.; Martinez-Botas, R.F.

    2015-01-01

    Highlights: • Bespoke non-adiabatic pressure loss boundary for pulse flow turbine modelling. • Predictions show convincing results against experimental and literature data. • Predicted pulse pressure propagation is in good agreement with literature data. • New methodology is time efficient and requires minimal geometrical inputs. - Abstract: This paper presents a simplified methodology of pulse flow turbine modelling, as an alternative over the meanline integrated methodology outlined in previous work, in order to make its application to engine cycle simulation codes much more straight forward. This is enabled through the development of a bespoke non-adiabatic pressure loss boundary to represent the turbine rotor. In this paper, turbocharger turbine pulse flow performance predictions are presented along with a comparison of computation duration against the previously established integrated meanline method. Plots of prediction deviation indicate that the mass flow rate and actual power predictions from both methods are highly comparable and are reasonably close to experimental data. However, the new boundary condition required significantly lower computational time and rotor geometrical inputs. In addition, the pressure wave propagation in this simplified unsteady turbine model at different pulse frequencies has also been found to be in agreement with data from the literature, thereby supporting the confidence in its ability to simulate the wave action encountered in turbine pulse flow operation

  14. Bruce Power's nuclear pressure boundary quality assurance program requirements, implementation and transition

    International Nuclear Information System (INIS)

    Krane, J.C.

    2009-01-01

    The development of a full scope nuclear pressure boundary quality assurance program in Canada requires extensive knowledge of the structure and detailed requirements of codes and standards published by the Canadian Standards Association (CSA) and American Society of Mechanical Engineers (ASME). Incorporation into company governance documents and implementation of these requirements while managing the transition to more recent revisions of these codes and standards represents a significant challenge for Bruce Power, Canada's largest independent nuclear operator. This paper explores the key developments and innovative changes that are used to ensure successful regulatory compliance and effective implementation of the Bruce Power Pressure Boundary Quality Assurance Program. Challenges and mitigating strategies to sustain this large compliance based program at Bruce Power's 8 unit nuclear power plant site will also be detailed. (author)

  15. Pressure and tension waves from bubble collapse near a solid boundary: A numerical approach.

    Science.gov (United States)

    Lechner, Christiane; Koch, Max; Lauterborn, Werner; Mettin, Robert

    2017-12-01

    The acoustic waves being generated during the motion of a bubble in water near a solid boundary are calculated numerically. The open source package OpenFOAM is used for solving the Navier-Stokes equation and extended to include nonlinear acoustic wave effects via the Tait equation for water. A bubble model with a small amount of gas is chosen, the gas obeying an adiabatic law. A bubble starting from a small size with high internal pressure near a flat, solid boundary is studied. The sequence of events from bubble growth via axial microjet formation, jet impact, annular nanojet formation, torus-bubble collapse, and bubble rebound to second collapse is described. The different pressure and tension waves with their propagation properties are demonstrated.

  16. CONTEMPT-4MOD3, LWR Containment Long-Term Pressure Distribution and Temperature Distribution in LOCA

    International Nuclear Information System (INIS)

    Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.; Ng, K.K.; Mirsky, S.M.

    2002-01-01

    term transients such as are encountered during degraded core accidents with hydrogen combustion. The user has the option of turning off the implicit routine through user input, if desired. 2 - Method of solution: Containment thermodynamic conditions of hydrogen/air/steam/liquid water mixtures are determined by using modularized equation-of-state subroutines and tabulated water properties. The numeric in the code are completely explicit except for the predictor-corrector technique used to estimate the heat structure effects on compartment conditions, an implicit calculation of junction flow with inertia, and an optional implicit routine for junction flow calculation approaching pressure equilibrium. 3 - Restrictions on the complexity of the problem: Maxima of: 999 lumped parameter compartments, 99 heat conducting structures using a variety of heat transfer options and boundary conditions. Inter-compartment flow junctions may be calculated for either a sharp-edge orifice (single phase homogeneous or two-phase flow) or a nozzle (vapor flow only). Containment cooling spray analytical models are provided for either single or double heat exchangers

  17. Description of the containment for a stationary pressurized water reactor

    International Nuclear Information System (INIS)

    Hermani, S.

    1986-01-01

    The function of the containment is to prevent the inadvertent release of radioactive fission products from the reactor coolant system to the atmosphere and to provide biological shielding during both normal and accident operation. Basically three different containment concepts 1) the dry containment, 2) the subatmospheric containment, and 3) the ice condenser containment, have been developed, based on how the accident energy release from the reactor coolant system is controlled. The containment structure can be either 1) reinforced concrets with inside liner, 2) prestressed concrete with inside, or 3) full steel cylinder or steel sphere with separate concrete shield. The size of the containment is largely dictated by the required net free volume, that satisfies the energy release criteria due to the design basic accident. The design and construction methods applied to this structure guarantee that the containment will carry out its safety function. This was proven by the Three Mile Island accident. (author)

  18. A two pressure-velocity approach for immersed boundary methods in three dimensional incompressible flows

    International Nuclear Information System (INIS)

    Sabir, O; Ahmad, Norhafizan; Nukman, Y; Tuan Ya, T M Y S

    2013-01-01

    This paper describes an innovative method for computing fluid solid interaction using Immersed boundary methods with two stage pressure-velocity corrections. The algorithm calculates the interactions between incompressible viscous flows and a solid shape in three-dimensional domain. The fractional step method is used to solve the Navier-Stokes equations in finite difference schemes. Most of IBMs are concern about exchange of the momentum between the Eulerian variables (fluid) and the Lagrangian nodes (solid). To address that concern, a new algorithm to correct the pressure and the velocity using Simplified Marker and Cell method is added. This scheme is applied on staggered grid to simulate the flow past a circular cylinder and study the effect of the new stage on calculations cost. To evaluate the accuracy of the computations the results are compared with the previous software results. The paper confirms the capacity of new algorithm for accurate and robust simulation of Fluid Solid Interaction with respect to pressure field

  19. Safety analysis of high pressure gasous fuel container punctures

    Energy Technology Data Exchange (ETDEWEB)

    Swain, M.R. [Univ. of Miami, Coral Gables, FL (United States)

    1995-09-01

    The following report is divided into two sections. The first section describes the results of ignitability tests of high pressure hydrogen and natural gas leaks. The volume of ignitable gases formed by leaking hydrogen or natural gas were measured. Leaking high pressure hydrogen produced a cone of ignitable gases with 28{degrees} included angle. Leaking high pressure methane produced a cone of ignitable gases with 20{degrees} included angle. Ignition of hydrogen produced larger overpressures than did natural gas. The largest overpressures produced by hydrogen were the same as overpressures produced by inflating a 11 inch child`s balloon until it burst.

  20. Pressure induced phase transitions in ceramic compounds containing tetragonal zirconia

    Energy Technology Data Exchange (ETDEWEB)

    Sparks, R.G.; Pfeiffer, G.; Paesler, M.A.

    1988-12-01

    Stabilized tetragonal zirconia compounds exhibit a transformation toughening process in which stress applied to the material induces a crystallographic phase transition. The phase transition is accompanied by a volume expansion in the stressed region thereby dissipating stress and increasing the fracture strength of the material. The hydrostatic component of the stress required to induce the phase transition can be investigated by the use of a high pressure technique in combination with Micro-Raman spectroscopy. The intensity of Raman lines characteristic for the crystallographic phases can be used to calculate the amount of material that has undergone the transition as a function of pressure. It was found that pressures on the order of 2-5 kBar were sufficient to produce an almost complete transition from the original tetragonal to the less dense monoclinic phase; while a further increase in pressure caused a gradual reversal of the transition back to the original tetragonal structure.

  1. Free-boundary Full-pressure Island Healing in a Stellarator: Coil-healing

    International Nuclear Information System (INIS)

    Hudson, S.R.; Reiman, A.; Strickler, D.; Brooks, A.; Monticello, D.A.; Hirshman, S.P.

    2002-01-01

    The lack of axisymmetry in stellarators guarantees that in general magnetic islands and chaotic magnetic field lines will exist. As particle transport is strongly tied to the magnetic field lines, magnetic islands and chaotic field lines result in poor plasma confinement. For stellarators to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux-surfaces, and the suppression of magnetic islands is a critical issue for stellarator coil design, particularly for small aspect ratio devices. A procedure for modifying stellarator coil designs to eliminate magnetic islands in free-boundary full-pressure magnetohydrodynamic equilibria is presented. Islands may be removed from coil-plasma free-boundary equilibria by making small changes to the coil geometry and also by variation of trim coil currents. A plasma and coil design relevant to the National Compact Stellarator Experiment is used to illustrate the technique

  2. Implicit Large-Eddy Simulations of Zero-Pressure Gradient, Turbulent Boundary Layer

    Science.gov (United States)

    Sekhar, Susheel; Mansour, Nagi N.

    2015-01-01

    A set of direct simulations of zero-pressure gradient, turbulent boundary layer flows are conducted using various span widths (62-630 wall units), to document their influence on the generated turbulence. The FDL3DI code that solves compressible Navier-Stokes equations using high-order compact-difference scheme and filter, with the standard recycling/rescaling method of turbulence generation, is used. Results are analyzed at two different Re values (500 and 1,400), and compared with spectral DNS data. They show that a minimum span width is required for the mere initiation of numerical turbulence. Narrower domains ((is) less than 100 w.u.) result in relaminarization. Wider spans ((is) greater than 600 w.u.) are required for the turbulent statistics to match reference DNS. The upper-wall boundary condition for this setup spawns marginal deviations in the mean velocity and Reynolds stress profiles, particularly in the buffer region.

  3. New Models for Velocity/Pressure-Gradient Correlations in Turbulent Boundary Layers

    Science.gov (United States)

    Poroseva, Svetlana; Murman, Scott

    2014-11-01

    To improve the performance of Reynolds-Averaged Navier-Stokes (RANS) turbulence models, one has to improve the accuracy of models for three physical processes: turbulent diffusion, interaction of turbulent pressure and velocity fluctuation fields, and dissipative processes. The accuracy of modeling the turbulent diffusion depends on the order of a statistical closure chosen as a basis for a RANS model. When the Gram-Charlier series expansions for the velocity correlations are used to close the set of RANS equations, no assumption on Gaussian turbulence is invoked and no unknown model coefficients are introduced into the modeled equations. In such a way, this closure procedure reduces the modeling uncertainty of fourth-order RANS (FORANS) closures. Experimental and direct numerical simulation data confirmed the validity of using the Gram-Charlier series expansions in various flows including boundary layers. We will address modeling the velocity/pressure-gradient correlations. New linear models will be introduced for the second- and higher-order correlations applicable to two-dimensional incompressible wall-bounded flows. Results of models' validation with DNS data in a channel flow and in a zero-pressure gradient boundary layer over a flat plate will be demonstrated. A part of the material is based upon work supported by NASA under award NNX12AJ61A.

  4. Correlations for modeling transitional boundary layers under influences of freestream turbulence and pressure gradient

    International Nuclear Information System (INIS)

    Suluksna, Keerati; Dechaumphai, Pramote; Juntasaro, Ekachai

    2009-01-01

    This paper presents mathematical expressions for two significant parameters which control the onset location and length of transition in the γ-Re θ transition model of Menter et al. [Menter, F.R., Langtry, R.B., Volker, S., Huang, P.G., 2005. Transition modelling for general purpose CFD codes. In: ERCOFTAC International Symposium on Engineering Turbulence Modelling and Measurements]. The expressions are formulated and calibrated by means of numerical experiments for predicting transitional boundary layers under the influences of freestream turbulence and pressure gradient. It was also found that the correlation for transition momentum thickness Reynolds number needs only to be expressed in terms of local turbulence intensity, so that the more complex form that includes pressure gradient effects is unnecessary. Transitional boundary layers on a flat plate both with and without pressure gradients are employed to assess the performance of these two expressions for predicting the transition. The results show that the proposed expressions can work well with the model of Menter et al. (2005)

  5. Derivation of Zagarola-Smits scaling in zero-pressure-gradient turbulent boundary layers

    Science.gov (United States)

    Wei, Tie; Maciel, Yvan

    2018-01-01

    This Rapid Communication derives the Zagarola-Smits scaling directly from the governing equations for zero-pressure-gradient turbulent boundary layers (ZPG TBLs). It has long been observed that the scaling of the mean streamwise velocity in turbulent boundary layer flows differs in the near surface region and in the outer layer. In the inner region of small-velocity-defect boundary layers, it is generally accepted that the proper velocity scale is the friction velocity, uτ, and the proper length scale is the viscous length scale, ν /uτ . In the outer region, the most generally used length scale is the boundary layer thickness, δ . However, there is no consensus on velocity scales in the outer layer. Zagarola and Smits [ASME Paper No. FEDSM98-4950 (1998)] proposed a velocity scale, U ZS=(δ1/δ ) U∞ , where δ1 is the displacement thickness and U∞ is the freestream velocity. However, there are some concerns about Zagarola-Smits scaling due to the lack of a theoretical base. In this paper, the Zagarola-Smits scaling is derived directly from a combination of integral, similarity, and order-of-magnitude analysis of the mean continuity equation. The analysis also reveals that V∞, the mean wall-normal velocity at the edge of the boundary layer, is a proper scale for the mean wall-normal velocity V . Extending the analysis to the streamwise mean momentum equation, we find that the Reynolds shear stress in ZPG TBLs scales as U∞V∞ in the outer region. This paper also provides a detailed analysis of the mass and mean momentum balance in the outer region of ZPG TBLs.

  6. An optimal control method for fluid structure interaction systems via adjoint boundary pressure

    Science.gov (United States)

    Chirco, L.; Da Vià, R.; Manservisi, S.

    2017-11-01

    In recent year, in spite of the computational complexity, Fluid-structure interaction (FSI) problems have been widely studied due to their applicability in science and engineering. Fluid-structure interaction systems consist of one or more solid structures that deform by interacting with a surrounding fluid flow. FSI simulations evaluate the tensional state of the mechanical component and take into account the effects of the solid deformations on the motion of the interior fluids. The inverse FSI problem can be described as the achievement of a certain objective by changing some design parameters such as forces, boundary conditions and geometrical domain shapes. In this paper we would like to study the inverse FSI problem by using an optimal control approach. In particular we propose a pressure boundary optimal control method based on Lagrangian multipliers and adjoint variables. The objective is the minimization of a solid domain displacement matching functional obtained by finding the optimal pressure on the inlet boundary. The optimality system is derived from the first order necessary conditions by taking the Fréchet derivatives of the Lagrangian with respect to all the variables involved. The optimal solution is then obtained through a standard steepest descent algorithm applied to the optimality system. The approach presented in this work is general and could be used to assess other objective functionals and controls. In order to support the proposed approach we perform a few numerical tests where the fluid pressure on the domain inlet controls the displacement that occurs in a well defined region of the solid domain.

  7. A new and self-contained presentation of the theory of boundary operators for slit diffraction and their logarithmic approximations

    Energy Technology Data Exchange (ETDEWEB)

    Gorenflo, Norbert [Beuth Hochschule fuer Technik Berlin (Germany). Fachbereich II; Kunik, Matthias [Magdeburg Univ. (Germany). Inst. fuer Analysis und Numerik

    2009-07-01

    We present a new and self-contained theory for mapping properties of the boundary operators for slit diffraction occurring in Sommerfeld's diffraction theory, covering two different cases of the polarisation of the light. This theory is entirely developed in the context of the boundary operators with a Hankel kernel and not based on the corresponding mixed boundary value problem for the Helmholtz equation. For a logarithmic approximation of the Hankel kernel we also study the corresponding mapping properties and derive explicit solutions together with certain regularity results. (orig.)

  8. Effect of air content and mass inflow on the pressure rise in a containment during blowdown

    International Nuclear Information System (INIS)

    Marshall, J.; Holland, P.G.

    1977-01-01

    Experiments were made to investigate conditions arising during blowdown of a vessel filled with saturated steam/water at 7 MPa pressure into a containment vessel. The initial air pressure in the containment vessel was varied from one atmosphere to near vacuum. The initial water content of the high pressure vessel was varied. Pressure and temperature distributions were measured during the blowdown transient and compared with calculations based on a simple lumped-parameter model. The effect of condensation heat transfer on the containment pressure is discussed and attention drawn to the inadequacy of most available data. (Author)

  9. Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

    International Nuclear Information System (INIS)

    Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

    1990-03-01

    Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective was to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems. 2 refs., 4 figs., 5 tabs

  10. Risk-based priorities for inspection of nuclear pressure boundary components at selected LWRs

    International Nuclear Information System (INIS)

    Vo, T.V.; Simonen, F.A.; Gore, B.F.; Doctor, S.R.; Smith, B.W.

    1990-01-01

    Data from existing probabilistic risk assessments for eight representative nuclear power plants were used to identify and prioritize the most relevant systems to plant safety. The objective of this paper is to assess current in-service inspection requirements for pressure boundary systems and components, and to develop recommendations for improvements. This study demonstrates the feasibility of using risk-based methods to develop plant-specific inspection plans. Results for the eight representative plants also indicate generic trends that suggest improvements in current inspection plans now based on priorities set in accordance with code definitions of Class 1, 2, and 3 systems

  11. Environment sensitive cracking in pressure boundary materials of light water reactors

    International Nuclear Information System (INIS)

    Hanninen, H.; Aho-Mantila, I.; Torronen, K.

    1987-08-01

    A review of the various forms of environment sensitive cracking in pressure boundary materials of light water reactors is presented. The available methods and the most promising future possibilities of preventive maintenance to counteract the environmental degradation are evaluated. 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 strength Ni-base alloys as well as on corrosion fatigue of low alloy and stainless steels. Additionally, some general ideas on how to predict, reduce, monitor or eliminate environment sensitive cracking in service are presented

  12. A low pressure filter system for new containment concepts

    Energy Technology Data Exchange (ETDEWEB)

    Dillmann, H.G.; Pasler, H. [Kernforschungszentrum Karlsruhe GmbH Laboratorium fuer Isotopentechnik, Karlsruhe (Germany)

    1995-02-01

    It is demonstrated that after severe accidents the decay heat can be removed in a passive mode in a convective flow, i.e. without needing a fan. The filter components with sufficiently low pressure drop values which are required for this purpose will be described and the results indicated.

  13. Study on effective prestressing effects on concrete containment under the design-basis pressure condition

    International Nuclear Information System (INIS)

    Sun Feng; Pan Rong; Wang Lu; Mao Huan; Yang Yu

    2013-01-01

    Prestressing technology is widely used in nuclear power plant containment building, and the durability of containment structure is affected directly by the distribution and loss of prestressing value under design-basis pressure. Containment structure and the distribution of prestressing system are introduced briefly. Furthermore, the calculating process of horizontal prestressing bunch loss near the equipment hatch hole is put forward in details, and the containment structure prestressing loss when 5-year pressure test is obtained. Based above analysis, the finite element model of the prestressed concrete containment structure is built by using ANSYS code, the prestressing effect on concrete containment is analysed. The results show that most of the design pressure is bore by the prestressing system under the design-basis pressure, so the containment structure is safe. These conclusions are consistent with prestressing containment system design concepts, which can provide reference to the engineering staff. (authors)

  14. Active Brownian particles near straight or curved walls: Pressure and boundary layers

    Science.gov (United States)

    Duzgun, Ayhan; Selinger, Jonathan V.

    2018-03-01

    Unlike equilibrium systems, active matter is not governed by the conventional laws of thermodynamics. Through a series of analytic calculations and Langevin dynamics simulations, we explore how systems cross over from equilibrium to active behavior as the activity is increased. In particular, we calculate the profiles of density and orientational order near straight or circular walls and show the characteristic width of the boundary layers. We find a simple relationship between the enhancements of density and pressure near a wall. Based on these results, we determine how the pressure depends on wall curvature and hence make approximate analytic predictions for the motion of curved tracers, as well as the rectification of active particles around small openings in confined geometries.

  15. Analysis of grain boundaries, twin boundaries, and Te precipitates in CdZnTe grown by high-pressure Bridgeman method

    International Nuclear Information System (INIS)

    Heffelfinger, J.R.; Medlin, D.L.; James, R.B.

    1998-03-01

    Grain boundaries and twin boundaries in commercial Cd 1-x Zn x Te, which is prepared by a high pressure Bridgeman technique, have been investigated with transmission electron microscopy, scanning electron microscopy, infrared light microscopy and visible light microscopy. Boundaries inside these materials were found to be decorated with Te precipitates. The shape and local density of the precipitates were found to depend on the particular boundary. For precipitates that decorate grain boundaries, their microstructure was found to consist of a single, saucer shaped grain of hexagonal Te (space group P3 1 21). Analysis of a Te precipitate precipitates by selected area diffraction revealed the Te to be aligned with the surrounding Cd 1-x Zn x Te grains. This alignment was found to match the (111) Cd 1-x Z x Te planes with the (1 bar 101) planes of hexagonal Te. Crystallographic alignments between the Cd 1-x Zn x Te grains were also observed for a high angle grain boundary. The structure of the grain boundaries and the Te/Cd 1-x Zn x Te interface are discussed

  16. Containers, particularly prestressed concrete pressure vessels for nuclear reactor plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.; Mitterbacher, P.

    1986-01-01

    Pressure and temperature changes act on the liner, which cause differential expansion between the liner and the prestressed concrete. So that there will be no overload or damage to the liner, its anchoring or the concrete structure, cutouts are provided in the concrete at deflection positions of the steel cladding, connections and penetrations. These cut-outs are filled with inserts made of elastic or plastic material. (DG) [de

  17. Analysis of events related to cracks and leaks in the reactor coolant pressure boundary

    Energy Technology Data Exchange (ETDEWEB)

    Ballesteros, Antonio, E-mail: Antonio.Ballesteros-Avila@ec.europa.eu [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Sanda, Radian; Peinador, Miguel; Zerger, Benoit [JRC-IET: Institute for Energy and Transport of the Joint Research Centre of the European Commission, Postbus 2, NL-1755 ZG Petten (Netherlands); Negri, Patrice [IRSN: Institut de Radioprotection et de Sûreté Nucléaire (France); Wenke, Rainer [GRS: Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH (Germany)

    2014-08-15

    Highlights: • The important role of Operating Experience Feedback is emphasised. • Events relating to cracks and leaks in the reactor coolant pressure boundary are analysed. • A methodology for event investigation is described. • Some illustrative results of the analysis of events for specific components are presented. - Abstract: The presence of cracks and leaks in the reactor coolant pressure boundary may jeopardise the safe operation of nuclear power plants. Analysis of cracks and leaks related events is an important task for the prevention of their recurrence, which should be performed in the context of activities on Operating Experience Feedback. In response to this concern, the EU Clearinghouse operated by the JRC-IET supports and develops technical and scientific work to disseminate the lessons learned from past operating experience. In particular, concerning cracks and leaks, the studies carried out in collaboration with IRSN and GRS have allowed to identify the most sensitive areas to degradation in the plant primary system and to elaborate recommendations for upgrading the maintenance, ageing management and inspection programmes. An overview of the methodology used in the analysis of cracks and leaks related events is presented in this paper, together with the relevant results obtained in the study.

  18. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    The primary mission of the Containment Group is to ensure that underground nuclear tests are satisfactorily contained. The main goal is the development of sound technical bases for containment-related methodology. Major areas of activity include siting, geologic description, emplacement hole stemming, and phenomenological predictions. Performance results of sanded gypsum concrete plugs on the Jefferson, Panamint, Cornucopia, Labquark, and Bodie events are given. Activities are also described in the following areas: computational capabilities site description, predictive modeling, and cavity-pressure measurement. Containment publications are listed. 8 references

  19. Pressure-temperature response of a full-pressure PWR containment to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Misak, J.

    1976-01-01

    A mathematical model and computer code TRACO III for pressure-temperature transients in the full-pressure containment of PWR during LOCA is described. Main attention is devoted to the analysis of parametric calculations with respect to the estimation of effect of various factors on the transient process and to the comparison of the theoretical and the experimental results on CVTR. (author)

  20. Friction and Wear Management Using Solvent Partitioning of Hydrophilic-Surface-Interactive Chemicals Contained in Boundary Layer-Targeted Emulsions

    Science.gov (United States)

    Richmond, Robert Chaffee (Inventor); Schramm, Jr., Harry F. (Inventor); Defalco, Francis G. (Inventor)

    2015-01-01

    Lubrication additives of the current invention require formation of emulsions in base lubricants, created with an aqueous salt solution plus a single-phase compound such that partitioning within the resulting emulsion provides thermodynamically targeted compounds for boundary layer organization thus establishing anti-friction and/or anti-wear. The single-phase compound is termed "boundary layer organizer", abbreviated BLO. These emulsion-contained compounds energetically favor association with tribologic surfaces in accord with the Second Law of Thermodynamics, and will organize boundary layers on those surfaces in ways specific to the chemistry of the salt and BLO additives. In this way friction modifications may be provided by BLOs targeted to boundary layers via emulsions within lubricating fluids, wherein those lubricating fluids may be water-based or oil-based.

  1. Effect of mixing rule boundary conditions on high pressure (liquid + liquid) equilibrium prediction

    International Nuclear Information System (INIS)

    Hsieh, Min-Kang; Lin, Shiang-Tai

    2012-01-01

    Highlights: ► Prediction of LLE from the combined use of EOS and liquid model are examined. ► The mixing rule used affects the predicted pressure dependence of LLE. ► MHV1 mixing rule predicts decent LLE at low pressures. ► WS mixing rule predicts more accurate excess volume and LLE at high pressures. ► The hybrid of MHV1 and WS mixing rule gives overall the best predictions. - Abstract: We examine the prediction of high pressure (liquid + liquid) equilibrium (LLE) from the Peng–Robinson equation with three excess Gibbs free energy (G ex )-based mixing rules (MR): the first order modified Huron–Vidal (MHV1), the Wong–Sandler (WS), and a hybrid of these two (referred to as G ex B 2 ). These mixing rules differ by the boundary conditions used for determination of the temperature and composition dependence of parameters a and b in the PR EOS. The condition of matching the excess Gibbs free energy from the EOS at zero pressure to that from the G ex model, used in MHV1 and G ex B 2 MR, leads to a similar miscibility gap from PR EOS and the G ex model used. On the other hand, the condition of matching excess Helmholtz energy from the EOS at infinite pressure to that from the G ex model, used in the WS MR, shows remarkable deviations. The condition of quadratic composition dependence in the second virial coefficient (B 2 ), used in WS and G ex B 2 MR, allows for both positive and negative values in the molar excess volume. Depending on the mixture, either the increase or decrease of the miscibility gap with pressure can be observed when the WS or the G ex B 2 MR is used. The condition of linear combination of molecular sizes of each component used in the MHV1 MR, however, often leads to small, positive molar excess volumes. As a consequence, the predicted LLE from using the MHV1 MR are insensitive to pressure. Therefore, we find that the G ex B 2 mixing rule provides the best predictive power for the LLE over a wide range of temperature and pressure.

  2. Analysis code for pressure in reactor containment vessel of ATR. CONPOL

    International Nuclear Information System (INIS)

    1997-08-01

    For the evaluation of the pressure and temperature in containment vessels in the events which are classified in the abnormal change of pressure, atmosphere and others in reactor containment vessels in accident among the safety evaluation events of the ATR, the analysis code for the pressure in reactor containment vessels CONPOL is used. In this report, the functions of the analysis code and the analysis model are shown. By using this analysis code, the rise of the pressure and temperature in a containment vessel is evaluated when loss of coolant accident occurs, and high temperature, high pressure coolant flows into it. This code possesses the functions of computing blow-down quantity and heat dissipation from reactor cooling facility, steam condensing heat transfer to containment vessel walls, and the cooling effect by containment vessel spray system. As for the analysis techniques, the models of reactor cooling system, containment vessel and steam discharge pool, and the computation models for the pressure and temperature in containment vessels, wall surface temperature, condensing heat transfer, spray condensation and blow-down are explained. The experimental analysis of the evaluation of the pressure and temperature in containment vessels at the time of loss of coolant accident is reported. (K.I.)

  3. Analysis of containment pressure and temperature changes following loss of coolant accident (LOCA)

    International Nuclear Information System (INIS)

    Nguyen Van Thai; Kieu Ngoc Dung

    2015-01-01

    This paper present a preliminary thermal-hydraulics analysis of AP1000 containment following loss of coolant accident events such as double-end cold line break (DECLB) or main steam line break (MSLB) using MELCOR code. A break of this type will produce a rapid depressurization of the reactor pressure vessel (primary system) and release initially high pressure water into the containment followed by a much smaller release of highly superheated steam. The high pressure liquid water will flash and rapidly pressurize the containment building. The performance of passive containment cooling system for steam removal by condensation on large steel containment structure is a major contributing process, controlling the pressure and temperature maximum reached during the accident event. The results are analyzed, discussed and compared with the similar work done by Sandia National Laboratories. (author)

  4. Inverse eigenvalue problems for Sturm-Liouville equations with spectral parameter linearly contained in one of the boundary conditions

    OpenAIRE

    Guliyev, Namig J.

    2008-01-01

    International audience; Inverse problems of recovering the coefficients of Sturm–Liouville problems with the eigenvalue parameter linearly contained in one of the boundary conditions are studied: 1) from the sequences of eigenvalues and norming constants; 2) from two spectra. Necessary and sufficient conditions for the solvability of these inverse problems are obtained.

  5. A PC-based computer program for simulation of containment pressurization

    International Nuclear Information System (INIS)

    Seifaee, F.

    1990-01-01

    This paper reports that a PC-based computer program has been developed to simulate a pressurized water reactor (PWR) containment during various transients. This containment model is capable of determining pressure and temperature history of a PWR containment in the event of a loss of coolant accident, as well as main steam line breaks inside the containment. Conservation of mass and energy equations are applied to the containment model. Development of the program is based on minimization of input specified information and user friendliness. Maximization of calculation efficiency is obtained by superseding the traditional trial and error procedure for determination of the state variables and implementation of an explicit solution for pressure. The program includes simplified models for active heat removal systems. The results are in close agreement between the present model and CONTEMPT-MOD5 computer code for pressure and temperature inside the containment

  6. Containment pressure monitoring method after severe accident in nuclear power plant

    International Nuclear Information System (INIS)

    Luo Chuanjie; Zhang Shishui

    2011-01-01

    The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

  7. Ultimate capacity and influenced factors analysis of nuclear RC containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Song Chenning; Hou Gangling; Zhou Guoliang

    2014-01-01

    Ultimate compressive bearing capacity, influenced factors and its rules of nuclear RC containment are key problems of safety assessment, accident treatment and structure design, etc. Ultimate compressive bearing capacity of nuclear RC containment is shown by concrete damaged plasticity model and steel double liner model of ABAQUS. The study shows that the concrete of nuclear RC containment cylinder wall becomes plastic when the internal pressure is up to 0.87 MPa, the maximum tensile strain of steel liner exceeds 3000 × 10 6 and nuclear RC containment reaches ultimate status when the internal pressure is up to 1.02 MPa. The result shows that nuclear RC containment is in elastic condition under the design internal pressure and the bearing capacity meets requirement. Prestress and steel liner play key parts in the ultimate internal pressure and failure mode of nuclear RC containment. The study results have value for the analysis of ultimate compressive bearing capacity, structure design and safety assessment. (authors)

  8. Lattice Boltzmann simulations of pressure-driven flows in microchannels using Navier–Maxwell slip boundary conditions

    KAUST Repository

    Reis, Tim

    2012-01-01

    We present lattice Boltzmann simulations of rarefied flows driven by pressure drops along two-dimensional microchannels. Rarefied effects lead to non-zero cross-channel velocities, nonlinear variations in the pressure along the channel. Both effects are absent in flows driven by uniform body forces. We obtain second-order accuracy for the two components of velocity the pressure relative to asymptotic solutions of the compressible Navier-Stokes equations with slip boundary conditions. Since the common lattice Boltzmann formulations cannot capture Knudsen boundary layers, we replace the usual discrete analogs of the specular diffuse reflection conditions from continuous kinetic theory with a moment-based implementation of the first-order Navier-Maxwell slip boundary conditions that relate the tangential velocity to the strain rate at the boundary. We use these conditions to solve for the unknown distribution functions that propagate into the domain across the boundary. We achieve second-order accuracy by reformulating these conditions for the second set of distribution functions that arise in the derivation of the lattice Boltzmann method by an integration along characteristics. Our moment formalism is also valuable for analysing the existing boundary conditions. It reveals the origin of numerical slip in the bounce-back other common boundary conditions that impose conditions on the higher moments, not on the local tangential velocity itself. © 2012 American Institute of Physics.

  9. Two-dimensional properties of n-inversion layers in InSb grain boundaries under high hydrostatic pressure

    International Nuclear Information System (INIS)

    Kraak, W.; Herrmann, R.; Nachtwei, G.

    1985-01-01

    Magnetotransport properties of n-inversion layers in grain boundaries of p-InSb bicrystals are investigated under high hydrostatic pressure up to 10 3 MPa. A rapid decrease of the carrier concentration in the inversion layer is observed when hydrostatic pressure is applied. A simple model taking into account the pressure dependence of the energy band structure of pure InSb is proposed to describe this behaviour. (author)

  10. Characterization of Rare Reverse Flow Events in Adverse Pressure Gradient Turbulent Boundary Layers

    Science.gov (United States)

    Kaehler, Christian J.; Bross, Matthew; Fuchs, Thomas

    2017-11-01

    Time-resolved tomographic flow fields measured in the viscous sublayer region of a turbulent boundary layer subjected to an adverse pressure gradient (APG) are examined with the aim to resolve and characterize reverse flow events at Reτ = 5000. The fields were measured using a novel high resolution tomographic particle tracking technique. It is shown that this technique is able to fully resolve mean and time dependent features of the complex three-dimensional flow with high accuracy down to very near-wall distances ( 10 μm). From time resolved Lagrangian particle trajectories, statistical information as well as instantaneous topological features of near-wall flow events are deduced. Similar to the zero pressure gradient case (ZPG), it was found that individual events with reverse flow components still occur relatively rarely under the action of the pressure gradient investigated here. However, reverse flow events comprised of many individual events, are shown to appear in relatively organized groupings in both spanwise and streamise directions. Furthermore, instantaneous measurements of reverse flow events show that these events are associated with the motion of low-momentum streaks in the near-wall region. This work is supported by the Priority Programme SPP 1881 Turbulent Superstructures and the individual project Grant KA1808/8-2 of the Deutsche Forschungsgemeinschaft.

  11. The needs of the nuclear pressure boundary industry in the 1990s

    International Nuclear Information System (INIS)

    Amano, Makio

    1990-01-01

    In order to meet the increasing demand for electric power, it is recognized in Japan that light water reactors (BWR and PWR) will continue to play an important role in the 1990s. Some technical developments and research are considered necessary in the 1990s for the further establishment of the structural integrity of the light water reactors. Based on a review of a series of problems experienced at pressure boundaries, the desired improvements and the prospects for their achievement are discussed in the following 3 fields. (1) Improvements in order to attain availability: some new techniques and the importance of preventive maintenance, (2) Nuclear plant life extension: The integrity assessment method of aged plants and the development of diagnostic and monitoring techniques, and (3) Human factor considerations in the NSSS Vendor: Technology transfer to the next generation. (orig.)

  12. Investigation of the Condensation Effect at IRWST Pool Surface on Containment Back Pressure in APR1400 Containment

    International Nuclear Information System (INIS)

    Lee, Eui Jong; Lee, Jin Yong; Lee, Byung Chul

    2006-01-01

    The APR1400 has several new design concepts in order to improve the plant safety functions during a postulated accident. The In-Containment Refueling Water Storage Tank (IRWST) is one of the new design concepts of APR1400 and installed at the bottom of containment building to promote the plant safety functions by simplifying emergency core cooling water source and preventing release of the fission product during an accidents. This design feature, however, brings about uncertainty factors which may necessitate conventional prediction of temperature and pressure of containment building improved or revised under accident conditions. The hot steam which is released from RCS break enters into the IRWST through four Pressure Relief Dampers (PRDs). It is expected to be condensed with water stored in IRWST, colder than incoming steam. The purpose of this study is to examine closely the influence of the condensation effect at IRWST on containment back pressure in APR1400 containment building using the GOTHIC code which can predict the steam condensation on IRWST pool surface

  13. The online sealing performance test of the primary circuit pressure boundary check valve in nuclear power plants

    International Nuclear Information System (INIS)

    Yang Yunfei; Huang Huimin

    2013-01-01

    The primary circuit pressure boundary check valves of 320 MW pressurized water reactor is a nuclear grade I key equipment. The sealing demand is very high, which is directly related to the internal leakage rate of the primary circuit system. After the welding check valve is repaired, the sealing performance is judged by color printing checks. If there is water or humid vapor in the pipe, it will affect the accuracy of the color printing checks. For the particularity of the online check valve tightness test, online detecting device is designed by the hydraulic pressure drop method in other nuclear power plants, but the method has some shortcomings and restrictions. In this paper, we design a reliable and portable test equipment by the low-pressure gas seal test flow measurement, which make accurate and quantitative judgment of sealing property after the pressure boundary check valves are repaired. (authors)

  14. Problems identified in quantifying leak before break in pressure containing structures

    International Nuclear Information System (INIS)

    Darlaston, B.J.L.; Connors, D.C.; Hellen, R.A.J.

    1979-01-01

    The leak before break approach is often applied to pressure containing plant as part of the safety assessment. The assumptions used in this approach are sometimes very pessimistic. It is therefore desirable to be able to quantify the concept more precisely. The two aspects which are of considerable importance are the way the crack profile develops and what happens when the remaining ligament below the crack fails. These two aspects are receiving attention and together with the development of the basic concept of 'leak before break' form the basis of this paper. Some thirty burst tests have been carried out on straight pipes of various dimensions. The results have been analysed using the CEGB Failure Assessment Route for structures containing defects. It was shown that in most cases the leaks and the breaks could be separated by this procedure. However all these tests involved machined rather than fatigue grown defects. A complementary program on pipes has the objective of examining defect growth under cyclic loads. The tests on the 152 mm diameter pipes showed that these defects did not grow in a uniform manner but after a while began to tunnel through the wall locally leading to failure of part of the ligament. This implies that some defects considered to be in the break category would only lead to leaks. As a consequence of these results the experimental programme was redesigned to concentrate on the growth of defects which it was thought would span the boundary of leak and break. For the pipe dimensions and materials used, this represented long defects which would penetrate well into the wall before ligament failure occurred. The analysis and interpretation of this aspect of the programme is part analytical part empirical. (orig.)

  15. Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

    International Nuclear Information System (INIS)

    Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

    2015-01-01

    The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

  16. High-Reynolds-number turbulent-boundary-layer wall-pressure fluctuations with dilute polymer solutions

    Science.gov (United States)

    Elbing, Brian R.; Winkel, Eric S.; Ceccio, Steven L.; Perlin, Marc; Dowling, David R.

    2010-08-01

    Wall-pressure fluctuations were investigated within a high-Reynolds-number turbulent boundary layer (TBL) modified by the addition of dilute friction-drag-reducing polymer solutions. The experiment was conducted at the U.S. Navy's Large Cavitation Channel on a 12.9 m long flat-plate test model with the surface hydraulically smooth (k+<0.2) and achieving downstream-distance-based Reynolds numbers to 220×106. The polymer (polyethylene oxide) solution was injected into the TBL through a slot in the surface. The primary flow diagnostics were skin-friction drag balances and an array of flush-mounted dynamic pressure transducers 9.8 m from the model leading edge. Parameters varied included the free-stream speed (6.7, 13.4, and 20.2 m s-1) and the injection condition (polymer molecular weight, injection concentration, and volumetric injection flux). The behavior of the pressure spectra, convection velocity, and coherence, regardless of the injection condition, were determined primarily based on the level of drag reduction. Results were divided into two regimes dependent on the level of polymer drag reduction (PDR), nominally separated at a PDR of 40%. The low-PDR regime is characterized by decreasing mean-square pressure fluctuations and increasing convection velocity with increasing drag reduction. This shows that the decrease in the pressure spectra with increasing drag reduction is due in part to the moving of the turbulent structures from the wall. Conversely, with further increases in drag reduction, the high-PDR regime has negligible variation in the mean-squared pressure fluctuations and convection velocity. The convection velocity remains constant at approximately 10% above the baseline-flow convection velocity, which suggests that the turbulent structures no longer move farther from the wall with increasing drag reduction. In light of recent numerical work, the coherence results indicate that in the low-PDR regime, the turbulent structures are being elongated in

  17. Interaction between a normal shock wave and a turbulent boundary layer at high transonic speeds. I - Pressure distribution

    Science.gov (United States)

    Messiter, A. F.

    1980-01-01

    Asymptotic solutions are derived for the pressure distribution in the interaction of a weak normal shock wave with a turbulent boundary layer. The undisturbed boundary layer is characterized by the law of the wall and the law of the wake for compressible flow. In the limiting case considered, for 'high' transonic speeds, the sonic line is very close to the wall. Comparisons with experiment are shown, with corrections included for the effect of longitudinal wall curvature and for the boundary-layer displacement effect in a circular pipe.

  18. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    International Nuclear Information System (INIS)

    Schulz, K.C.

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K Q due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail

  19. Kinetic boundaries and phase transformations of ice i at high pressure

    Science.gov (United States)

    Wang, Yu; Zhang, Huichao; Yang, Xue; Jiang, Shuqing; Goncharov, Alexander F.

    2018-01-01

    Raman spectroscopy in diamond anvil cells has been employed to study phase boundaries and transformation kinetics of H2O ice at high pressures up to 16 GPa and temperatures down to 15 K. Ice i formed at nearly isobaric cooling of liquid water transforms on compression to high-density amorphous (HDA) ice at 1.1-3 GPa at 15-100 K and then crystallizes in ice vii with the frozen-in disorder (ice vii') which remains stable up to 14.1 GPa at 80 K and 15.9 GPa at 100 K. Unexpectedly, on decompression of ice vii', it transforms to ice viii in its domain of metastability, and then it relaxes into low-density amorphous (LDA) ice on a subsequent pressure release and warming up. On compression of ice i at 150-170 K, ice ix is crystallized and no HDA ice is found; further compression of ice ix results in the sequential phase transitions to stable ices vi and viii. Cooling ice i to 210 K at 0.3 GPa transforms it to a stable ice ii. Our extensive investigations provide previously missing information on the phase diagram of water, especially on the kinetic paths that result in formation of phases which otherwise are not accessible; these results are keys for understanding the phase relations including the formation of metastable phases. Our observations inform on the ice modifications that can occur naturally in planetary environments and are not accessible for direct observations.

  20. Effects of irradiation and thermal aging upon fatigue-crack growth behavior of reactor pressure boundary materials. [Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    James, L. A.

    1978-10-01

    Two processes that have the potential to produce degradation in the properties of pressure boundary materials are neutron irradiation and long-time thermal aging. This paper uses linear-elastic fracture mechanics techniques to assess the effect of these two processes upon the fatigue-crack growth behavior of a number of alloys commonly employed in reactor pressure boundaries. The materials evaluated include ferritic steels, austenitic stainless steels, and nickel-base alloys typical of those employed in a number of reactor types including water-cooled, gas-cooled, and liquid-metal-cooled designs.

  1. Study of Boundary Layer Convective Heat Transfer with Low Pressure Gradient Over a Flat Plate Via He's Homotopy Perturbation Method

    International Nuclear Information System (INIS)

    Fathizadeh, M.; Aroujalian, A.

    2012-01-01

    The boundary layer convective heat transfer equations with low pressure gradient over a flat plate are solved using Homotopy Perturbation Method, which is one of the semi-exact methods. The nonlinear equations of momentum and energy solved simultaneously via Homotopy Perturbation Method are in good agreement with results obtained from numerical methods. Using this method, a general equation in terms of Pr number and pressure gradient (λ) is derived which can be used to investigate velocity and temperature profiles in the boundary layer.

  2. Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima, E-mail: dayanefs@ipen.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2017-07-01

    The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

  3. Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

    International Nuclear Information System (INIS)

    Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima

    2017-01-01

    The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

  4. Anomalous composition dependence of the band gap pressure coefficients in In-containing nitride semiconductors

    DEFF Research Database (Denmark)

    Gorczyca, I.; Kamińska, A.; Staszczak, G.

    2010-01-01

    The pressure-induced changes in the electronic band structures of In-containing nitride alloys, InxGa1-xN and InxAl1-xN are examined experimentally as well as by ab initio calculations. It is found that the band gap pressure coefficients, dEg/dp, exhibit very large bowing with x, and calculations...

  5. Analysis on the effect of risk from containment failure by over-pressurization during the operation of containment filtered venting system

    International Nuclear Information System (INIS)

    Ham, Jaehyun; Kang, Hyun Gook; Chang, Soon Heung

    2015-01-01

    Passive safety systems which are operated without power source are suggested as a solution SBO. For containment protection system, Containment Filtered Venting System (CFVS) is suggested. CFVS controls the containment pressure by releasing the containment gas through filter passively without any power source. But because still small amount of radioactive material have no choice but to release to the environment, starting time and operation method of CFVS have to be determined carefully. Later starting time brings not only lower release but also higher risk from containment failure by over-pressurization, so it is a problem. In this research, the effect of risk from containment failure by over-pressurization during the operation of containment filtered venting system was analyzed. In this research, optimized values for variables of the CFVS operation method are found as 0.67 MPa, 9 cm, 0.1 MPa each for open pressure, pressure interval, and vent pipe diameter when DF as a function of time and risk from containment over-pressurization failure are considered. Generally in this research, release without risk get lower values in higher pressure, and lower vent pipe diameter. Release with risk get sharply high values when the containment pressure exceeds the design pressure because of the effect of risk from containment failure by over-pressurization. In conclusion, highest pressure, and lowest vent pipe diameter which are not influenced by risk is the optimized values for CFVS operation method because amount of risk is much larger than release through the CFVS

  6. Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

    International Nuclear Information System (INIS)

    Park, S.Y.; Ahn, K.I.

    2014-01-01

    Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

  7. Experimental investigation on the behaviour of pressure suppression containment systems by the SOPRE-1 facility

    International Nuclear Information System (INIS)

    Cerullo, N.; Delli Gatti, A.; Marinelli, M.; Mazzini, M.; Mazzoni, A.; Sbrana, A.; Todisco, P.

    1977-01-01

    The SOPRE-1 test facility is an integral model (scale 1:13) of a MARK II pressure suppression containment system. It was set up at the University of Pisa in order to study the pressure-temperature transient in pressure suppression containment systems during LOCAs. Knowledge of this transient is necessary to perform a correct structural analysis of reactor containment. The containment system behaviour is studied by changing the principal parameters which affect the transient (blow-down mass and energy release, suppression pool water temperature, vent pipe number and submergence heat transfer coefficients). The first series of tests involved: A) 13 tests with break area of 1.8 cm 2 , B) 8 tests with break area of 20.0 cm 2 . The following experimental conditions were changed: - position of the simulated break (from liquid or steam zone), - water pressure (20-85 Kgsub(p)/cm 2 ) and mass (45-70Kg) in the vessel model. Tests A): the CONTEMPT codes correctly forecast the pressure-temperature history, both in dry- and in wet-well. Tests B): the experimental runs have shown that increasing of blow-down flowrate produces dry-well pressure spatial differences and anomalous vent pipe behaviour. This results in damped oscillations of dry- and wet-well pressure, probably due to alterbating air bubble over-expansion and collapse, and in vent pipe opening and reclosing. (Auth.)

  8. Evaluation of CANDU NPP containment structure subjected to aging and internal pressure increase

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xu [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Kwon, Oh-Sung, E-mail: os.kwon@utoronto.ca [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Bentz, Evan [Department of Civil Engineering, University of Toronto, Toronto M5S 1A4 (Canada); Tcherner, Julia [Candu Energy Inc. a member of SNC-Lavalin Group, Mississauga L5K 1B1 (Canada)

    2017-04-01

    Highlights: • The aging effects on the performance of a nuclear containment structure is evaluated. • A numerical model of the structure is subjected to increasing internal pressure. • No through-thickness cracks are predicted under the design level internal pressure. • The structure is predicted to be ductile up to large internal pressure levels. - Abstract: The objective of this study is to investigate the long-term performance of a typical CANDU® containment structure. A three-dimensional nonlinear finite element model was built to realistically evaluate the performance of the structure under service load as well as a hypothetical beyond-design level internal pressure. Consideration is given to the time-dependent effects, such as shrinkage, creep, and relaxation of prestressing tendons, over a 60-year timeframe. In addition, the sensitivity of the response of the containment structure against support condition, internal temperature profile and temporary construction openings was also investigated. The accuracy of the numerical model was validated against structural measurements made during a routine leak rate test. The analysis results show that the containment structure would develop a ductile mechanism if the internal pressure significantly exceeded the design pressure. The pressure-deformation relationship of the structure is sensitive to the considered time-dependent parameters.

  9. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Khattab, M; Ibrahim, N A; Bedrose, C D [Reactors department, nuclear research center, atomic energy authority, Cairo, (Egypt)

    1995-10-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs.

  10. Methodology for predicting ultimate pressure capacity of the ACR-1000 containment structure

    International Nuclear Information System (INIS)

    Saudy, A.M.; Awad, A.; Elgohary, M.

    2006-01-01

    The Advanced CANDU Reactor or the ACR-1000 is developed by Atomic Energy of Canada Limited (AECL) to be the next step in the evolution of the CANDU product line. It is based on the proven CANDU technology and incorporates advanced design technologies. The ACR containment structure is an essential element of the overall defense in depth approach to reactor safety, and is a physical barrier against the release of radioactive material to the environment. Therefore, it is important to provide a robust design with an adequate margin of safety. One of the key design requirements of the ACR containment structure is to have an ultimate pressure capacity that is at least twice the design pressure Using standard design codes, the containment structure is expected to behave elastically at least up to 1.5 times the design pressure. Beyond this pressure level, the concrete containment structure with reinforcements and post-tension tendons behaves in a highly non-linear manner and exhibits a complex response when cracks initiate and propagate. To predict the structural non-linear responses, at least two critical features are involved. These are: the structural idealization by the geometry and material property models, and the adopted solution algorithm. Therefore, detailed idealization of the concrete structure is needed in order to accurately predict its ultimate pressure capacity. This paper summarizes the analysis methodology to be carried out to establish the ultimate pressure capacity of the ACR containment structure and to confirm that the structure meets the specified design requirements. (author)

  11. Pressure behavior in nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    Khattab, M.; Ibrahim, N.A.; Bedrose, C.D.

    1995-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The seniors of small, medium and large LOCA at 2%, 15%, and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs

  12. Pressure behaviour in a nuclear reactor containment following a loss of coolant accident

    International Nuclear Information System (INIS)

    KHattab, M.S.; Ibrahim, N.A.; Bedrose, S.D.

    1994-01-01

    The scenarios of pressure variation following a loss of coolant accident (LOCA) inside the containment of pressurized water reactor (PWR) have been investigated. Critical mass flow rushing out from high pressure leg through pipe break, is used to calculate the rate of coolant. The energy added to the containment atmosphere is determined to specify the rate of growth of pressure and temperature. The scenarios of small, medium and large LOCA at 2%, 15% and 25% flow released are investigated. Safety water spray system is initiated as the pressure reaches the containment design safety limit at about 3 bar to depressurise and to cooldown the system and thereby to reduce the concentration of radioactivity release in the containment atmosphere. The pressure response before and after operation of safety spray system is predicted in each size of LOCA using a typical design of westinghouse PWR system. The results of large LOCA showed good agreement with westinghouse calculations of the same design. The heat removal from the containment environment is rejected into the sump by drop-wise condensation mechanism. The effect of initial droplets diameters injected from the nozzles of the spray system is investigated. The results show that the droplet diameter of 3 mm gives best performance. 6 figs., 1 tab

  13. Fracture Analysis of CNG High Pressure Container using Fractography and Measurement of Property

    Directory of Open Access Journals (Sweden)

    Kim Eui-Soo

    2017-01-01

    Full Text Available Bursting accidents of pressure containers due to design and manufacturing defects are frequently occurring. Due to high-pressure gas or harmful substances, when this vessel is fractured, it can lead to catastrophic disasters. Especially, in the event of bursting accident of composite pressure vessel for CNG bus, many unspecified people can be damaged. Most of the accidents were caused by problems in the manufacturing process. The manufacturing process for TYPE2 pressure vessel is very complicated such as three drawing processes, two ironing processes and one spinning process. In the middle of process, various heat treatments are performed for imparting toughness and removing residual stresses. It should cause a serious problem such as bursting and fragmentation of the pressure container due to defects of this process. In this research, the fracture cause of CNG vessel is evaluated through fractography and measuring material property using IIT and analysis of chemical composition.

  14. Experimental investigation on the behavior of pressure suppression containment systems by the SOPRE-1 facility

    International Nuclear Information System (INIS)

    Cerullo, N.; Delli Gatti, A.; Marinelli, M.; Mazzini, M.; Mazzoni, A.; Sbrana, A.; Todisco, P.

    1977-01-01

    The SOPRE-1 test facility is an integral model (scale 1:13) of a MARK II pressure suppression containment system. It was set up at the University of Pisa in order to study the pressure-temperature transient in pressure suppression containment systems during LOCAs. Knowledge of this transient is necessary to perform a correct structural analysis of reactor containment. The containment system behavior is studied by changing the principal parameters which affect the transient (blow-down mass and energy release, suppression pool water temperature, vent pipe number and submergence, heat transfer coefficients). The first series of tests involved: A) 13 tests with break area of 1.8 cm 2 , B) 8 tests with break area of 20.0 cm 2 . The following experimental conditions were changed: position of the simulated break (from liquid or steam zone), water pressure (20-85 Kg/cm 2 ) and mass (45-70 Kg) in the vessel model. Tests A): the CONTEMPT codes correctly forecast the pressure-temperature history, both in dry- and in wet-well. Tests B): the experimental runs have shown that increasing of blow-down flowrate produces dry-well pressure spatial differences and anomalous vent pipe behavior. This results in damped oscillations of dry- and wet-well pressure, probably due to alternating air bubble over-expansion and collapse, and in vent pipe opening and reclosing. Dry-well pressure maxima at the end of blow-down are greater than those forecasted by currently applied codes: these codes use an homogeneous model, and do not take into account the above mentioned dynamic phenomena. In some tests other interesting phenomena were observed, such as some local pressure peaks in the suppression pool greater than dry-well pessure maxima at the end of blow-down. At present, all these phenomena are under study; they could be important for the structural analysis of containment systems

  15. Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection

    Energy Technology Data Exchange (ETDEWEB)

    Vo, T.V.; Smith, B.W.; Simonen, F.A.; Gore, B.F.

    1994-08-01

    The goals of the Evaluation and Improvement of Non-destructive Examination Reliability for the In-service Inspection of Light Water Reactors Program sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory (PNL) are to (1) assess current ISI techniques and requirements for all pressure boundary systems and components, (2) determine if improvements to the requirements are needed, and (3) if necessary, develop recommendations for revising the applicable ASME Codes and regulatory requirements. In evaluating approaches that could be used to provide a technical basis for improved inservice inspection plans, PNL has developed and applied a method that uses results of probabilistic risk assessment (PRA) to establish piping system ISI requirements. In the PNL program, the feasibility of generic ISI requirements is being addressed in two phases. Phase I involves identifying and prioritizing the systems most relevant to plant safety. The results of these evaluations will be later consolidated into requirements for comprehensive inservice inspection of nuclear power plant components that will be developed in Phase II. This report presents Phase I evaluations for eight selected plants and attempts to compare these PRA-based inspection priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there are generic insights that can be extrapolated from the selected plants to specific classes of light water reactors.

  16. Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection

    International Nuclear Information System (INIS)

    Vo, T.V.; Smith, B.W.; Simonen, F.A.; Gore, B.F.

    1994-08-01

    The goals of the Evaluation and Improvement of Non-destructive Examination Reliability for the In-service Inspection of Light Water Reactors Program sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory (PNL) are to (1) assess current ISI techniques and requirements for all pressure boundary systems and components, (2) determine if improvements to the requirements are needed, and (3) if necessary, develop recommendations for revising the applicable ASME Codes and regulatory requirements. In evaluating approaches that could be used to provide a technical basis for improved inservice inspection plans, PNL has developed and applied a method that uses results of probabilistic risk assessment (PRA) to establish piping system ISI requirements. In the PNL program, the feasibility of generic ISI requirements is being addressed in two phases. Phase I involves identifying and prioritizing the systems most relevant to plant safety. The results of these evaluations will be later consolidated into requirements for comprehensive inservice inspection of nuclear power plant components that will be developed in Phase II. This report presents Phase I evaluations for eight selected plants and attempts to compare these PRA-based inspection priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there are generic insights that can be extrapolated from the selected plants to specific classes of light water reactors

  17. Constructing integrable full-pressure full-current free-boundary stellarator magnetohydrodynamic equilibrium solutions

    International Nuclear Information System (INIS)

    Hudson, S.R.

    2002-01-01

    For stellarators to be feasible candidates for fusion power stations it is essential that the magnetic field lines lie on nested flux surfaces; however, the lack of a continuous symmetry implies that magnetic islands, caused by Pfirsch-Schlueter currents, diamagnetic currents and resonant coil fields, are guaranteed to exist. The challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary full-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the PIES code [Comp. Phys. Comm., 43:157, 1986] which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to lie in the nullspace of certain measures of engineering acceptability and kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied to a candidate plasma and coil design for NCSX [Phys. Plas., 7:1911, 2000]. (author)

  18. Three-dimensional local ALE-FEM method for fluid flow in domains containing moving boundaries/objects interfaces

    Energy Technology Data Exchange (ETDEWEB)

    Carrington, David Bradley [Los Alamos National Laboratory (LANL), Los Alamos, NM (United States); Monayem, A. K. M. [Univ. of New Mexico, Albuquerque, NM (United States); Mazumder, H. [Univ. of New Mexico, Albuquerque, NM (United States); Heinrich, Juan C. [Univ. of New Mexico, Albuquerque, NM (United States)

    2015-03-05

    A three-dimensional finite element method for the numerical simulations of fluid flow in domains containing moving rigid objects or boundaries is developed. The method falls into the general category of Arbitrary Lagrangian Eulerian methods; it is based on a fixed mesh that is locally adapted in the immediate vicinity of the moving interfaces and reverts to its original shape once the moving interfaces go past the elements. The moving interfaces are defined by separate sets of marker points so that the global mesh is independent of interface movement and the possibility of mesh entanglement is eliminated. The results is a fully robust formulation capable of calculating on domains of complex geometry with moving boundaries or devises that can also have a complex geometry without danger of the mesh becoming unsuitable due to its continuous deformation thus eliminating the need for repeated re-meshing and interpolation. Moreover, the boundary conditions on the interfaces are imposed exactly. This work is intended to support the internal combustion engines simulator KIVA developed at Los Alamos National Laboratories. The model's capabilities are illustrated through application to incompressible flows in different geometrical settings that show the robustness and flexibility of the technique to perform simulations involving moving boundaries in a three-dimensional domain.

  19. Parametric studies on containment thermal hydraulic loads during high pressure melt ejection in a BWR

    Energy Technology Data Exchange (ETDEWEB)

    Silde, A.; Lindholm, I. [VTT Energy, Espoo (Finland)

    1997-12-01

    The containment thermal hydraulic loads during high pressure melt ejection in a Nordic BWR are studied parametrically with the CONTAIN and the MELCOR codes. The work is part of the Nordic RAK-2 project. The containment analyses were divided into two categories according to composition of the discharged debris: metallic and oxidic debris cases. In the base case with highly metallic debris, all sources from the reactor coolant system to the containment were based on the MELCOR/BH calculation. In the base case with the oxidic debris, the source data was specified assuming that {approx} 15% of the whole core material inventory and 34,000 kg of saturated water was discharged from the reactor pressure vessel (RPV) during 30 seconds. In this case, the debris consisted mostly of oxides. The highest predicted containment pressure peaks were about 8.5 bar. In the scenarios with highly metallic debris source, very high gas temperature of about 1900 K was predicted in the pedestal, and about 1400 K in the upper drywell. The calculations with metallic debris were sensititive to model parameters, like the particle size and the parameters, which control the chemical reaction kinetics. In the scenarios with oxidic debris source, the predicted pressure peaks were comparable to the cases with the metallic debris source. The maximum gas temperatures (about 450-500 K) in the containment were, however, significantly lower than in the respective metallic debris case. The temperatures were also insensitive to parametric variations. In addition, one analysis was performed with the MELCOR code for benchmarking of the MELCOR capabilities against the more detailed CONTAIN code. The calculations showed that leak tightness of the containment penetrations could be jeopardized due to high temperature loads, if a high pressure melt ejection occurred during a severe accident. Another consequence would be an early containment venting. (au). 28 refs.

  20. On the Unsteadiness of a Transitional Shock Wave-Boundary Layer Interaction Using Fast-Response Pressure-Sensitive Paint

    Science.gov (United States)

    Lash, E. Lara; Schmisseur, John

    2017-11-01

    Pressure-sensitive paint has been used to evaluate the unsteady dynamics of transitional and turbulent shock wave-boundary layer interactions generated by a vertical cylinder on a flat plate in a Mach 2 freestream. The resulting shock structure consists of an inviscid bow shock that bifurcates into a separation shock and trailing shock. The primary features of interest are the separation shock and an upstream influence shock that is intermittently present in transitional boundary layer interactions, but not observed in turbulent interactions. The power spectral densities, frequency peaks, and normalized wall pressures are analyzed as the incoming boundary layer state changes from transitional to fully turbulent, comparing both centerline and outboard regions of the interaction. The present study compares the scales and frequencies of the dynamics of the separation shock structure in different boundary layer regimes. Synchronized high-speed Schlieren imaging provides quantitative statistical analyses as well as qualitative comparisons to the fast-response pressure sensitive paint measurements. Materials based on research supported by the U.S. Office of Naval Research under Award Number N00014-15-1-2269.

  1. KAPP-3 and 4 containment pressure following postulated severe accident along with SAMG implementation

    International Nuclear Information System (INIS)

    Sharma, Sanjeev Kr.; Bhartia, D.K.; Mohan, Nalini; Malhotra, P.K.; Ghadge, S.G.; Chandra, Umesh

    2011-01-01

    Containment is an ultimate safety barrier which is designed to enclose whole reactor systems and to prevent the spread of active air-borne fission products. Studies are done to access its performance following severe accident i.e. Loss of Coolant Accident (LOCA) along with failure of Emergency Core Cooling System (ECCS), moderator and calandria vault water cooling system. The accident progression begins with the double ended break in reactor outlet/inlet header with simultaneous failure of ECCS followed by failure of moderator and calandria vault water cooling system. Initially decay heat and metal water reaction energy are assumed to be added to moderator water resulting in boiling of moderator and re-pressurization of containment due to steam addition. Subsequent to moderator boiling, decay heat and metal water reaction energy are assumed to be added to calandria vault water resulting in boiling and re-pressurization of containment due to steam addition. After moderator and calandria vault water have completely boiled off, rapid hydrogen generation would take place due to oxidation of pressure tubes and calandria tubes. In such accident scenario, the core is severely damaged. It will also lead to release of a large quantity of radio nuclides to containment atmosphere. To arrest the progression of accident, which can result in Severe Core damage and large amount of hydrogen production, which could leads to containment failure due to hydrogen deflagration or detonation, application of Severe Accident Management Guidelines (SAMG) has been studied. SAMG involve addition of water to calandria and calandria vault. It would result the boiling of the added water and consequent pressurization of containment. This paper presents the analysis for pressure-temperature of KAPP-3 and 4 containment following the postulated accident along with the application of Severe Accident Management Guidelines (SAMG). SAMG initiated action helps in arresting the progression of core

  2. MDEP Technical Report TR-CSWG-02. Technical Report on Lessons Learnt on Achieving Harmonisation of Codes and Standards for Pressure Boundary Components in Nuclear Power Plants

    International Nuclear Information System (INIS)

    2013-01-01

    This report was prepared by the Multinational Design Evaluation Program's (MDEP's) Codes and Standards Working Group (CSWG). The primary, long-term goal of MDEP's CSWG is to achieve international harmonisation of codes and standards for pressure-boundary components in nuclear power plants. The CSWG recognised early on that the first step to achieving harmonisation is to understand the extent of similarities and differences amongst the pressure-boundary codes and standards used in various countries. To assist the CSWG in its long-term goals, several standards developing organisations (SDOs) from various countries performed a comparison of their pressure-boundary codes and standards to identify the extent of similarities and differences in code requirements and the reasons for their differences. The results of the code-comparison project provided the CSWG with valuable insights in developing the subsequent actions to take with SDOs and the nuclear industry to pursue harmonisation of codes and standards. The results enabled the CSWG to understand from a global perspective how each country's pressure-boundary code or standard evolved into its current form and content. The CSWG recognised the important fact that each country's pressure-boundary code or standard is a comprehensive, living document that is continually being updated and improved to reflect changing technology and common industry practices unique to each country. The rules in the pressure-boundary codes and standards include comprehensive requirements for the design and construction of nuclear power plant components including design, materials selection, fabrication, examination, testing and overpressure protection. The rules also contain programmatic and administrative requirements such as quality assurance; conformity assessment (e.g., third-party inspection); qualification of welders, welding equipment and welding procedures; non-destructive examination (NDE) practices and

  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. An Evaluation of a Phase-Lag Boundary Condition for Francis Hydroturbine Simulations Using a Pressure-Based Solver

    Science.gov (United States)

    Wouden, Alex; Cimbala, John; Lewis, Bryan

    2014-11-01

    While the periodic boundary condition is useful for handling rotational symmetry in many axisymmetric geometries, its application fails for analysis of rotor-stator interaction (RSI) in multi-stage turbomachinery flow. The inadequacy arises from the underlying geometry where the blade counts per row differ, since the blade counts are crafted to deter the destructive harmonic forces of synchronous blade passing. Therefore, to achieve the computational advantage of modeling a single blade passage per row while preserving the integrity of the RSI, a phase-lag boundary condition is adapted to OpenFOAM® software's incompressible pressure-based solver. The phase-lag construct is accomplished through restating the implicit periodic boundary condition as a constant boundary condition that is updated at each time step with phase-shifted data from the coupled cells adjacent to the boundary. Its effectiveness is demonstrated using a typical Francis hydroturbine modeled as single- and double-passages with phase-lag boundary conditions. The evaluation of the phase-lag condition is based on the correspondence of the overall computational performance and the calculated flow parameters of the phase-lag simulations with those of a baseline full-wheel simulation. Funded in part by DOE Award Number: DE-EE0002667.

  5. Ultimate analysis of PWR prestressed concrete containment subjected to internal pressure

    International Nuclear Information System (INIS)

    Hu, H.-T.; Lin, Y.-H.

    2006-01-01

    Numerical analyses are carried out by using the ABAQUS finite element program to predict the ultimate pressure capacity and the failure mode of the PWR prestressed concrete containment at Maanshan nuclear power plant. Material nonlinearity such as concrete cracking, tension stiffening, shear retention, concrete plasticity, yielding of prestressing tendon, yielding of steel reinforcing bar and degradation of material properties due to high temperature are all simulated with proper constitutive models. Geometric nonlinearity due to finite deformation has also been considered. The results of the analysis show that when the prestressed concrete containment fails, extensive cracks take place at the apex of the dome, the junction of the dome and cylinder, and the bottom of the cylinder connecting to the base slab. In addition, the ultimate pressure capacity of the containment is higher than the design pressure by 86%

  6. Acid pressure leaching of a concentrate containing uranium, thorium and rare earth elements

    International Nuclear Information System (INIS)

    Lan Xinghua; Peng Ruqing.

    1987-01-01

    The acid pressure leaching of a concentrate containing rinkolite for recovering uranium, thorium and rare earth elements is described. The laboratory and the pilot plant test results are given. Under the optimum leaching conditions, the recovery of uranium, thorium and rare earth elements are 82.9%, 86.0% and 88.3% respectively. These results show that the acid pressure leaching process is a effective process for treating the concentrate

  7. Northwest Boundary Containment/Treatment System Baseline Conditions, System Startup, and Operational Assessment. Volume 1.

    Science.gov (United States)

    1987-12-01

    combination pressure reducing check valve that is manually I controlled. A shutoff valve is installed on each well discharge line to iso - late the well from...system. This profile exhibits a relatively gentle, stable gradient between wells 27003 and 22051 which is due to the alignment of this portion of the...isoconcentration lines. As a result, the iso - 3 concentration maps for this period were constructed using isoconcentration maps for the later years as a

  8. Constructing Integrable High-pressure Full-current Free-boundary Stellarator Magnetohydrodynamic Equilibrium Solutions

    International Nuclear Information System (INIS)

    Hudson, S.R.; Monticello, D.A.; Reiman, A.H.; Strickler, D.J.; Hirshman, S.P.; Ku, L-P; Lazarus, E.; Brooks, A.; Zarnstorff, M.C.; Boozer, A.H.; Fu, G-Y.; Neilson, G.H.

    2003-01-01

    For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch-Schluter currents, diamagnetic currents, and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver [Reiman and Greenside, Comp. Phys. Comm. 43 (1986) 157] which iterate s the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator Experiment [Reiman, et al., Phys. Plasmas 8 (May 2001) 2083

  9. Constructing integrable high-pressure full-current free-boundary stellarator magnetohydrodynamic equilibrium solutions

    International Nuclear Information System (INIS)

    Hudson, S.R.; Monticello, D.A.; Reiman, A.H.

    2003-01-01

    For the (non-axisymmetric) stellarator class of plasma confinement devices to be feasible candidates for fusion power stations it is essential that, to a good approximation, the magnetic field lines lie on nested flux surfaces; however, the inherent lack of a continuous symmetry implies that magnetic islands responsible for breaking the smooth topology of the flux surfaces are guaranteed to exist. Thus, the suppression of magnetic islands is a critical issue for stellarator design, particularly for small aspect ratio devices. Pfirsch-Schlueter currents, diamagnetic currents and resonant coil fields contribute to the formation of magnetic islands, and the challenge is to design the plasma and coils such that these effects cancel. Magnetic islands in free-boundary high-pressure full-current stellarator magnetohydrodynamic equilibria are suppressed using a procedure based on the Princeton Iterative Equilibrium Solver (Reiman and Greenside 1986 Comput. Phys. Commun. 43 157) which iterates the equilibrium equations to obtain the plasma equilibrium. At each iteration, changes to a Fourier representation of the coil geometry are made to cancel resonant fields produced by the plasma. The changes are constrained to preserve certain measures of engineering acceptability and to preserve the stability of ideal kink modes. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible, the plasma is stable to ideal kink modes, and the coils satisfy engineering constraints. The method is applied to a candidate plasma and coil design for the National Compact Stellarator eXperiment (Reiman et al 2001 Phys. Plasma 8 2083). (author)

  10. Preliminary thermal design of a pressurized water reactor containment for handling severe accident consequences

    International Nuclear Information System (INIS)

    Abdullah, A.M.; Karameldin, A.

    1998-01-01

    A one-dimensional mathematical model has been developed for a 4250 MW(th) Advanced Pressurized Water Reactor containment analysis following a severe accident. The cooling process of the composite containment-steel shell and concrete shield- is achievable by natural circulation of atmospheric air. However, for purpose of gettering higher degrees of safety margin, the present study undertakes two objectives: (1) Installment of a diesel engine-driven air blower to force air through the annular space between the steel shell and concrete shield. The engine can be remotely operated to be effective in case of station blackout. (ii) Fixing longitudinally plate fins on the circumference of the inside and outside containment steel shell. These fins increase the heat transfer areas and hence the rate of heat removal from the containment atmosphere. In view of its importance - from the safety viewpoint - the long term behaviour of the containment which is a quasi-steady state problem, is formulated through a system of coupled nonlinear algebraic equations which describe the thermal-hydraulic and thermodynamic behaviour of the double shell containment. The calculated results revealed the following: (i) the passively air cooled containment can remove maximum heat load of 11.5 MW without failure, (ii) the effect of finned surface in the air passage tends to decrease the containment pressure by 20 to 30%, depending on the heat load, (iii) the effect of condensing fins is negligible for the proposed fin dimensions and material. However, by reducing the fin width, increasing their thickness, doubling their number, and using a higher conductive metal than the steel, it is expected that the containment pressure can be further reduced by 10% or more, (iv) the fins' dimensions and their number must be optimized via maximizing the difference or the ratio between the heat removed and pressure drop to get maximum heat flow rate

  11. Sensitivity of break-flow-partition on the containment pressure and temperature

    International Nuclear Information System (INIS)

    Kwon, Young Min; Song, Jin Ho; Lee, Sang Yong

    1994-01-01

    For the case of RCS blowdown into the vapor region of a containment at low pressure, the blowdown mixture will start to boil at the containment pressure and liquid will separate from the flow near the break location. The flashed steam is added to the containment atmosphere and liquid is falled to the sump. Analytically, the break flow can be divided into steam and liquid in a number of ways. Discussed in this study is three partition models and Instantaneous Mixing(IM) Model employed in different containment analysis computer codes. IM model is employed in the CONTRANS code developed by ABB-CE for containment thermodynamic analysis. The various partition models were applied to the double ended discharge leg slot break (DEDLS) LOCA which is containment design base accident (CDBA) for Ulchin 3 and 4 PSAR. It was shown that IM model is the most conservative for containment design pressure analysis. Results of the CONTRANS analyses are compared with those of UCN PSAR for which CONTEMPT-LT code was used

  12. Probabilistic evaluation of concrete containment capacity for beyond design basis internal pressures

    International Nuclear Information System (INIS)

    Tang, H.T.; Dameron, R.A.; Rashid, Y.R.

    1995-01-01

    For beyond design basis internal pressure loading, experimental studies have demonstrated that the most probable failure mode governing the ultimate functional capacity of concrete containments is leak rather than break. Based on leak rates measured in experiments, a prediction formula for leak rate as functions of containment liner size and internal pressure has been postulated. The determination of liner tear is cast in a probabilistic framework. In calculating leakage, particular attention is paid to the evaluation of leakage versus rupture and the loading rates that may be required to leapfrog over a leakage mode. (orig.)

  13. BBRV post-tensioning systems as applied to reactor containments and prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Thorpe, W.; Speck, F.E.

    1976-01-01

    Nuclear containments and pressure vessels can be post-tensioned by using two basically different methods: tendons and winding. The fundamental differences between the two concepts are shown by introductory examples. A discussion of tendon units, usually lying in the range 4000 to 10,000 kN, is followed by a detailed presentation of the BBRV winding system. After giving a short comment to factors influencing the choice of a post-tensioning system the authors discuss specific aspects of some application groups: cable layout with containments and pressure vessels, conditions for a wrapped design, corrosion protection. (author)

  14. New W-and Mo-containing perovskites sythesized at high pressure

    Energy Technology Data Exchange (ETDEWEB)

    Sevast' yanova, L G; Burdina, K P; Zubova, E V; Venevtsev, Yu N [Moskovskij Gosudarstvennyj Univ. (USSR); Nauchno-Issledovatel' skij Fiziko-Khimicheskij Inst., Moscow (USSR))

    1979-11-01

    The possibility of synthesizing complex oxide W and Mo-containing compounds having a perovskite structure is shown. The optimum synthesis conditions have been defined. Critical pressure Psub(cr) has been found to equal 70 kbar, above which the perovskite structure can still exist at room temperature. The ''pressure-temperature'' diagram was used to define the stability region of perovskite of Pb(HgMo)sub(1/2)Osub(3)composition, bound by pressure p=35 to 50 kbar and a temperature of 700 deg C.

  15. Enhancement of fatigue crack growth rates in pressure boundary materials due to light-water-reactor environments

    International Nuclear Information System (INIS)

    VanDerSluys, W.A.; Emanuelson, R.H.

    1988-01-01

    The high level of reliability required of the primary-coolant pressure boundary in a nuclear reactor system leads to a continuing interest in the interaction among the coolant, pressure boundary materials, and service loadings. One area of concern involves the possible enhancement of the growth rate of fatigue cracks due to the coolant. Advances have occurred recently toward a better understanding of the variables influencing the material/environment interactions and methods of addressing this interaction. Sulfur now appears to be one of the principal agents responsible for the observed enhancement of the fatigue crack growth rates in light-water-reactor (LWR) environments. This paper presents the results of investigations on the effect of sulfur in the steel, bulk water environment, and at the crack tip

  16. Aging characteristics of containment building and sensitivity on ultimate pressure capacity

    International Nuclear Information System (INIS)

    Seo, Jeong Moon; Choun, Young Sun; Choi, In Kil; Ha, Jae Joo

    1998-03-01

    For the reliable safety assessment of the containment building, structural and material conditions can be investigated in detail and pertinent assessment technologies have to be established. Also, an understanding on the aging-related degradations for the construction materials is required to predict long-term structural safety of the containment building. For the development of reliable aging prediction models, an extensive data base system related to aging properties of the containment building has to be prepared. The objectives of this research are to develop aging models representing long-term degradation of materials and a structural performance assessment program for containment building considering aging-related degradation. According to the results of sensitivity analysis, as the mechanical properties of the constituent materials degrade, the ultimate pressure capacity of containment building may decrease and severe damage may occur around the mid-level of the containment wall. (author). 28 refs., 11 tabs., 36 figs

  17. Bounding analysis of containment of high pressure melt ejection in advanced light water reactors

    International Nuclear Information System (INIS)

    Additon, S.L.; Fontana, M.H.; Carter, J.C.

    1990-01-01

    This paper reports on the loadings on containment due to direct containment heating (DCH) as a result of high pressure melt ejection (HPME) in advanced light water reactors (ALWR) which were estimated using conservative, bounding analyses. The purpose of the analyses was to scope the magnitude of the possible loadings and to indicate the performance needed from potential mitigation methods, such as a cavity configuration that limits energy transfer to the upper containment volume. Analyses were performed for three cases which examined the effect of availability of high pressure reactor coolant system water at the time of reactor vessel melt through and the effect of preflooding of the reactor cavity. The amount of core ejected from the vessel was varied from 100% to 0% for all cases. Results indicate that all amounts of core debris dispersal could be accommodated by the containment for the case where the reactor cavity was preflooded. For the worst case, all the energy from in-vessel hydrogen generation and combustion plus that from 45% of the entire molten core would be required to equilibrate with the containment upper volume in order to reach containment failure pressure

  18. Effect of Operating Pressure on Hydrogen Risk in Filtered Containment Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Na, Young Su; Cho, Song-Won; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The FCVS (Filtered Containment Venting System) has the main objectives of both the depressurization in the containment building and the decontamination of fission products generated under a severe accident. One of the commercial wet-type FCVSs consists of a cylindrical pressure vessel including a scrubbing solution and filters. A FCVS vessel can be installed on the outside of the containment building, and is connected with the containment through a pipe. When the pressure in the containment building approaches the setting value, a valve on a pipe between the containment and the FCVS opens to operate the FCVS. The amount of steam and gas mixtures generated under a severe accident can be released into the FCVS, where the nozzles of a pipe are submerged into a scrubbing solution in a FCVS vessel. Non-condensable gases and fine aerosols can enter a scrubbing solution, and they then pass the filters. The decontaminated gases are finally discharged from the FCVS into the outside environment. Previous studies have introduced critical issues with the operation of the FCVS. Reference [2] assessed the effect of the operating pressure of the FCVS on the hydrogen risk in a FCVS vessel. The volumetric concentrations of hydrogen and steam in a postulated FCVS with a 3 m diameter and 6.5 m height were calculated using the MELCOR computer code (v. 1.8.6). After the operation of the FCVS, the pressure and temperature in the FCVS vessel jumped from the initial conditions of the atmosphere pressure and room temperature. For the FCVS operating pressure of 5 bar, the hydrogen concentration increased from 6% in the containment to 14% in a FCVS vessel, whereas the steam concentration decreased from 58% in the containment to 3% in a FCVS vessel. The increased hydrogen concentration with air in a FCVS vessel can exists within the region of the burn limit in the Shapiro diagram. This possibility of the hydrogen combustion can threaten the integrity of the FCVS. To mitigate the hydrogen risk

  19. Development of a double containment concept for the European pressurized water reactor

    International Nuclear Information System (INIS)

    Costaz, J.L.; Bonhomme, N.; L'Huby, Y.; Sidaner, J.F.

    1994-01-01

    This paper addresses the development of a double containment concept for the European Pressurized Water Reactor. Specification of containment leak tightness during severe hazards resulting from core melt scenarios is part of the safety goals defined for the EPR project. These safety goals include retention of molten core, mitigation of hydrogen deflagration or explosion risks and decay heat removal. The main new containment structural design loads which have been defined, including containment pressure and temperature conditions following possible postulated-core melt events, are recalled in the paper. The feasibility of a double containment with a prestressed concrete inner containment taking into account these new design loads but based upon experience gained within the well tested concept of concrete double wall containment used in 1400 MW nuclear power plants which have already been built in France, is presented. The main characteristics of such a prestressed inner containment are described. Limits and further possible optimization for even more severe design loads (including liner option) are indicated. Experimental works including a large scale mock up are already under way. (author). 2 refs., 4 figs

  20. Recyclability of mixed office waste papers containing pressure sensitive adhesives and silicone release liners

    Science.gov (United States)

    Julie Hess; Roberta Sena-Gomes; Lisa Davie; Marguerite Sykes

    2001-01-01

    Increased use of pressure sensitive adhesives for labels and stamps has introduced another contaminant into the office paper stream: silicone- coated release liners. This study examines methods and conditions for removal of contaminants, including these liners, from a typical batch of discarded office papers. Removal of contaminants contained in the furnish were...

  1. ZOCO V - a computer code for the calculation of time-dependent spatial pressure distribution in reactor containments

    International Nuclear Information System (INIS)

    Mansfeld, G.; Schally, P.

    1978-06-01

    ZOCO V is a computer code which can calculate the time- and space- dependent pressure distribution in containments of water-cooled nuclear power reactors (both full pressure containments and pressure suppression systems) following a loss-of-coolant accident, caused by the rupture of a main coolant or steam pipe

  2. Hydrogen behavior in a large-dry pressurized water reactor containment building during a severe accident

    International Nuclear Information System (INIS)

    Hsu Wensheng; Chen Hungpei; Hung Zhenyu; Lin Huichen

    2014-01-01

    Following severe accidents in nuclear power plants, large quantities of hydrogen may be generated after core degradation. If the hydrogen is transported from the reactor vessel into the containment building, an explosion might occur, which might threaten the integrity of the building; this can ultimately cause the release of radioactive materials. During the Fukushima Daiichi nuclear accident in 2011, the primary containment structures remained intact but contaminated fragments broke off the secondary containment structures, which disrupted mitigation activities and triggered subsequent explosions. Therefore, the ability to predict the behavior of hydrogen after severe accidents may facilitate the development of effective nuclear reactor accident management procedures. The present study investigated the behavior of hydrogen in a large-dry pressurized water reactor (PWR). The amount of hydrogen produced was calculated using the Modular Accident Analysis Program. The hydrogen transport behavior and the effect of the explosion on the PWR containment building were simulated using the Flame Acceleration Simulator. The simulation results showed that the average hydrogen volume fraction is approximately 7% in the containment building and that the average temperature is 330 K. The maximum predicted pressure load after ignition is 2.55 bar, which does not endanger the structural integrity of the containment building. The results of this investigation indicate that the hydrogen mitigation system should be arranged on both the upper and lower parts of the containment building to reduce the impact of an explosion. (author)

  3. Reynolds stress structures in a self-similar adverse pressure gradient turbulent boundary layer at the verge of separation.

    Science.gov (United States)

    Atkinson, C.; Sekimoto, A.; Jiménez, J.; Soria, J.

    2018-04-01

    Mean Reynolds stress profiles and instantaneous Reynolds stress structures are investigated in a self-similar adverse pressure gradient turbulent boundary layer (APG-TBL) at the verge of separation using data from direct numerical simulations. The use of a self-similar APG-TBL provides a flow domain in which the flow gradually approaches a constant non-dimensional pressure gradient, resulting in a flow in which the relative contribution of each term in the governing equations is independent of streamwise position over a domain larger than two boundary layer thickness. This allows the flow structures to undergo a development that is less dependent on the upstream flow history when compared to more rapidly decelerated boundary layers. This APG-TBL maintains an almost constant shape factor of H = 2.3 to 2.35 over a momentum thickness based Reynolds number range of Re δ 2 = 8420 to 12400. In the APG-TBL the production of turbulent kinetic energy is still mostly due to the correlation of streamwise and wall-normal fluctuations, 〈uv〉, however the contribution form the other components of the Reynolds stress tensor are no longer negligible. Statistical properties associated with the scale and location of sweeps and ejections in this APG-TBL are compared with those of a zero pressure gradient turbulent boundary layer developing from the same inlet profile, resulting in momentum thickness based range of Re δ 2 = 3400 to 3770. In the APG-TBL the peak in both the mean Reynolds stress and the production of turbulent kinetic energy move from the near wall region out to a point consistent with the displacement thickness height. This is associated with a narrower distribution of the Reynolds stress and a 1.6 times higher relative number of wall-detached negative uv structures. These structures occupy 5 times less of the boundary layer volume and show a similar reduction in their streamwise extent with respect to the boundary layer thickness. A significantly lower percentage

  4. High pressure melt ejection (HPME) and direct containment heating (DCH): state-of-the-art report

    International Nuclear Information System (INIS)

    1996-12-01

    This report first address the accident considerations leading to conditions with the reactor pressure vessel at a significant pressure. It also address those accident management actions that could prevent such a pressurized state and the effectiveness of operator actions since this is a principal focus of how a HPME could be prevented. Furthermore, it also investigates those situations, while very unlikely, in which the RCS could be at a significant pressure and possibly experience RPV failure. This represents a significant set of experimental information that, coupled with the integral effects models, provides the necessary insights for issue resolution for a number of containment types. Lastly, conclusions and recommendations are developed to be presented to the CSNI

  5. Mark II containment 1/6-scale pressure suppression test program: data report no. 2

    International Nuclear Information System (INIS)

    Kukita, Yutaka; Okazaki, Motoaki; Namatame, Ken; Shiba, Masayoshi

    1979-08-01

    This report documents experimental data from the first test phase of the Mark II Containment 1/6-Scale Pressure Suppression Test. The 1/6-Scale Test was initiated in December, 1976, to investigate the thermohydraulic responses of a BWR Mark II pressure suppression system to a postulated loss-of-coolant accident (LOCA), by means of scale model experiments. From January to June, 1977, a series of tests were performed for the Japanese BWR Owners' Group. These tests consisted of eight air-blowdown pool swell tests, three steam-blowdown pool swell tests, and twelve steam condensation tests. The dynamic responses of pressure and pool water level during the blowdown, pressure oscillation and chugging phenomena associated with unsteady condensation of steam were measured. (author)

  6. Nonlinear failure analysis of a reinforced concrete containment under internal pressure

    International Nuclear Information System (INIS)

    Sharma, S.; Wang, Y.K.; Reich, M.

    1984-01-01

    A detailed nonlinear finite element model is used to investigate the failure response of the Indian Point containment building under severe accident pressures. Refined material models are used to describe the complex stress-strain behavior of the liner and rebar steels, the plain concrete and the reinforced concrete. Structural geometry of the containment is idealized by eight layers of axisymmetric finite elements through the wall thickness in order to closely model the actual placement of the rebars. Soil stiffness under the containment base mat is modeled by a series of nonlinear spring elements. Numerical results presented in the paper describe cracking and plastic deformation (in compression) of the concrete, yielding of the liner and rebar steels and eventual loss of the load carrying capacity of the containment. The results are compared with available data from the previous studies for this containment. 8 references, 9 figures

  7. Use of Dimples to Suppress Boundary Layer Separation on a Low Pressure Turbine Blade

    Science.gov (United States)

    2002-12-01

    thermocouples. A Druck LPM 5481 pressure transducer is connected to an SCXI-1121 signal conditioning card. It has a range of -0.2 to 0.8 in H2O...tapped blades. 71 4.2.1 Pressure Instrumentation The primary interface for all measurements taken during this research is the Druck LPM 5481...tester. Figure 48 shows a schematic of the Pressurements V1600/ 3D dead-weight tester. Force = (m)(g) Regulator Volume Volume Supply Pressure

  8. The structure of a three-dimensional boundary layer subjected to streamwise-varying spanwise-homogeneous pressure gradient

    International Nuclear Information System (INIS)

    Bentaleb, Y.; Leschziner, M.A.

    2013-01-01

    Highlights: • We study a spatially-evolving three-dimensional boundary layer. • We impose a streamwise-varying spanwise-homogeneous pressure gradient. • A collateral flow is formed close to the wall, and this is investigated alongside the skewed upper part of the boundary layer. • A wide range of flow-physical properties have been studied. -- Abstract: A spatially-evolving three-dimensional boundary layer, subjected to a streamwise-varying spanwise-homogeneous pressure gradient, equivalent to a body force, is investigated by way of direct numerical simulation. The pressure gradient, prescribed to change its sign half-way along the boundary layer, provokes strong skewing of the velocity vector, with a layer of nearly collateral flow forming close to the wall up to the position of maximum spanwise velocity. A wide range of flow-physical properties have been studied, with particular emphasis on the near-wall layer, including second-moments, major budget contributions and wall-normal two-point correlations of velocity fluctuations and their angles, relative to wall-shear fluctuations. The results illustrate the complexity caused by skewing, including a damping in turbulent mixing and a significant lag between strains and stresses. The study has been undertaken in the context of efforts to develop and test novel hybrid LES–RANS schemes for non-equilibrium near-wall flows, with an emphasis on three-dimensional near-wall straining. Fundamental flow-physical issues aside, the data derived should be of particular relevance to a priori studies of second-moment RANS closure and the development and validation of RANS-type near-wall approximations implemented in LES schemes for high-Reynolds-number complex flows

  9. Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

  10. High-pressure catalytic chemical vapor deposition of ferromagnetic ruthenium-containing carbon nanostructures

    Energy Technology Data Exchange (ETDEWEB)

    Khavrus, Vyacheslav O., E-mail: V.Khavrus@ifw-dresden.de; Ibrahim, E. M. M.; Bachmatiuk, Alicja; Ruemmeli, Mark H.; Wolter, A. U. B.; Hampel, Silke; Leonhardt, Albrecht [IFW Dresden (Germany)

    2012-06-15

    We report on the high-pressure catalytic chemical vapor deposition (CCVD) of ruthenium nanoparticles (NPs) and single-walled carbon nanotubes (SWCNTs) by means of gas-phase decomposition of acetonitrile and ruthenocene in a tubular quartz flow reactor at 950 Degree-Sign C and at elevated pressures (between 2 and 8 bar). The deposited material consists of Ru metal cores with sizes ranging between 1 and 3 nm surrounded by a carbon matrix. The high-pressure CCVD seems to be an effective route to obtain composite materials containing metallic NPs, Ru in this work, inside a nanostructured carbon matrix protecting them from oxidation in ambient air. We find that in contradiction to the weak paramagnetic properties characterizing bulk ruthenium, the synthesized samples are ferromagnetic as predicted for nanosized particles of nonmagnetic materials. At low pressure, the very small ruthenium catalyst particles are able to catalyze growth of SWCNTs. Their yield decreases with increasing reaction pressure. Transmission electron microscopy, selected area energy-dispersive X-ray analysis, Raman spectroscopy, and magnetic measurements were used to analyze and confirm properties of the synthesized NPs and nanotubes. A discussion on the growth mechanism of the Ru-containing nanostructures is presented.

  11. Development of pressure containment and damage tolerance technology for composite fuselage structures in large transport aircraft

    Science.gov (United States)

    Smith, P. J.; Thomson, L. W.; Wilson, R. D.

    1986-01-01

    NASA sponsored composites research and development programs were set in place to develop the critical engineering technologies in large transport aircraft structures. This NASA-Boeing program focused on the critical issues of damage tolerance and pressure containment generic to the fuselage structure of large pressurized aircraft. Skin-stringer and honeycomb sandwich composite fuselage shell designs were evaluated to resolve these issues. Analyses were developed to model the structural response of the fuselage shell designs, and a development test program evaluated the selected design configurations to appropriate load conditions.

  12. Valency state changes in lanthanide-contained systems under high pressure

    Energy Technology Data Exchange (ETDEWEB)

    Jayaraman, A

    1980-08-01

    Changes in valency state induced by pressure in samarium sulphide SmS remind one of alchemy, as the mat black initial substance shines golden after the electron transition. The alchemist's dream is of course not realized, however the compound does exhibit an unusually interesting behaviour in the new state. The valency state of samarium as newly appeared fluctuated very rapidly between two electron configurations. Manipulation of the valency state by pressure or chemical substitution can basically change the physical properties of systems containing lanthanides. The phenomena are described and discussed in the following survey.

  13. Research on the behaviour of pressure suppression containment systems carried out at the University of Pisa

    International Nuclear Information System (INIS)

    Bigi, R.; Bovalini, R.; Mazzini, M.; Micheletti, E.

    1978-01-01

    A research programme has been carried out at the University of Pisa to study the thermo-hydraulic transient in pressure suppression containment systems during a LOCA. In the first series of experimental tests remarkable oscillations of pressure were observed both in dry and in wet-well. In order to describe these dynamic phenomena, a mathematical model has been set up; the main out-lines of this model are briefly described and the comparison between the calculated and experimental results is reported. (author)

  14. Axisymmetric global structural analysis of BARC prestressed concrete containment model for beyond design pressure

    International Nuclear Information System (INIS)

    Singh, Tarvinder; Singh, R.K.; Ghosh, A.K.

    2008-10-01

    In order to check the adequacy of the Indian Pressurized Heavy Water Reactor (PHWR) containment structure to withstand severe accident induced internal pressure load, the ultimate load capacity assessment is required. Reactor Safety Division (RSD) of Bhabha Atomic Research Centre (BARC) has initiated an experimental program at BARC Tarapur Containment Test Facility to evaluate the ultimate load capacity of Indian PHWR containment. For this study, BARC Containment Model (BARCOM), which is 1:4 scale representation of Tarapur Atomic Power Station (TAPS) unit-3 and 4 540 MWe PHWR Inner Containment of Pre-stressed Concrete has been constructed. The model includes all the important major design features of the prototype containment and simulates Main Air Lock (MAL), Steam Generator (SG), Emergency Air Lock (EAL) and Fueling Machine Air Lock (FMAL) openings. The design pressure (Pd) of BARCOM is 1.44kg/cm 2 (g), which is same as the prototype. The pretest analysis of BARCOM has been performed with finite element axi-symmetric modeling. The objective of this simulation was to understand the behavior of containment model under internal pressure and find out the various failure modes and critical locations important for instrumentation during the experiment. The structural response of the containment model is assessed in terms of wall and dome displacement; cracking of concrete, longitudinal and hoop strains and stresses. Another objective of the analysis was to predict the various failure modes of BARCOM with regard to the concrete cracking, reinforcement yielding and tendon inelastic behavior along with the estimation of the ultimate load capacity of the containment model. It is noted that the BARCOM has an ultimate load capacity factor of 3.54 Pd. However, further analysis is needed to quantify the factor of safety with detail 3D model, which should account for the local structural behavior due to various openings. Meanwhile, this preliminary simplified analysis helps to

  15. From Topos to Oikos: The Standardization of Glass Containers as Epistemic Boundaries in Modern Laboratory Research (1850-1900).

    Science.gov (United States)

    Espahangizi, Kijan

    2015-09-01

    Glass vessels such as flasks and test tubes play an ambiguous role in the historiography of modern laboratory research. In spite of the strong focus on the role of materiality in the last decades, the scientific glass vessel - while being symbolically omnipresent - has remained curiously neglected in regard to its materiality. The popular image or topos of the transparent, neutral, and quasi-immaterial glass container obstructs the view of the physico-chemical functionality of this constitutive inner boundary in modern laboratory environments and its material historicity. In order to understand how glass vessels were able to provide a stable epistemic containment of spatially enclosed experimental phenomena in the new laboratory ecologies emerging in the nineteenth and early twentieth century, I will focus on the history of the material standardization of laboratory glassware. I will follow the rise of a new awareness for measurement errors due to the chemical agency of experimental glass vessels, then I will sketch the emergence of a whole techno-scientific infrastructure for the improvement of glass container quality in late nineteenth-century Germany. In the last part of my argument, I will return to the laboratory by looking at the implementation of this glass reform that created a new oikos for the inner experimental milieus of modern laboratory research.

  16. Transient integral boundary layer method to calculate the translesional pressure drop and the fractional flow reserve in myocardial bridges

    Directory of Open Access Journals (Sweden)

    Möhlenkamp Stefan

    2006-06-01

    Full Text Available Abstract Background The pressure drop – flow relations in myocardial bridges and the assessment of vascular heart disease via fractional flow reserve (FFR have motivated many researchers the last decades. The aim of this study is to simulate several clinical conditions present in myocardial bridges to determine the flow reserve and consequently the clinical relevance of the disease. From a fluid mechanical point of view the pathophysiological situation in myocardial bridges involves fluid flow in a time dependent flow geometry, caused by contracting cardiac muscles overlying an intramural segment of the coronary artery. These flows mostly involve flow separation and secondary motions, which are difficult to calculate and analyse. Methods Because a three dimensional simulation of the haemodynamic conditions in myocardial bridges in a network of coronary arteries is time-consuming, we present a boundary layer model for the calculation of the pressure drop and flow separation. The approach is based on the assumption that the flow can be sufficiently well described by the interaction of an inviscid core and a viscous boundary layer. Under the assumption that the idealised flow through a constriction is given by near-equilibrium velocity profiles of the Falkner-Skan-Cooke (FSC family, the evolution of the boundary layer is obtained by the simultaneous solution of the Falkner-Skan equation and the transient von-Kármán integral momentum equation. Results The model was used to investigate the relative importance of several physical parameters present in myocardial bridges. Results have been obtained for steady and unsteady flow through vessels with 0 – 85% diameter stenosis. We compare two clinical relevant cases of a myocardial bridge in the middle segment of the left anterior descending coronary artery (LAD. The pressure derived FFR of fixed and dynamic lesions has shown that the flow is less affected in the dynamic case, because the distal

  17. Heat removal tests for pressurized water reactor containment spray by largescale facility

    International Nuclear Information System (INIS)

    Motoki, Y.; Hashimoto, K.; Kitani, S.; Naritomi, M.; Nishio, G.; Tanaka, M.

    1983-01-01

    Heat removal tests for pressurized water reactor (PWR) containment spray were carried out to investigate effectiveness of the depressurization by Japan Atomic Energy Research Institute model containment (7-m diameter, 20 m high, and 708-m 3 volume) with PWR spray nozzles. The depressurization rate is influenced by the spray heat transfer efficiency and the containment wall surface heat transfer coefficient. The overall spray heat transfer efficiency was investigated with respect to spray flow rate, weight ratio of steam/air, and spray height. The spray droplet heat transfer efficiency was investigated whether the overlapping of spray patterns gives effect or not. The effect was not detectable in the range of large value of steam/air, however, it was better in the range of small value of it. The experimental results were compared with the calculated results by computer code CONTEMPT-LT/022. The overall spray heat transfer efficiency was almost 100% in the containment pressure, ranging from 2.5 to 0.9 kg/cm 2 X G, so that the code was useful on the prediction of the thermal hydraulic behavior of containment atmosphere in a PWR accident condition

  18. High pressure sample container for thermal neutron spectroscopy and diffraction on strongly scattering fluids

    International Nuclear Information System (INIS)

    Verkerk, P.; Pruisken, A.M.M.

    1979-01-01

    A description is presented of the construction and performance of a container for thermal neutron scattering on a fluid sample with about 1.5 cm -1 macroscopic cross section (neglecting absorption). The maximum pressure is about 900 bar. The container is made of 5052 aluminium capillary with inner diameter 0.75 mm and wall thickness 0.25 mm; it covers a neutron beam with a cross section of 9 X 2.5 cm 2 . The container has been successfully used in neutron diffraction and time-of-flight experiments on argon-36 at 120 K and several pressures up to 850 bar. It is shown that during these measurements the temperature gradient over the sample as well as the error in the absolute temperature were both less than 0.05 K. Subtraction of the Bragg peaks due to container scattering in diffraction experiments may be dfficult, but seems feasible because of the small amount of aluminium in the neutron beam. Correction for container scattering and multiple scattering in time-of-flight experiments may be difficult only in the case of coherently scattering samples and small scattering angles. (Auth.)

  19. 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)

  20. A three-dimensional rupture analysis of steel liners anchored to concrete pressure and containment vessels

    International Nuclear Information System (INIS)

    Bangash, Y.

    1987-01-01

    Steel liners or plates are anchored to concrete pressure and containment vessels for nuclear and offshore facilities. Due to extreme loading conditions a liner may buckle due to the pull-out or shearing of anchors from the base metal and concrete. Under certain conditions attributed to loadings, liner metal deterioration and cracking of concrete behind the liner, the liner may fail by rupture. This paper presents a three-dimensional analysis of steel-concrete elements, using finite elements analysis in which a provision is made for liner instability, anchor strength and stiffness, concrete cracking and finally liner rupture. The analysis is tested first on an octagonal slab with and without an anchored steel liner. It is then extended to concrete pressure and containment vessels. The analytical results obtained are compared well with those available from the experimental tests and other sources. (author)

  1. Over-pressure test on BARCOM pre-stressed concrete containment

    Energy Technology Data Exchange (ETDEWEB)

    Parmar, R.M.; Singh, Tarvinder; Thangamani, I.; Trivedi, Neha; Singh, Ram Kumar, E-mail: rksingh@barc.gov.in

    2014-04-01

    Bhabha Atomic Research Centre (BARC), Trombay has organized an International Round Robin Analysis program to carry out the ultimate load capacity assessment of BARC Containment (BARCOM) test model. The test model located in BARC facilities Tarapur; is a 1:4 scale representation of 540 MWe Pressurized Heavy Water Reactor (PHWR) pre-stressed concrete inner containment structure of Tarapur Atomic Power Station (TAPS) unit 3 and 4. There are a large number of sensors installed in BARCOM that include vibratory wire strain gauges of embedded and spot-welded type, surface mounted electrical resistance strain gauges, dial gauges, earth pressure cells, tilt meters and high resolution digital camera systems for structural response, crack monitoring and fracture parameter measurement to evaluate the local and global behavior of the containment test model. The model has been tested pneumatically during the low pressure tests (LPTs) followed by proof test (PT) and integrated leakage rate test (ILRT) during commissioning. Further the over pressure test (OPT) has been carried out to establish the failure mode of BARCOM Test-Model. The over-pressure test will be completed shortly to reach the functional failure of the test model. Pre-test evaluation of BARCOM was carried out with the results obtained from the registered international round robin participants in January 2009 followed by the post-test assessment in February 2011. The test results along with the various failure modes related to the structural members – concrete, rebars and tendons identified in terms of prescribed milestones are presented in this paper along with the comparison of the pre-test predictions submitted by the registered participants of the Round Robin Analysis for BARCOM test model.

  2. F-8 supercritical wing flight pressure, Boundary layer, and wake measurements and comparisons with wind tunnel data

    Science.gov (United States)

    Montoya, L. C.; Banner, R. D.

    1977-01-01

    Data for speeds from Mach 0.50 to Mach 0.99 are presented for configurations with and without fuselage area-rule additions, with and without leading-edge vortex generators, and with and without boundary-layer trips on the wing. The wing pressure coefficients are tabulated. Comparisons between the airplane and model data show that higher second velocity peaks occurred on the airplane wing than on the model wing. The differences were attributed to wind tunnel wall interference effects that caused too much rear camber to be designed into the wing. Optimum flow conditions on the outboard wing section occurred at Mach 0.98 at an angle of attack near 4 deg. The measured differences in section drag with and without boundary-layer trips on the wing suggested that a region of laminar flow existed on the outboard wing without trips.

  3. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1975-01-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

  4. Theoretical and experimental investigations into the filtration of the atmosphere within the containments of pressurized water reactors after serious reactor accidents

    International Nuclear Information System (INIS)

    Dillmann, H.G.; Pasler, H.

    1981-01-01

    For serious accidents in nuclear power stations equipped with pressurized water reactors and with boundary conditions assumed, a conservative evaluation was made of the condition of the atmosphere within the reactor containment, particularly referring to pressure, temperature, air humidity and activity release. Based on these data the loads were calculated of accident filter systems of different designs as a function of parameters such as the course of releases and the volume flow through the filter systems. A number of experimental results are indicated on the behaviour of iodine sorption materials under extreme conditions including the least favorable temperature, humidity and pressure derived from the calculations above. Reference is made to the targets of future R and D work on aerosol removal

  5. Experimental study of the structural behavior of the reinforced concrete containment vessel beyond design pressure

    International Nuclear Information System (INIS)

    Oyamada, O.; Saito, H.; Muramatsu, Y.; Hasegawa, T.; Tanaka, N.

    1990-01-01

    The first Advanced Boiling Water Reactor (ABWR) including a reinforced concrete containment vessel (RCCV) is scheduled to be constructed in the 1990s, in Japan. As the RCCV is new to Japan, we performed a trial design, several series of fundamental experiments and partial/total model experiments. This paper presents a summary of the 'TOP SLAB EXPERIMENT' carried out as one of partial model experiments, in which the structural behavior of the RCCV was examined under internal pressure. (orig.)

  6. Probabilistic analysis of Millstone Unit 3 ultimate containment failure probability given high pressure: Chapter 14

    International Nuclear Information System (INIS)

    Bickel, J.H.

    1983-01-01

    The quantification of the containment event trees in the Millstone Unit 3 Probabilistic Safety Study utilizes a conditional probability of failure given high pressure which is based on a new approach. The generation of this conditional probability was based on a weakest link failure mode model which considered contributions from a number of overlapping failure modes. This overlap effect was due to a number of failure modes whose mean failure pressures were clustered within a 5 psi range and which had uncertainties due to variances in material strengths and analytical uncertainties which were between 9 and 15 psi. Based on a review of possible probability laws to describe the failure probability of individual structural failure modes, it was determined that a Weibull probability law most adequately described the randomness in the physical process of interest. The resultant conditional probability of failure is found to have a median failure pressure of 132.4 psia. The corresponding 5-95 percentile values are 112 psia and 146.7 psia respectively. The skewed nature of the conditional probability of failure vs. pressure results in a lower overall containment failure probability for an appreciable number of the severe accident sequences of interest, but also probabilities which are more rigorously traceable from first principles

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  8. Application of smart differential pressure transmitters (DPTS) for containment studies facility (CSF)

    International Nuclear Information System (INIS)

    Shanware, V.M.; Gole, N.V.; Sebastian, A.; Subramaniam, K.

    2001-01-01

    Containment Studies Facility (CSF) is being set up in BARC for studying various containment related thermal hydraulic and other processes during simulated conditions of pipe rupture. The set up consists of a model reactor containment vessel with a model primary heat transport system. Besides, provisions exist to introduce aerosols and hydrogen also in the containment model. The instrumentation includes measurement of the process temperatures, pressures, levels, flows, humidity, etc. Differential Pressure Transmitters (DPT) will be used for measurement of levels and flows in the CSF. The procured DPTs for this facility are smart. Conventional transmitters have a rangeability specification of 5 or 6. But the smart transmitters have rangeability varying between 40-100. Smart transmitters have facility to change its operating range online. This enables the provision of zooming in on the selected range and narrowing the range around the point of measurement. This facility can be exploited to realise the maximum possible accuracy at the smallest possible range around the point of measurement. This paper describes how the smart DPTs function, how the Highway Addressable Remote Transmitter (HART) protocol works and how we propose to use the on-line rangeability of these DPTs get the highest resolution in our measurements. (author)

  9. Experimental investigation of iodine removal and containment depressurization in containment spray system test facility of 700 MWe Indian pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jain, Manish [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Kandar, T.K.; Vhora, S.F.; Mohan, Nalini [Directorate of Technology Development, Nuclear Power Corporation of India Limited, Mumbai (India); Iyer, K.N. [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, I.I.T., Bombay, Powai, Mumbai (India)

    2017-05-15

    Highlights: • Depressurization rate in a scaled down vessel filled with air and steam is studied. • Iodine removal rate in a scaled down vessel filled with steam/air is investigated. • Effect of SMD and vessel pressure on depressurization rate is studied. • Depressurization rate decreases with the increase in the droplet size (590 μm – 1 mm) • Decrease in pressure and iodine concentration with time follow exponential trend. - Abstract: As an additional safety measure in the new 700 MWe Indian pressurized heavy water reactors, the first of a kind system called containment Spray System is introduced. The system is designed to cater/mitigate the conditions after design basis accidents i.e., loss of coolant accident and main steam line break. As a contribution to the safety analysis of condition following loss-of-coolant accidents, experiments are carried out to establish the performance of the system. The loss of coolant is simulated by injecting saturated steam and iodine vapors into the containment vessel in which air is enclosed at atmospheric and room temperature, and then the steam-air mixture is cooled by sprays of water. The effect of water spray on the containment vessel pressure and the iodine scrubbing in a scaled down facility is investigated for the containment spray system of Indian pressurized heavy water reactors. The experiments are carried out in the scaled down vessel of the diameter of 2.0 m and height of 3.5 m respectively. Experiments are conducted with water at room temperature as the spray medium. Two different initial vessel pressure i.e. 0.7 bar and 1.0 bar are chosen for the studies as they are nearing the loss of coolant accident & main steam line break pressures in Indian pressurized heavy water reactors. These pressures are chosen based on the containment resultant pressures after a design basis accident. The transient temperature and pressure distribution of the steam in the vessel are measured during the depressurization

  10. 16 CFR 1500.46 - Method for determining flashpoint of extremely flammable contents of self-pressurized containers.

    Science.gov (United States)

    2010-01-01

    ... extremely flammable contents of self-pressurized containers. 1500.46 Section 1500.46 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FEDERAL HAZARDOUS SUBSTANCES ACT REGULATIONS HAZARDOUS SUBSTANCES AND... extremely flammable contents of self-pressurized containers. Use the apparatus described in § 1500.43a. Use...

  11. Cost-benefit evaluation of containment related engineered safety features of Indian pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Bajaj, S.S.; Bhawal, R.N.; Rustagi, R.S.

    1984-01-01

    The typical containment system for a commercial nuclear reactor uses several engineered safety features to achieve its objective of limiting the release of radioactive fission products to the environment in the event of postulated accident conditions. The design of containment systems and associated features for Indian Pressurized Heavy Water Reactors (PHWRs) has undergone progressive improvement in successive projects. In particular, the current design adopted for the Narora Atomic Power Project (NAPP) has seen several notable improvements. The paper reports on a cost-benefit study in respect of three containment related engineered safety features and subsystems of NAPP, viz. (i) secondary containment envelope, (ii) primary containment filtration and pump-back system, and (iii) secondary containment filtration, recirculation and purge system. The effect of each of these systems in reducing the environmental releases of radioactivity following a design basis accident is presented. The corresponding reduction in population exposure and the associated monetary value of this reduction in exposure are also given. The costs of the features and subsystem under consideration are then compared with the monetary value of the exposures saved, as well as other non-quantified benefits, to arrive at conclusions regarding the usefulness of each subsystem. This study clearly establishes for the secondary containment envelope the benefit in terms of reduction in public exposure giving a quantitative justification for the costs involved. In the case of the other two subsystems, which involve relatively low costs, while all benefits have not been quantified, their desirability is justified on qualitative considerations. It is concluded that the engineered safety features adopted in the current containment system design of Indian PHWRs contribute to reducing radiation exposures during accident conditions in accordance with the ALARA ('as low as reasonably achievable') principle

  12. Eliminating Islands in High-pressure Free-boundary Stellarator Magnetohydrodynamic Equilibrium Solutions

    International Nuclear Information System (INIS)

    Hudson, S.R.; Monticello, D.A.; Reiman, A.H.; Boozer, A.H.; Strickler, D.J.; Hirshman, S.P.; Zarnstorff, M.C.

    2002-01-01

    Magnetic islands in free-boundary stellarator equilibria are suppressed using a procedure that iterates the plasma equilibrium equations and, at each iteration, adjusts the coil geometry to cancel resonant fields produced by the plasma. The coils are constrained to satisfy certain measures of engineering acceptability and the plasma is constrained to ensure kink stability. As the iterations continue, the coil geometry and the plasma simultaneously converge to an equilibrium in which the island content is negligible. The method is applied with success to a candidate plasma and coil design for the National Compact Stellarator eXperiment [Physics of Plasma, 7 (2000) 1911

  13. Thermal - hydraulic analysis of pressurizer water reactors using the model of open lateral boundary

    International Nuclear Information System (INIS)

    Borges, R.C.

    1980-10-01

    A computational method is developed for thermal-hydraulic analysis, where the channel may be analysed by more than one independent steps of calculation. This is made possible by the incorporation of the model of open lateral boundary in the code COBRA-IIIP, which permits the determination of the subchannel of an open lattice PWR core in a multi-step calculation. The thermal-hydraulic code COBRA-IIIP, developed at the Massachusetts Institute of Technology, is used as the basic model for this study. (Author) [pt

  14. NFAP calculation of pressure response of 1/6th scale model containment structure

    International Nuclear Information System (INIS)

    Costantino, C.J.; Pepper, S.; Reich, M.

    1988-01-01

    The details associated with the NFAP calculation of the pressure response of the 1/6th scale model containment structure are discussed in this paper. Comparisons are presented of some of the primary items of interest with those determined from the experiment. It was found from this comparison that the hoop response of the containment wall was adequately predicted by the NFAP finite element calculation, including the response in the high pressure, high strain range at which cracking of the concrete and yielding of the hoop reinforcement occurred. In the vertical or meridional direction, it was found that the model was significantly softer than predicted by the finite element calculation; that is, the vertical strains in the test were three to four times larger than computed in the NFAP calculation. These differences were noted even at low strain levels at which the concrete would not be expected to be cracked under tensile loadings. Simplified calculations for the containment indicate that the vertical stiffness of the wall is similar to that which would be determined by assuming the concrete fully cracked. Thus, the experiment indicates an anomalous behavior in the vertical direction

  15. Extreme accident mitigation - analysis of a low pressure secondary containment building

    International Nuclear Information System (INIS)

    Vaughan, G.J.; Dunbar, I.H.

    1987-01-01

    Although whole core accidents are sufficiently unlikely as to be beyond the design basis, the Secondary Containment Building [SCB] is expected to have some effect in mitigating the consequences of such accidents. From a design point of view there are many advantages in having a low pressure SCB fitted with a filtered vent, so studies have been undertaken of the response of such a building to the large sodium fires that might follow a severe accident. The behaviour of the sodium oxide aerosols has been studied using the code AEROSIM. The efficiency of an aerosol scrubber has been investigated experimentally. A simple code, SECCONTAIN, has been developed to model the effects of sodium fires in buildings, and has been applied to a specific design of a low pressure SCB. (author)

  16. Effect of boundary conditions on pressure behavior of finite-conductivity fractures in bounded stratified reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Osman, Mohammed E.; Abou-Kassem, J.H. [Chemical and Petroleum Engineering Department, UAE University, Al-Ain (United Arab Emirates)

    1996-08-15

    In this study, a mathematical model was developed to model the pressure behavior of a well located in a bounded multilayer reservoir and crossed by a finite-conductivity vertical fracture. It was found that the dimensionless pressure function and its derivative strongly depend on fracture conductivity and fracture extension during early times. The effect of reservoir heterogeneity on the pressure function is negligible compared to that on the pressure derivative. Both functions exhibit four flow periods: bilinear, formation linear, pseudoradial and pseudosteady-state which are separated by transition periods. One or more of these flow periods may be missing. Data obtained from a long test and which are characterized by a unit slope line indicate that the well is intercepted by deeply extended fractures. It has been found that the fractional production rates of different layers are a good measure of reservoir and fracture characteristics. Flowmeter survey data can be used to eliminate the non-uniqueness problem when using the type curves presented in this study

  17. Discrete multi-physics simulations of diffusive and convective mass transfer in boundary layers containing motile cilia in lungs.

    Science.gov (United States)

    Ariane, Mostapha; Kassinos, Stavros; Velaga, Sitaram; Alexiadis, Alessio

    2018-04-01

    In this paper, the mass transfer coefficient (permeability) of boundary layers containing motile cilia is investigated by means of discrete multi-physics. The idea is to understand the main mechanisms of mass transport occurring in a ciliated-layer; one specific application being inhaled drugs in the respiratory epithelium. The effect of drug diffusivity, cilia beat frequency and cilia flexibility is studied. Our results show the existence of three mass transfer regimes. A low frequency regime, which we called shielding regime, where the presence of the cilia hinders mass transport; an intermediate frequency regime, which we have called diffusive regime, where diffusion is the controlling mechanism; and a high frequency regime, which we have called convective regime, where the degree of bending of the cilia seems to be the most important factor controlling mass transfer in the ciliated-layer. Since the flexibility of the cilia and the frequency of the beat changes with age and health conditions, the knowledge of these three regimes allows prediction of how mass transfer varies with these factors. Copyright © 2018 Elsevier Ltd. All rights reserved.

  18. Creep deformation-induced antiphase boundaries in L12-containing single-crystal cobalt-base superalloys

    International Nuclear Information System (INIS)

    Eggeler, Yolita M.; Titus, Michael S.; Suzuki, Akane; Pollock, Tresa M.

    2014-01-01

    Creep-induced antiphase boundaries (APBs) in new Co-base single-crystal superalloys with coherent embedded L1 2 -γ′ precipitates have been observed. APBs formed during single-crystal tensile creep tests performed at 900 °C under vacuum at stresses between 275 and 310 MPa. The alloys investigated contained 30–39 at.% Ni, which was added to the Co–Al–W ternary system to expand the γ–γ′ phase field and increase the γ′-solvus. Transmission electron microscopy (TEM) using two-beam conditions with fundamental and superlattice reflections was performed for defect characterization. The Burgers vector b of dislocations associated with the APBs was determined to be of type b = a 0 /2[011] and a 0 /2[011 ¯ ]. The displacement vectors, R, of the APBs matched the dislocation Burgers vectors, with R = b = a 0 /2[011]. APBs were observed in nearly every precipitate beyond 0.5% creep strain for the compositions investigated. The implications for high-temperature properties are discussed

  19. Instrumentation availability for a pressurized water reactor with a large dry containment during severe accidents

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1991-03-01

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a pressurized water reactor with a large dry containment. Results from this evaluation include the following: (a) identification of plant conditions that would impact instrument performance and information needs during severe accidents, (b) definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences, and (c) assessment of the availability of plant instrumentation during severe accidents. 16 refs., 3 figs., 4 tabs

  20. Structure of the AZ91 alloy pressure castings fabricated of home scrap containing charge

    Directory of Open Access Journals (Sweden)

    Z. Konopka

    2011-04-01

    Full Text Available The influence of the AZ91 alloy home scrap addition to the metal charge on both the structure and the selected mechanical propertiesof pressure castings was examined in this article. Two heats were made using different components, the first with only pure AZ91 alloyingots in the charge, and the second containing 30 wt % of home scrap. The hot chamber 3 MN machine was used for casting. Thestructures of the castings and their Brinell hardness were examined for both cases. A strong refinement of crystals was observed in castings made with the contribution of the recycled material. Any significant differences in castings hardness were not observed.

  1. Patterns in new dimensionless quantities containing melting temperature, and their dependence on pressure

    Directory of Open Access Journals (Sweden)

    U. WALZER

    1980-06-01

    Full Text Available The relationships existing between melting temperature and other
    macroscopic physical quantities are investigated. A new dimensionless
    quantity Q(1 not containing the Grtineisen parameter proves to be suited for serving in future studies as a tool for the determination of the melting temperature in the outer core of the Earth. The pressure dependence of more general dimensionless quantities Q„ is determined analytically and, for the chemical elements, numerically, too. The patterns of various interesting dimensionless quantities are shown in the Periodic Table and compared.

  2. Boundary-Layer Separation Control under Low-Pressure Turbine Airfoil Conditions using Glow-Discharge Plasma Actuators

    Science.gov (United States)

    Hultgren, Lennart S.; Ashpis, David E.

    2003-01-01

    Modem low-pressure turbines, in general, utilize highly loaded airfoils in an effort to improve efficiency and to lower the number of airfoils needed. Typically, the airfoil boundary layers are turbulent and fully attached at takeoff conditions, whereas a substantial fraction of the boundary layers on the airfoils may be transitional at cruise conditions due to the change of density with altitude. The strong adverse pressure gradients on the suction side of these airfoils can lead to boundary-layer separation at the latter low Reynolds number conditions. Large separation bubbles, particularly those which fail to reattach, cause a significant degradation of engine efficiency. A component efficiency drop of the order 2% may occur between takeoff and cruise conditions for large commercial transport engines and could be as large as 7% for smaller engines at higher altitude. An efficient means of of separation elimination/reduction is, therefore, crucial to improved turbine design. Because the large change in the Reynolds number from takeoff to cruise leads to a distinct change in the airfoil flow physics, a separation control strategy intended for cruise conditions will need to be carefully constructed so as to incur minimum impact/penalty at takeoff. A complicating factor, but also a potential advantage in the quest for an efficient strategy, is the intricate interplay between separation and transition for the situation at hand. Volino gives a comprehensive discussion of several recent studies on transition and separation under low-pressure-turbine conditions, among them one in the present facility. Transition may begin before or after separation, depending on the Reynolds number and other flow conditions. If the transition occurs early in the boundary layer then separation may be reduced or completely eliminated. Transition in the shear layer of a separation bubble can lead to rapid reattachment. This suggests using control mechanisms to trigger and enhance early

  3. On a free boundary problem for a strongly degenerate quasilinear parabolic equation with an application to a model of pressure filtration

    Energy Technology Data Exchange (ETDEWEB)

    Buerger, R.; Frid, H.; Karlsen, K.H.

    2002-07-01

    We consider a free boundary problem of a quasilinear strongly degenerate parabolic equation arising from a model of pressure filtration of flocculated suspensions. We provide definitions of generalized solutions of the free boundary problem in the framework of L2 divergence-measure fields. The formulation of boundary conditions is based on a Gauss-Green theorem for divergence-measure fields on bounded domains with Lipschitz deformable boundaries and avoids referring to traces of the solution. This allows to consider generalized solutions from a larger class than BV. Thus it is not necessary to derive the usual uniform estimates on spatial and time derivatives of the solutions of the corresponding regularized problem requires in the BV approach. We first prove existence and uniqueness of the solution of the regularized parabolic free boundary problem and then apply the vanishing viscosity method to prove existence of a generalized solution to the degenerate free boundary problem. (author)

  4. MDEP Technical Report TR-CSWG-03. Technical Report: fundamental attributes for the design and construction of reactor coolant pressure-boundary components

    International Nuclear Information System (INIS)

    2014-01-01

    The primary, long-term goal of MDEP's CSWG is to achieve international harmonisation of codes and standards for pressure boundary components in nuclear power plants that are important to reactor safety. The key to achieving harmonisation is to understand the extent of similarities and differences amongst the pressure boundary codes and standards used in various countries. To assist the CSWG in its long-term goals, several standards development organisations (SDOs) from various countries performed a comparison of their pressure boundary codes and standards to identify the extent of similarities and differences in code requirements and the reasons for their differences. This CSWG document provides the fundamental attributes which have been developed for the codes and standards used in the design and construction of reactor coolant pressure boundary components in nuclear power plants. The fundamental attributes are the basic concepts to be considered in the design, materials, fabrication, installation, examination, testing and over-pressure protection requirements for pressure boundary components

  5. Modeling Heat Transfer and Pressurization of Polymeric Methylene Diisocyanate (PMDI) Polyurethane Foam in a Sealed Container.

    Energy Technology Data Exchange (ETDEWEB)

    Scott, Sarah Nicole

    2018-01-01

    Polymer foam encapsulants provide mechanical, electrical, and thermal isolation in engineered systems. It can be advantageous to surround objects of interest, such as electronics, with foams in a hermetically sealed container to protect the electronics from hostile en vironments, such as a crash that produces a fire. However, i n fire environments, gas pressure from thermal decomposition of foams can cause mechanical failure of the sealed system . In this work, a detailed study of thermally decomposing polymeric methylene diisocyanate (PMDI) - polyether - polyol based polyurethane foam in a sealed container is presented . Both experimental and computational work is discussed. Three models of increasing physics fidelity are presented: No Flow, Porous Media, and Porous Media with VLE. Each model us described in detail, compared to experiment , and uncertainty quantification is performed. While the Porous Media with VLE model matches has the best agreement with experiment, it also requires the most computational resources.

  6. NPP Krsko on-line low pressure containment tightness monitoring implementation

    International Nuclear Information System (INIS)

    Dudas, M.; Basic, I.

    2004-01-01

    Containment Integrated Leak Rate Test (CILRT) 1999 in NPP Krsko was completely performed following regulation of 10CFR50 Appendix J Option A and ANSI/ANS 56.8-1987 at a design pressure (3.15 kp/cm2). In 2001 NPP Krsko proposed to Slovenian Nuclear Safety Administration (SNSA) the Technical Specification (TS) and Updated Safety Analysis Report (USAR) changes that describe implementation of new test intervals for Type A, B and C tests according to 10CFR50, Appendix J, Option B. After the positive final independent review of proposed changes by Authorized Institution, NPP Krsko received the License Amendment requiring from NPP Krsko to define technical solution for surveillance of containment tightness between two 10-years CILRT. This paper intends to discuss proposed methods by NPP Krsko, test equipment, performed measurements in 2004, associated analyses and evaluation.(author)

  7. Review of ultimate pressure capacity test of containment structure and scale model design techniques

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Moon; Choi, In Kil [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    This study was performed to obtain the basic knowledge of the scaled model test through the review of experimental studies conducted in foreign countries. The results of this study will be used for the wall segment test planed in next year. It was concluded from the previous studies that the larger the model, the greater the trust of the community in the obtained results. It is recommended that a scale model 1/4 - 1/6 be suitable considering the characteristics of concrete, reinforcement, liner and tendon. Such a large scale model test require large amounts of time and budget. Because of these reasons, it is concluded that the containment wall segment test with analytical studies is efficient for the verification of the ultimate pressure capacity of the containment structures. 57 refs., 46 figs., 11 tabs. (Author)

  8. Not a mystery. Inner containment of the pressurized water reactor (EPR trademark type)

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

    The containment of the advanced pressurized water reactor EPR trademark type is developed on the basis of the French nuclear power plant operational experience and consists of - The reinforced outer containment structure, designed to withstand external hazards (e.g. APC), - The pre-stressed inner containment structure, designed to bear the loads resulting from internal hazards (LOCA), - The steel liner, designed to provide leak tightness resulting from internal hazards. The main advantage of the pre-stressed inner containment design is that the structure remains in linear-elastic behavior during the whole life-time. Even in case of postulated design accidents (LOCA) concrete tensile strains are strongly limited. Due to pre-stressing the concrete structure remains practically free of cracks. Due to pre-stressing the leak tightness ensuring steel liner, embedded into the inner concrete shell, is exposed to more favorable compression loads. In addition to detailed calculations several test programs have been performed to verify and confirm the predicted behavior in normal operation and in accident condition. (orig.)

  9. Creep deformation and crack growth in a low alloy steel welded pressure vessel containing defects

    International Nuclear Information System (INIS)

    Coleman, M.C.

    1982-01-01

    A full-size pressure vessel was tested for effects of welding residual stresses on creep deformation and crack growth. The vessel, based on 1/2 Cr 1/2 Mo 1/4 V main steam pipe, contained four 2CrMo manual metal arc welds, two in the as-welded condition and two stress-relieved. All the welds contained pre-existing defects machined in the heat affected zones. Testing was carried out at two internal steam pressures, 250 and 350 bar, and 565 0 C. Cracked and uncracked areas of the vessel were monitored continuously. Results are presented for the continuous creep deformation observed in both the hoop and axial directions of the welds throughout the 11,400 h of testing, as well as the intermittent strain data obtained during inspections. Crack growth observations are described based on nondestructive examination. The residual stresses measured are also given for both the as-welded and stress relieved weldments. Results obtained are discussed in terms of the effects of welding residual stress on the hoop and axial deformations observed in the welds. Similarly, the effects of residual stress on creep crack growth are considered together with compositional and microstructural implications. 9 figures, 5 tables

  10. State of the art review of pressure liquefied gas container failure modes and associated projectile hazards

    Energy Technology Data Exchange (ETDEWEB)

    Leslie, I.R.M.; Birk, A.M.

    1989-08-01

    A study was carried out to investigate the state of knowledge about the failure of pressure liquified gas transport and storage tanks. A comprehensive literature search and review was carried out to assess the level of knowledge relating to the causes and characteristics of vessel ruptures. Specific parameters of interest were: the effect of vessel initial conditions (fill level, initial temperature, etc.) on rupture severity; the ability to predict the occurrence of boiling liquid expanding vapor explosions (BLEVE); and the effects of explosions such as blast waves and missile generation. The review revealed that there are several areas where knowledge is weak. These areas include: the effects of blast on structures, the prediction of hazards from, and size of, fireballs, and the understanding of failure modes of pressure liquified gas containers. It was concluded that an experimental program should be initiated to investigate the effects of container size, shape and loading conditions on the consequences of vessel rupture. 68 refs., 16 figs., 10 tabs.

  11. Computational study of sheath structure in oxygen containing plasmas at medium pressures

    Science.gov (United States)

    Hrach, Rudolf; Novak, Stanislav; Ibehej, Tomas; Hrachova, Vera

    2016-09-01

    Plasma mixtures containing active species are used in many plasma-assisted material treatment technologies. The analysis of such systems is rather difficult, as both physical and chemical processes affect plasma properties. A combination of experimental and computational approaches is the best suited, especially at higher pressures and/or in chemically active plasmas. The first part of our study of argon-oxygen mixtures was based on experimental results obtained in the positive column of DC glow discharge. The plasma was analysed by the macroscopic kinetic approach which is based on the set of chemical reactions in the discharge. The result of this model is a time evolution of the number densities of each species. In the second part of contribution the detailed analysis of processes taking place during the interaction of oxygen containing plasma with immersed substrates was performed, the results of the first model being the input parameters. The used method was the particle simulation technique applied to multicomponent plasma. The sheath structure and fluxes of charged particles to substrates were analysed in the dependence on plasma pressure, plasma composition and surface geometry.

  12. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    International Nuclear Information System (INIS)

    Kalkahoran, Omid Noori; Ahangari, Rohollah; Shirani, Amir Saied

    2016-01-01

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results

  13. Simulation of Containment Pressurization in a Large Break-Loss of Coolant Accident Using Single-Cell and Multicell Models and CONTAIN Code

    Directory of Open Access Journals (Sweden)

    Omid Noori-Kalkhoran

    2016-10-01

    Full Text Available Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model. In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code’s results.

  14. Simulation of containment pressurization in a large break-loss of coolant accident using single-cell and multicell models and CONTAIN code

    Energy Technology Data Exchange (ETDEWEB)

    Kalkahoran, Omid Noori; Ahangari, Rohollah [Reactor Research School, Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of); Shirani, Amir Saied [Faculty of Engineering, Shahid Beheshti University, Tehran (Iran, Islamic Republic of)

    2016-10-15

    Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

  15. Behaviours of reinforced concrete containment models under thermal gradient and internal pressure

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Ohnuma, H.; Yoshioka, Y.; Okada, K.; Ueda, M.

    1979-01-01

    The provisions for design concepts in Japanese Technical Standard of Concrete Containments for Nuclear Power Plants require to take account of thermal effects into design. The provisions also propose that the thermal effects could be relieved according to the degree of crack formation and creep of concrete, and may be neglected in estimating the ultimate strength capacity in extreme environmental loading conditions. This experimental study was carried out to clarify the above provisions by investigating the crack and deformation behaviours of two identical reinforced cylindrical models with dome and basement (wall outer diameter 160 cm, and wall thickness 10 cm). One of these models was hydraulically pressurized up to failure at room temperature and the other was subjected to similar internal pressure combined with the thermal gradient of approximately 40 to 50 0 C across the wall. Initial visual cracks were recognized when the stress induced by the thermal gradient reached at about 85% of bending strength of concrete used. The thermal stress of reinforcement calculated with the methods proposed by the authors using an average flexural rigidity considering the contribution of concrete showed good agreement with test results. The method based on the fully cracked section, however, was recognized to underestimate the measured stress. These cracks considerably reduced the initial deformation caused by subsequent internal pressure. (orig.)

  16. Analysis of the behaviour of pressure and temperature of the containment of a PWR reactor, submitted to a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Silva, D.E. da; Arrieta, L.A.J.; Costa, J.R.; Camargo, C.; Santos, C.M. dos; Rochedo, E.R.R.

    1979-12-01

    The main purpose of this work is to analyse the pressure and temperature behaviour of the metalic containment of a PWR building, submitted to a postulated loss-of-coolant accident (LOCA) caused by a double-ended rupture in the main line of the primary circuit. The scope of the study was directed to verify the Final Safety Analysis Report (FSAR) results for the integrity of the metalic containment of the Angra I power plant. The highest containment pressure peak for this unit is expected for a break in the suction line of one of the main pumps of the primary coolant. Using the same input data, our results are very similar to those presented in the FSAR which shows a reasonable equivalence between the two analytical models. Using as input data the results of a previous LOCA study at CNEN, which yields to more conservative boundary conditions than those presented by the FSAR, the pressure and temperature peak values determined by our model are quite larger than those presented by the cited Safety Report. (author) [pt

  17. CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1978-04-01

    The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

  18. Structural integrity of water reactor pressure boundary components. Progress report ending 29 February 1976

    International Nuclear Information System (INIS)

    Loss, F.J.

    1976-01-01

    The report describes progress in the following areas: (a) fatigue crack propagation in reactor pressure vessel steels in an air environment, (b) dynamic fracture toughness of 1-in. (25-mm) and precracked Charpy-V bend specimens under impact loading, (c) postirradiation notch ductility and properties recovery in reactor vessel steels, (d) factors contributing to variable resistance of structural steels to radiation embrittlement, and (e) the initial program plan to investigate the phenomena of warm prestress and plastic net ligament in support of thermal shock studies

  19. Flutter Sensitivity to Boundary Layer Thickness, Structural Damping, and Static Pressure Differential for a Shuttle Tile Overlay Repair Concept

    Science.gov (United States)

    Scott, Robert C.; Bartels, Robert E.

    2009-01-01

    This paper examines the aeroelastic stability of an on-orbit installable Space Shuttle patch panel. CFD flutter solutions were obtained for thick and thin boundary layers at a free stream Mach number of 2.0 and several Mach numbers near sonic speed. The effect of structural damping on these flutter solutions was also examined, and the effect of structural nonlinearities associated with in-plane forces in the panel was considered on the worst case linear flutter solution. The results of the study indicated that adequate flutter margins exist for the panel at the Mach numbers examined. The addition of structural damping improved flutter margins as did the inclusion of nonlinear effects associated with a static pressure difference across the panel.

  20. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    International Nuclear Information System (INIS)

    Lydell, B.

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research

  1. Failure rates in Barsebaeck-1 reactor coolant pressure boundary piping. An application of a piping failure database

    Energy Technology Data Exchange (ETDEWEB)

    Lydell, B. [RSA Technologies, Vista, CA (United States)

    1999-05-01

    This report documents an application of a piping failure database to estimate the frequency of leak and rupture in reactor coolant pressure boundary piping. The study used Barsebaeck-1 as reference plant. The study tried two different approaches to piping failure rate estimation: 1) PSA-style, simple estimation using Bayesian statistics, and 2) fitting of statistical distribution to failure data. A large, validated database on piping failures (like the SKI-PIPE database) supports both approaches. In addition to documenting leak and rupture frequencies, the SKI report describes the use of piping failure data to estimate frequency of medium and large loss of coolant accidents (LOCAs). This application study was co sponsored by Barsebaeck Kraft AB and SKI Research 41 refs, figs, tabs

  2. Unit Reynolds number, Mach number and pressure gradient effects on laminar-turbulent transition in two-dimensional boundary layers

    Science.gov (United States)

    Risius, Steffen; Costantini, Marco; Koch, Stefan; Hein, Stefan; Klein, Christian

    2018-05-01

    The influence of unit Reynolds number (Re_1=17.5× 106-80× 106 {m}^{-1}), Mach number (M= 0.35-0.77) and incompressible shape factor (H_{12} = 2.50-2.66) on laminar-turbulent boundary layer transition was systematically investigated in the Cryogenic Ludwieg-Tube Göttingen (DNW-KRG). For this investigation the existing two-dimensional wind tunnel model, PaLASTra, which offers a quasi-uniform streamwise pressure gradient, was modified to reduce the size of the flow separation region at its trailing edge. The streamwise temperature distribution and the location of laminar-turbulent transition were measured by means of temperature-sensitive paint (TSP) with a higher accuracy than attained in earlier measurements. It was found that for the modified PaLASTra model the transition Reynolds number (Re_{ {tr}}) exhibits a linear dependence on the pressure gradient, characterized by H_{12}. Due to this linear relation it was possible to quantify the so-called `unit Reynolds number effect', which is an increase of Re_{ {tr}} with Re_1. By a systematic variation of M, Re_1 and H_{12} in combination with a spectral analysis of freestream disturbances, a stabilizing effect of compressibility on boundary layer transition, as predicted by linear stability theory, was detected (`Mach number effect'). Furthermore, two expressions were derived which can be used to calculate the transition Reynolds number as a function of the amplitude of total pressure fluctuations, Re_1 and H_{12}. To determine critical N-factors, the measured transition locations were correlated with amplification rates, calculated by incompressible and compressible linear stability theory. By taking into account the spectral level of total pressure fluctuations at the frequency of the most amplified Tollmien-Schlichting wave at transition location, the scatter in the determined critical N-factors was reduced. Furthermore, the receptivity coefficients dependence on incidence angle of acoustic waves was used to

  3. PREST, Pressure Temperature Transients, I Inhalation in Containment Building from LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Gaggero, G [CETIS, EURATOM C.C.R., 21020 - Ispra - Varese (Italy); Gerini, P M [CISE, Segrate, Milano (Italy); Leoni, G [AGIP Nucleare, San Donato Milanese - Milano (Italy); Van Erp, J B [EURATOM C.C.R., 21020 - Ispra - Varese (Italy)

    1969-06-01

    1 - Nature of physical problem solved: The programme is intended for the determination of pressure and temperature transient inside the containment building, following a loss-of-coolant accident due to a rupture in the primary cooling system of a nuclear power plant having water as the primary coolant. The model includes the calculation of the radiation doses incurred to the thyroid due to inhalation of radioactive iodine released outside the containment building. 2 - Method of solution: The energy equation is solved at each time step by using the Newton method. In order to determine the heat exchange with structures inside the containment building as well as with the outside atmosphere, the structures are treated in slab geometry. The resulting Fourier equations for heat conduction are solved numerically by using an implicit form to avoid stability problems. 3 - Restrictions on the complexity of the problem: max. number of internal slabs - 6; max. number of external slabs - 4; max. number of meshes in each slab - 100.

  4. Analysis of hydrodynamical pressure of cavitation flow on the boundary surface

    International Nuclear Information System (INIS)

    Volin, V.E.; Donchenko, E.G.; Chepajkin, G.A.; Lunatsi, E.D.; Chernishov, P.S.; Shvartser, A.L.

    1976-01-01

    This paper substantiates the necessity of receiving test data for creation of the methods of cavitation impact impulses on the hydraulic machines and hydraulic structures. The paper describes the methodics of experimental research of intensity of impact cavitation impulses on the elements of flowing canals at different regimes of operation; the method of determining the expected erosion in flowing canals; the method of measuring the parameters of cavitation impacts on the wall of flowing canals with the use of easily damaged varnished coverings, piezo-electric pressure transducers and amplitude and spectrum analysators. The form of a separate cavitation impact is established, the sequence of impact frequency is determined and the amplitude spectra of impacts are obtained. The analysis of test results is given

  5. PA171 Containers on a Wood Pallet with Metal Top Adapter, Air Pressure Tests During MIL-STD-1660 Tests

    National Research Council Canada - National Science Library

    2004-01-01

    ... (PM-MAS) to conduct Air Pressure Tests during MIL-STD-1660, "Design Criteria for Ammunition Unit Loads" testing on the PA171 containers on a wood pallet with metal top adapter as manufactured by Alliant Tech...

  6. Modelling of pressurized water reactor fuel, rod time dependent radial heat flow with boundary element method; Modeliranje spremenljivega radijalnega toplotnega toka tlacnovodne gorivne palice z metodo robnih elementov

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1987-07-01

    The basic principles of the boundary element method numerical treatment of the radial flow heat diffusion equation are presented. The algorithm copes the time dependent Dirichlet and Neumann boundary conditions, temperature dependent material properties and regions from different materials in thermal contact. It is verified on the several analytically obtained test cases. The developed method is used for the modelling of unsteady radial heat flow in pressurized water reactor fuel rod. (author)

  7. Desizing of Starch Containing Cotton Fabrics Using Near Atmospheric Pressure, Cold DC Plasma Treatment

    Science.gov (United States)

    Prasath, A.; Sivaram, S. S.; Vijay Anand, V. D.; Dhandapani, Saravanan

    2013-03-01

    An attempt has been made to desize the starch containing grey cotton fabrics using the DC plasma with oxygen as the gaseous medium. Process conditions of the plasma reactor were optimized in terms of distance between the plates (3.2 cm), applied voltage (600 V) and applied pressure (0.01 bar) to obtain maximum desizing efficiency. No discolouration was observed in the hot water extracts of the desized sample in presence of iodine though relatively higher solvent extractable impurities (4.53 %) were observed in the plasma desized samples compared to acid desized samples (3.38 %). Also, significant weight loss, improvements in plasma desized samples were observed than that of grey fabrics in terms of drop absorbency.

  8. Ultimate Pressure Capacity of Prestressed Concrete Containment Vessels with Steel Fibers

    Energy Technology Data Exchange (ETDEWEB)

    Hahm, Dae Gi; Choun, Young Sun; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The ultimate pressure capacity (UPC) of the prestressed concrete containment vessel (PCCV) is very important since the PCCV are final protection to prevent the massive leakage of a radioactive contaminant caused by the severe accident of nuclear power plants (NPPs). The tensile behavior of a concrete is an important factor which influence to the UPC of PCCVs. Hence, nowadays, it is interested that the application of the steel fiber to the PCCVs since that the concrete with steel fiber shows an improved performance in the tensile behavior compared to reinforced concrete (RC). In this study, we performed the UPC analysis of PCCVs with steel fibers corresponding to the different volume ratio of fibers to verify the effectiveness of steel fibers on PCCVs

  9. Pressure suppression pool hydrodynamic studies for horizontal vent exit of Indian PHWR containment

    International Nuclear Information System (INIS)

    Mohan, N.; Bajaj, S.S.; Saha, P.

    1994-01-01

    The standard Indian PHWR incorporates a pressure suppression type of containment system with a suppression pool.The design of KAPS (Kakrapar Atomic Power Station) suppression pool system adopts a modified system of downcomers having horizontal vents as compared to vertical vents of NAPS (Narora Atomic Power Station). Hydrodynamic studies for vertical vents have been reported earlier. This paper presents hydrodynamic studies for horizontal type vent system during LOCA. These studies include the phenomenon of vent clearing (where the water slug standing in downcomer initially is injected to wetwell due to rapid pressurization of drywell) followed by pool swell (elevation of pool water due to formation of bubbles due to air mass entering pool at the exit of horizontal vents from drywell). The analysis performed for vent clearing and pool swell is based on rigorous thermal hydraulic calculation consisting of conservation of air-steam mixture mass, momentum and thermal energy and mass of air. Horizontal vent of downcomer is modelled in such a way that during steam-air flow, variation of flow area due to oscillating water surface in downcomer could be considered. Calculation predicts that the vent gets cleared in about 1.0 second and the corresponding downward slug velocity in the downcomer is 4.61 m/sec. The maximum pool swell for a conservative lateral expansion is calculated to be 0.56 m. (author). 3 refs., 12 figs

  10. Vapour pressures and osmotic coefficients of binary mixtures containing alcohol and pyrrolidinium-based ionic liquids

    International Nuclear Information System (INIS)

    Calvar, N.; Domínguez, Á.; Macedo, E.A.

    2013-01-01

    Highlights: • Osmotic coefficients of alcohols with pyrrolidinium ILs are determined. • Experimental data were correlated with extended Pitzer model of Archer and MNRTL. • Mean molal activity coefficients and excess Gibbs free energies were calculated. • The results have been interpreted in terms of interactions. -- Abstract: The osmotic and activity coefficients and vapour pressures of mixtures containing primary (1-propanol, 1-butanol and 1-pentanol) and secondary (2-propanol and 2-butanol) alcohols with pyrrolidinium-based ionic liquids (1-butyl-1-methyl pyrrolidinium bis(trifluoromethylsulfonyl)imide, C 4 MpyrNTf 2 , and 1-butyl-1-methyl pyrrolidinium trifluoromethanesulfonate, C 4 MpyrTFO) have been experimentally determined at T = 323.15 K. For the experimental measurements, the vapour pressure osmometry technique has been used. The results on the influence of the structure of the alcohol and of the anion of the ionic liquid on the determined properties have been discussed and compared with literature data. For the correlation of the osmotic coefficients obtained, the Extended Pitzer model of Archer and the Modified Non-Random Two Liquids model were applied. The mean molal activity coefficients and the excess Gibbs energy for the studied mixtures were calculated from the parameters obtained in the correlation

  11. Damage evaluation of 500 MWe Indian pressurized heavy water reactor nuclear containment for air craft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh; Singh, R.K; Vaze, K.K; Kushwaha, H.S.

    2003-01-01

    Non-linear transient dynamic analysis of 500 MWe Indian Pressurized Heavy Water Reactor (PHWR) nuclear containment has been carried out for the impact of Boeing and Airbus category of aircraft operated in India. The impulsive load time history is generated based on the momentum transfer of the crushable aircraft (soft missiles) of Boeing and Airbus families on the containment structure. The case studies include the analyses of outer containment wall (OCW) single model and the combined model with outer and inner containment wall (ICW) for impulsive loading due to aircraft impact. Initially the load is applied on OCW single model and subsequently the load is transferred to ICW after the local perforation of the OCW is noticed in the transient simulation. In the first stage of the analysis it is demonstrated that the OCW would suffer local perforation with a peak local deformation of 117 mm for impact due to B707-320 and 196 mm due to impact of A300B4 without loss of the overall integrity. However, this first barrier (OCW) cannot absorb the full impulsive load. In the second stage of the analysis of the combined model, the ICW is subjected to lower impulse duration as the load is transferred after 0.19 sec for B707-320 and 0.24 sec for A300B4 due to the local perforation of OCW. This results in the local deformation of approx. 115 mm for B707-320 and 124 mm for A300B4 in ICW and together both the structures (OCW and ICW) are capable of absorbing the full impulsive load. The analysis methodology evolved in the present work would be useful for studying the behaviour of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due commercial aircraft operated in India. (author)

  12. In-core assembly configuration having a dual-wall pressure boundary for nuclear reactor

    International Nuclear Information System (INIS)

    Todt, W.H. Sr.; Playfoot, K.C.

    1988-01-01

    This patent describes an in-core detector assembly of the type having an in-core part and an out-of-core part and having an elongated outer hollow housing tube with a wall thickness, an inner hollow calibration tube with a wall thickness and disposed concentrically within the outer tube to define an annular space therewith, and a plurality of discrete, circular, rod-like elements extending through the annular space, the improvement comprising: the elements having outer diameters and being of a number to substantially occupy the entire annular space of both the incore and out-of-core parts without significant voids between elements; each of the elements including at least an outer sheath and interior highly compacted mineral insulation for the entire length of the element; a first number of the elements also including center lead means connected to condition responsive element means in the in-core part of the length of the assembly and a second, remaining number of the elements being non-operating elements. The wall thickness of the housing tube and the wall thickness of the calibration tube, taken together with the diameter of the elements, provide a thickness dimension adequate to meet code primary pressure requirements for normal nuclear reactor in-core conditions, while the wall thickness of the calibration tube alone provides a thickness dimension less than adequate to meet such requirements

  13. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hong Pyo, E-mail: hplee@kepri.re.k [Nuclear Power Laboratory, Korea Electric Power Research Institute, 103-16 Munji-Dong, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of)

    2011-02-15

    Research highlights: Finite element program with 9-node degenerated shell element was developed. The developed program was mainly forced to analyze nuclear containment building. Concrete material model is adapted Niwa and Yamada failure criteria. The performance of program developed is verified through various numerical examples. The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  14. Shell finite element of reinforced concrete for internal pressure analysis of nuclear containment building

    International Nuclear Information System (INIS)

    Lee, Hong Pyo

    2011-01-01

    Research highlights: → Finite element program with 9-node degenerated shell element was developed. → The developed program was mainly forced to analyze nuclear containment building. → Concrete material model is adapted Niwa and Yamada failure criteria. → The performance of program developed is verified through various numerical examples. → The numerical analysis results similar to the experimental data. - Abstract: This paper describes a 9-node degenerated shell finite element (FE), an analysis program developed for ultimate pressure capacity evaluation and nonlinear analysis of a nuclear containment building. The shell FE developed adopts the Reissner-Mindlin (RM) assumptions to consider the degenerated shell solidification technique and the degree of transverse shear strain occurring in the structure. The material model of the concrete determines the level of the concrete stress and strain by using the equivalent stress-equivalent strain relationship. When a crack occurs in the concrete, the material behavior is expressed through the tension stiffening model that takes adhesive stress into account and through the shear transfer mechanism and compressive strength reduction model of the crack plane. In addition, the failure envelope proposed by Niwa is adopted as the crack occurrence criteria for the compression-tension region, and the failure envelope proposed by Yamada is used for the tension-tension region. The performance of the program developed is verified through various numerical examples. The analysis based on the application of the shell FE developed from the results of verified examples produced results similar to the experiment or other analysis results.

  15. Self-contained high-pressure chambers for study on the Moessbauer effect at low temperatures

    International Nuclear Information System (INIS)

    Stepanov, G.N.

    1980-01-01

    Designs of two high-pressure chambers intended for studying the Moessbauer effect at low temperatures are described. The high-pressure chamber of the Bridgman anvil type is made of non magnetic materials and intended for operation at helium temperatures. The chamber employs a superconducting pressure gage. A sample and superconducting pressure gage are surrounded with a liquid medium of a high pressure at a room temperature. Measurements of the pressure were taken during heating the chamber in the vapours of liquid helium according to the known dependence of the lead superconducting transition temperature on pressure. The other high-pressure chamber of the piston-to-cylinder type can be used to study the Moessbauer effect at temperatures ranging from 4 to 300 K. Pressure in the chamber is measured by means of the superconducting pressure gage. The maximum pressure obtained in the chamber constitutes 25 kbar

  16. ZOCO VI - a computer code to calculate the time- and space-dependent pressure distribution in full pressure containments of water-cooled reactors

    International Nuclear Information System (INIS)

    Mansfeld, G.

    1974-12-01

    ZOCO VI is a computer code to investigate the time and space dependent pressure distribution in full pressure containment of water cooled nuclear power reactors following a loss-of-coolant accident, which is caused by the rupture of a main coolant or steam line. ZOCO VI is an improved version of the computer code ZOCO V with enlarged description of condensing events. (orig.) [de

  17. Evaluation of responses to IE Bulletin 82-02: degradation of threaded fasteners in reactor coolant pressure boundary of pressurized-water-reactor plants

    International Nuclear Information System (INIS)

    Anderson, W.; Sterner, P.

    1985-05-01

    IE Bulletin 82-02 was issued by the NRC on June 2, 1982, to notify licensees about incidents of severe degradation of threaded fasteners. The bulletin required appropriate action including submittal of information from pressurized water reactors having an operating license. Responses from 41 licensees included their recent experience with degradation of threaded fasteners in primary system components. Data from recent regular inspections of reactor coolant pressure boundary component connections of 6-in. size and larger are compiled for technical evaluation. Statistical analysis is used to determine significant factors related to frequency of leakage incidents in connections, occurrence of degradation of bolts and studs, and the need for bolt replacement. Factors examined include the age of the plant, types of components, use of lubricants and sealants, and differences between plants. The compiled data indicate that, on the average, 10% of the bolted connections show evidence of leaking during an 18-month period. Also, 80% of the connections that show evidence of leakage undergo some degradation of the bolting. Results of the analysis show a significant decrease in the occurrence of bolting degradation events as the age of the plant increases. The data also show that valves are less subject to bolting corrosion. A group of 5 of the 41 plants accounted for about one-half of the reported leakage and corrosion events. The common characteristic found for four of these five plants was the lubricant used. The use of nickel-graphite based lubricants appears to offer a significantly reduced incidence of leakage and corrosion, based on late corrections to the reported data. The data also permit the conclusion that the use of molybdenum-disulfide-based lubricants and graphite-based lubricants results in a significantly increased incidence of leakage and corrosion. Reporting of data on lubricants was of poor quality and detracted from the value of the bulletin responses

  18. The plant-specific impact of different pressurization rates in the probabilistic estimation of containment failure modes

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Yang, Joon Eon; Ha, Jae Joo

    2003-01-01

    The explicit consideration of different pressurization rates in estimating the probabilities of containment failure modes has a profound effect on the confidence of containment performance evaluation that is so critical for risk assessment of nuclear power plants. Except for the sophisticated NUREG-1150 study, many of the recent containment performance analyses (through level 2 PSAs or IPE back-end analyses) did not take into account an explicit distinction between slow and fast pressurization in their analyses. A careful investigation of both approaches shows that many of the approaches adopted in the recent containment performance analyses exactly correspond to the NUREG-1150 approach for the prediction of containment failure mode probabilities in the presence of fast pressurization. As a result, it was expected that the existing containment performance analysis results would be subjected to greater or less conservatism in light of the ultimate failure mode of the containment. The main purpose of this paper is to assess potential conservatism of a plant-specific containment performance analysis result in light of containment failure mode probabilities

  19. Storage of hydrogen in advanced high pressure container. Appendices; Lagring af brint i avancerede hoejtryksbeholdere. Appendiks 1

    Energy Technology Data Exchange (ETDEWEB)

    Bentzen, J.J.; Lystrup, A. [Forskningscenter Risoe, Roskilde (Denmark)

    2005-07-15

    The objective of the project has been to study barriers for a production of advanced high pressure containers especially suitable for hydrogen, in order to create a basis for a container production in Denmark. The project has primarily focused on future Danish need for hydrogen storage in the MWh area. One task has been to examine requirement specifications for pressure tanks that can be expected in connection with these stores. Six potential storage needs have been identified: (1) Buffer in connection with start-up/regulation on the power grid. (2) Hydrogen and oxygen production. (3) Buffer store in connection with VEnzin vision. (4) Storage tanks on hydrogen filling stations. (5) Hydrogen for the transport sector from 1 TWh surplus power. (6) Tanker transport of hydrogen. Requirements for pressure containers for the above mentioned use have been examined. The connection between stored energy amount, pressure and volume compared to liquid hydrogen and oil has been stated in tables. As starting point for production technological considerations and economic calculations of various container concepts, an estimation of laminate thickness in glass-fibre reinforced containers with different diameters and design print has been made, for a 'pure' fibre composite container and a metal/fibre composite container respectively. (BA)

  20. Minimized Capillary End Effect During CO2 Displacement in 2-D Micromodel by Manipulating Capillary Pressure at the Outlet Boundary in Lattice Boltzmann Method

    Science.gov (United States)

    Kang, Dong Hun; Yun, Tae Sup

    2018-02-01

    We propose a new outflow boundary condition to minimize the capillary end effect for a pore-scale CO2 displacement simulation. The Rothman-Keller lattice Boltzmann method with multi-relaxation time is implemented to manipulate a nonflat wall and inflow-outflow boundaries with physically acceptable fluid properties in 2-D microfluidic chip domain. Introducing a mean capillary pressure acting at CO2-water interface to the nonwetting fluid at the outlet effectively prevents CO2 injection pressure from suddenly dropping upon CO2 breakthrough such that the continuous CO2 invasion and the increase of CO2 saturation are allowed. This phenomenon becomes most pronounced at capillary number of logCa = -5.5, while capillary fingering and massive displacement of CO2 prevail at low and high capillary numbers, respectively. Simulations with different domain length in homogeneous and heterogeneous domains reveal that capillary pressure and CO2 saturation near the inlet are reproducible compared with those with a proposed boundary condition. The residual CO2 saturation uniquely follows the increasing tendency with increasing capillary number, corroborated by experimental evidences. The determination of the mean capillary pressure and its sensitivity are also discussed. The proposed boundary condition is commonly applicable to other pore-scale simulations to accurately capture the spatial distribution of nonwetting fluid and corresponding displacement ratio.

  1. Method of making Tl-Sr-Ca-Cu-oxide superconductors comprising heating at elevated pressures in a sealed container

    International Nuclear Information System (INIS)

    Lechtev, W.L.; Osofsky, M.S.; Skelton, E.F.; Toth, L.E.

    1992-01-01

    This patent describes a method of forming a Tl-Sr-Ca-Cu-oxide high T c superconductor. It comprises forming a reaction mixture of the oxides of Sr, Cu, Ca, and Tl in stoichiometric proportions to make a Tl-Sr-Ca-Cu-oxide high T c superconducting compound; compressing the reaction mixture into a hard body; placing the hard body into a container for containing thallium vapor; evacuating and sealing the hard body in the container; heating the hard body and the container at a temperature of about 800 degrees C to about 950 degrees C and under pressure of at least about 30,000 psi until the container metal around the hard body and the oxides of Tl, Sr, Ca, and Cu react to form a superconducting compound; and cooling the superconducting compound to room temperature and returning the superconducting compound to atmospheric pressure

  2. Thermo-hydraulic consequence of pressure suppression containment vessel during blowdown, 2

    International Nuclear Information System (INIS)

    Aya, Izuo; Nariai, Hideki; Kobayashi, Michiyuki

    1980-01-01

    As a part of the safety research works for the integral-type marine reactor, an analytical code SUPPAC-2V was developed to simulate the thermo-hydraulic consequence of a pressure suppression containment system during blowdown and the code was applied to the Model Experimental Facility of the Safety of Integral Type Marine Reactors (explained already in Part 1). SUPPAC-2V is much different from existing codes in the following points. A nonhomogeneous model for the gaseous region in the drywell, a new correlation for condensing heat transfer coefficient at drywell wall based on existing data and approximation of air bubbles in wetwell water by one dimensional bubble rising model are adopted in this code. In comparing calculational results with experimental results, values of predominant input parameters were evaluated and discussed. Moreover, the new code was applied also to the NSR-7 marine reactor, conceptually designed at the Shipbuilding Research Association in Japan, of which suppression system had been already analysed by CONTEMPT-PS. (author)

  3. Containment venting sliding pressure venting process for PWR and BWR plants

    International Nuclear Information System (INIS)

    Eckardt, B.

    1991-01-01

    In order to reduce the residual risk associated with hypothetical severe nuclear accidents, nuclear power plants in Germany as well as in certain other European countries have been or will be backfitted with a system for filtered containment venting. During venting system process design, particular importance is attached to the requirements regarding, for example, high aerosol loading capability, provision for decay heat removal from the scrubber unit, the aerosol spectrum to be retained and entirely passive functioning of the scrubber unit. The aerosol spectrum relevant for process design and testing varies depending on aerosol concentrations, the time at which venting is commenced and whether there is an upstream wetwell, etc. Because of this the Reactor Safety Commission in Germany has specified that SnO 2 with a mass mean diameter of approximately 0.5 μm should be used as an enveloping test aerosol. To meet the above-mentioned requirements, a combined venturi scrubber system was developed which comprises a venturi section and a filter demister section and is operated in the sliding pressure mode. This scrubber system was tested using a full-scale model and has now been installed in 14 PWR and BWR plants in Germany and Finland

  4. Stress concentration factors for an internally pressurized circular vessel containing a radial U-notch

    International Nuclear Information System (INIS)

    Carvalho, E.A. de

    2005-01-01

    This paper evaluates the stress concentration factors for an internally pressurized cylinder containing a radial U-notch along its length. This work studies the cases where the external to internal radius ratio (Ψ) is equal to 1.26, 1.52, 2.00, and 3.00 and the notch radius to internal radius ratio (Φ) is fixed and equal to 0.026. The U-notch depth varies from 0.1 to 0.6 of the wall thickness. Results are also presented for a fixed size semi-circular notch. Hoop stresses at the external wall are presented, showing regions where the stress matches the nominal one and the favourable places to install strain sensors. The finite element method is used to determine the stress concentration factors (K t ) for the above described situations and for a special case where a varying semi-circular notch is present with Ψ=3.00. This notch depth varies from 0.013 to 0.3 of the wall thickness. It is pointed out that even relatively small notches introduce large stress concentrations and disrupt the hoop stress distribution all over the cross section. Results are also compared to an example found in the literature for semi-circular notches and K t curves for both cases present the same shape

  5. Burning rates of hydrogen-air mixtures in containment buildings and the consequent pressure transients

    International Nuclear Information System (INIS)

    Tennankore, K.N.; Kumar, R.K.; Razzaghi, M.

    1987-01-01

    One-dimensional flame models are often used to predict the pressure transients caused by hydrogen combustion in containments during postulated severe accidents. In the absence of data, these models account for prevailing flame acceleration mechanisms, such as initial turbulence, venting and obstacle-induced turbulence, by using arbitrarily large burning velocities that are much higher than laminar burning velocities. Using an intermediate-scale test facility at the Whiteshell Nuclear Research Establishment we have obtained necessary data on the effects of flame acceleration mechanisms, to estimate the safety margin in the buring velocities used in the models. So far, data have been analyzed, with a one-dimensional model, to determine effective burning velocities and burning-rate enhancement factors. The results of the analyses indicate that the effect of initial turbulence on the burning rate can be bounded only if the effect of flame-generated turbulence is included. The effect of venting can be accounted for by using two burning velocities, one for the pre-vent duration and a second increased value during the vented-combustion stage. The enhancement factors due to these two mechanisms, for the different conditions analyzed, varied up to 5.4, and the effective burning velocities varied up to 8.4 m/s

  6. GENERAL RULES OF SIC FORMATION IN DIAMOND-CONTAINING COMPOSITION AT LOW PRESSURE

    Directory of Open Access Journals (Sweden)

    A. E. Zhuk

    2007-01-01

    Full Text Available Results of experimental investigations of structure-formation process of «diamond-carbide silicon» composite at low pressure which is obtained by liquid silicon impregnation of a porous blank made of diamond crystals with nano-coatings have made it possible to establish the following general rules of the process concerning a sintering reaction in the coating and composite material: vacuum magnetronic spraying of composite cathodes leads to formation of nano-coating which is made of silicon and hydrogen atoms or clusters, and their subsequent treatment with plasma of glow discharge is accompanied by formation of α-SiC at low temperatures in a hard phase; silicon impregnation at 1500 °C with given pyrolytic carbon in the charge may result in β-SiC matrix formation.The formed «diamond-carbide silicon» composite material contains a frame structure of diamond crystals with nano-coating impregnated by silicon carbide and is characterized by high physical and mechanical properties. 

  7. Reliability of an HTR-module primary circuit pressure boundary. Influences, sensitivity, and comparison with a PWR

    International Nuclear Information System (INIS)

    Staat, M.

    1995-01-01

    The reliability of the HTR-module for electricity and steam generation was analysed for normal operation, as well as accident conditions. The probabilistic fracture mechanics assessment was performed with a modification of the ZERBERUS code on the basis of widely used data. The calculated failure probabilities may thus be compared with similar investigations. The HTR-module primary circuit pressure boundary as a unit showed leak-before-break behaviour in a probabilistic sense, although a break was more probable than a leak for some of its parts.However, the findings may depend greatly on the stochastic data. Therefore a stochastic reference problem is defined and the results are compared with the Japanese round robin on a PWR section. Possible changes of failure probabilities and of the leak-before-break behaviour are discussed for different criteria for the events leading to a leak, and for modifications of the stochastic reference problem such as the inclusion of NDE. The results may be used to identify those stochastic variables which have the greatest influence on the computed failure probabilities, and to perhaps justify further work which would provide more detailed information on these probabilities. Furthermore, there is an obvious need for reduction of the non-statistical reasons for great variations of failure probabilities. (orig.)

  8. Evaluation of design, leak monitoring, dnd NDEA strategies to assure PBMR Helium pressure boundary reliability - HTR2008-58037

    International Nuclear Information System (INIS)

    Fleming, K. N.; Smit, K.

    2008-01-01

    This paper discusses the reliability and integrity management (RIM) strategies that have been applied in the design of the PBMR passive metallic components for the helium pressure boundary (HPB) to meet reliability targets and to evaluate what combination of strategies are needed to meet the targets. The strategies considered include deterministic design strategies to reduce or eliminate the potential for specific damage mechanisms, use of an on-line leak monitoring system and associated design provisions that provide a high degree of leak detection reliability, and periodic nondestructive examinations combined with repair and replacement strategies to reduce the probability that degradation would lead to pipe ruptures. The PBMR RIM program for passive metallic piping components uses a leak-before-break philosophy. A Markov model developed for use in LWR risk-informed in-service inspection evaluations was applied to investigate the impact of alternative RIM strategies and plant age assumptions on the pipe rupture frequencies as a function of rupture size. Some key results of this investigation are presented in this paper. (authors)

  9. Measurement strategy of the water leakage into a low pressure sodium boundary for a liquid metal reactor

    International Nuclear Information System (INIS)

    Hur, S.; Kim, D.H.; Seong, S.H.; Kim, S.O.

    2004-01-01

    This paper deals with the measurement strategy of a water leakage into sodium boundary for a liquid metal reactor. There are several methods including the chemical sensing method, pressure sensing methods, and non-destructive method including the acoustic monitoring technique to measure the leakage. As for the results of the analysis with respect to the event propagation characteristics, it has been recommended that the acoustic method has a capability to detect small and intermediate leaks within required response time of 10 seconds. A leak of one gram/sec could be currently detected within the required response time of 10 seconds with a high reliability. In the case of less than a one gram/sec leakage, the response time could not meet our requirements due to a complicated signal processing logic. Thus, the system configuration for a fast processing of a leak detection has been recommended. It is expected that this configuration of the leak detection system could reduce the response time due to the distributed and parallel processing scheme. (orig.)

  10. Reduction of PWR containment pressure after hypothetical accidents by water-cooling of the outer containment surface - annular space spray system

    International Nuclear Information System (INIS)

    Cremer, J.; Dietrich, D.P.; Roedder, P.

    1980-12-01

    The consequences of a core melt-out accident in the vicinity of a nuclear power station are determined by the integrity of the safety containment. This can be adversely affected by different events during the course of the core melt-out accident. The most important phenomenon is the contact between the melt and sump water. Due to the evaporation of the sump water, there is a continuous rise in pressure of the safety containment, which finally leads to failure due to excess pressure. In order to reduce the fission product release due to the resulting leakage, one must try to reduce the pressure as quickly as possible. As heat cannot be removed from the steel containment to the environment because of the thick concrete containment, it is best to bypass the insulating effect of the concrete by cooling the steel containment from outside. The aim of this investigation is therefore to work out a technically relatively simple system, which offers the possibility of backfitting, setting to work and repair. Such a system is an annular space spray system, by which the annular space between the concrete and steel containment has water pumped to the level of the dome and evenly sprayed over the top hemisphere. Mobile pumps on fire engines belonging to the fire brigade are sufficient to supply the cooling water and these will be available some hours after the accident occurs. The used spray water without any radioactive components is collected outside the reactor building and/or drained off. (orig./GL) [de

  11. Loads on EPR containment after RPV failure at high pressure; Belastungen des EPR-Containments in Falle eines RDB-Versagens bei hohem Druck

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, G.

    1995-08-01

    As regards the desgin of the EPR, the general strategy is to eliminate, the vessel failure at high pressure by preventive and mitigative measures. The design proposals involved trust in the reliability of dedicated devices (relief valves) for rapid depressurization. The aim is to attain a lower pressure level at the moment of vessel failure, so that the containment is capable to cope with the blowdown impact on the pit walls and the vessel supporting structures. Nevertheless, the potential of a high-pressure failure of the vessel must be kept in mind, whatever well thought-out and reliable preventive depressurization measures might be. Therefore, the reactor pressure blowdown has been studied in order to quantify the ultimate containment load, which might support future design requirements. The calculations were performed with the LWR transient analysis thermal-hydraulics computer code REALAP5/MOD3. In previous analyses, the nodalization of the problem was based on the geometrical conditions of a typical German 1300 MW(e) NPP. In the present analysis a new input model has been used, which was based on the EPR conditions. (orig./HP)

  12. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hsii, Y.H.

    1976-06-01

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

  13. Technical update on pressure suppression type containments in use in U.S. light water reactor nuclear power plants

    International Nuclear Information System (INIS)

    1978-07-01

    In 1972, Dr. S. H. Hanauer (Technical Advisor to the NRC's Executive Director for Operations) wrote a memorandum that raised several questions on the viability of pressure suppression containment concepts. The concerns raised by Dr. Hanauer have recently become the subject of considerable discussion by several members of the U.S. Congress and public. The report provides a response to these expressed concerns and a status summary for various technical matters that relate to the safety of pressure suppression type containments for light water cooled reactor plants

  14. Fructose containing sugars do not raise blood pressure or uric acid at normal levels of human consumption.

    Science.gov (United States)

    Angelopoulos, Theodore J; Lowndes, Joshua; Sinnett, Stephanie; Rippe, James M

    2015-02-01

    The impact of fructose, commonly consumed with sugars by humans, on blood pressure and uric acid has yet to be defined. A total of 267 weight-stable participants drank sugar-sweetened milk every day for 10 weeks as part of their usual, mixed-nutrient diet. Groups 1 and 2 had 9% estimated caloric intake from fructose or glucose, respectively, added to milk. Groups 3 and 4 had 18% of estimated caloric intake from high fructose corn syrup or sucrose, respectively, added to the milk. Blood pressure and uric acid were determined prior to and after the 10-week intervention. There was no effect of sugar type on either blood pressure or uric acid (interaction P>.05), and a significant time effect for blood pressure was noted (Pfructose at the 50th percentile level, whether consumed as pure fructose or with fructose-glucose-containing sugars, does not promote hyperuricemia or increase blood pressure. © 2014 Wiley Periodicals, Inc.

  15. Standard practice for acoustic emission examination of pressurized containers made of fiberglass reinforced plastic with balsa wood cores

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2007-01-01

    1.1 This practice covers guidelines for acoustic emission (AE) examinations of pressurized containers made of fiberglass reinforced plastic (FRP) with balsa cores. Containers of this type are commonly used on tank trailers for the transport of hazardous chemicals. 1.2 This practice is limited to cylindrical shape containers, 0.5 m [20 in.] to 3 m [120 in.] in diameter, of sandwich construction with balsa wood core and over 30 % glass (by weight) FRP skins. Reinforcing material may be mat, roving, cloth, unidirectional layers, or a combination thereof. There is no restriction with regard to fabrication technique or method of design. 1.3 This practice is limited to containers that are designed for less than 0.520 MPa [75.4 psi] (gage) above static pressure head due to contents. 1.4 This practice does not specify a time interval between examinations for re-qualification of a pressure container. 1.5 This practice is used to determine if a container is suitable for service or if follow-up NDT is needed before that...

  16. Crack growth behaviour of low alloy steels for pressure boundary components under transient light water reactor operating conditions (CASTOC)

    International Nuclear Information System (INIS)

    Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ernestova, M.; Zamboch, M.; Seifert, H.P.; Ritter, S.; Roth, A.; Devrient, B.; Ehrnsten, U.

    2004-01-01

    The CASTOC project addresses environmentally assisted cracking (EAC) phenomena in low alloy steels used for pressure boundary components in both Western type boiling water reactors (BWR) and Russian type pressurised water reactors (VVER). It comprises the four work packages (WP): inter-laboratory comparison test (WP1); EAC behaviour under static load (WP2), EAC behaviour under cyclic load and load transients (WP3); evaluation of the results with regard to their relevance for components in practice (WP4). The use of sophisticated test facilities and measurement techniques for the on-line detection of crack advances have provided a more detailed understanding of the mechanisms of environmentally assisted cracking and provided quantitative data of crack growth rates as a function of loading events and time, respectively. The effect of several major parameters controlling EAC was investigated with particular emphasis on the transferability of the results to components in service. The obtained crack growth rate data were reflected on literature data and on commonly applied prediction curves as presented in the appropriate Code. At relevant stress intensity factors it could be shown that immediate cessation of growing cracks occurs after changing from cyclic to static load in high purity oxygenated BWR water and oxygen-free VVER water corresponding to steady state operation conditions. Susceptibility to environmentally assisted cracking under static load was observed for a heat affected zone material in oxygenated high purity water and also in base materials during a chloride transient representing BWR water condition below Action Level 1 of the EPRI Water Chemistry Guidelines according to the lectrical conductivity of the water but in the range of Action Level 2 according to the content of chlorides. Time based crack growth was also observed in one Russian type base material in oxygenated VVER water and in one Western type base material in oxygenated high purity BWR

  17. Framing the impact of external magnetic field on bioconvection of a nanofluid flow containing gyrotactic microorganisms with convective boundary conditions

    Directory of Open Access Journals (Sweden)

    Tanmoy Chakraborty

    2018-03-01

    Full Text Available The intention of this study is to examine the combined impacts of magnetic field and convective boundary state on bioconvection of a nanofluid flow along an expanding sheet co-existed with gyrotactic microorganisms. The fundamental partial differential equations are reduced to a set of nonlinear ordinary differential equations taking a guide of some appropriate similarity transformations. The numerical fallouts are calculated by considering the bvp4c function of Matlab. The impacts of magnetic field parameter, surface convection parameter, Eckert number and Peclet number on non-dimensional velocity, nanoparticle concentration, temperature and density of self-moving microorganisms are interpreted through graphs and charts. The fluid velocity near the surface and the Nusselt number is lessen with magnetic field. Surface convection parameter enhances the self-moving microorganism flux but a reverse result is noticed for Peclet number. Also, the contrast between the present results with formerly visited outcomes is in excellent harmony. Keywords: Nanofluid, Bioconvection, Gyrotactic microorganisms, Magnetic field, Convective boundary condition

  18. Reactor pressure boundary material - A study on evaluation and improvement of non-radiational characteristics low alloy steel weld

    Energy Technology Data Exchange (ETDEWEB)

    Lee, C. H.; Kim, K. S.; Ahn, H. J. [Hanyang University, Seoul (Korea)

    2000-04-01

    Metallurgical factors influencing on the toughness of the Intercritically Reheated Coarse-Grained Heat Affected Zone (ICCGHAZ) of multiple welded SA508-c1.3 Reactor Pressure vessel steel was evaluated. The ICCGAZ is formed when the CGHZ is reheated to the intercritical temperature range(between A{sub C1} and A{sub C3} temperature) in which {gamma} phase and {alpha} phase coexist. During heating, austenite was formed along the prior austenite grain boundaries and lath interface. On cooling, the newly formed austenite was transformed into bainite and/or martensite. The newly formed martensite always included some retained austenite(M-A constituents). The characteristics(amount, hardness, density, and size) of M-A constituents were found to be strongly associated with the last pass peak temperature and cooling time ({delta} t{sub 8/592)}). Toughness in the ICCGHAZ was deteriorated with increasing amount of M-A constituents of with was increased with increasing the last peak temperature within the intercritical temperature range. Meanwhile, for the same intercritical peak temperature, toughness decreased with increasing cooling time. When cooling time was short, the dominant factor influencing on the toughness of the ICCGHAZ was amount of M-A constituents. However, when the cooling time was long, the dominant factor was found to be the hardness difference between M-A constituents and softened matrix (tempered martensite). When the restraint was applied, restraint didn't affect on transformation temperature of martensite. But the transformation in austenite to banitic ferrite was found to be greatly affected. This austenited {gamma}-bainitic phase transformation was lowered hardness but raised toughness. Slightly, especially, when the cooling rate was relatively fast, toughness was greatly improved, but a longer cooling time did not change the characteristics considerably. 23 refs., 81 figs., 7 tabs. (Author)

  19. Assessment of Mechanisms Impacting N-Nitrosodimethylamine Fate Within the North Boundary Containment System, Rocky Mountain Arsenal

    National Research Council Canada - National Science Library

    Gunnison, Douglas

    1997-01-01

    Rocky Mountain Arsenal (RMA) was for many years a site of military chemical weapons manufacturing activities, including manufacture and assembly of weapons containing intermediate and toxic chemical end-products, incendiary...

  20. 16 CFR 1500.45 - Method for determining extremely flammable and flammable contents of self-pressurized containers.

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 2 2010-01-01 2010-01-01 false Method for determining extremely flammable and flammable contents of self-pressurized containers. 1500.45 Section 1500.45 Commercial Practices CONSUMER PRODUCT SAFETY COMMISSION FEDERAL HAZARDOUS SUBSTANCES ACT REGULATIONS HAZARDOUS SUBSTANCES AND...

  1. Excessive leakage measurement using pressure decay method in containment building local leakage rate test at nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Kyu; Kim, Chang Soo; Kim, Wang Bae [KHNP, Central Research Institute, Daejeon (Korea, Republic of)

    2016-06-15

    There are two methods for conducting the containment local leakage rate test (LLRT) in nuclear power plants: the make-up flow rate method and the pressure decay method. The make-up flow rate method is applied first in most power plants. In this method, the leakage rate is measured by checking the flow rate of the make-up flow. However, when it is difficult to maintain the test pressure because of excessive leakage, the pressure decay method can be used as a complementary method, as the leakage rates at pressures lower than normal can be measured using this method. We studied the method of measuring over leakage using the pressure decay method for conducting the LLRT for the containment building at a nuclear power plant. We performed experiments under conditions similar to those during an LLRT conducted on-site. We measured the characteristics of the leakage rate under varies pressure decay conditions, and calculated the compensation ratio based on these data.

  2. Significance of grain boundaries and stacking faults on hydrogen storage properties of Mg2Ni intermetallics processed by high-pressure torsion

    International Nuclear Information System (INIS)

    Hongo, Toshifumi; Edalati, Kaveh; Arita, Makoto; Matsuda, Junko; Akiba, Etsuo; Horita, Zenji

    2015-01-01

    Mg 2 Ni intermetallics are processed using three different routes to produce three different microstructural features: annealing at high temperature for coarse grain formation, severe plastic deformation through high-pressure torsion (HPT) for nanograin formation, and HPT processing followed by annealing for the introduction of stacking faults. It is found that both grain boundaries and stacking faults are significantly effective to activate the Mg 2 Ni intermetallics for hydrogen storage at 423 K (150 °C). The hydrogenation kinetics is also considerably enhanced by the introduction of large fractions of grain boundaries and stacking faults while the hydrogenation thermodynamics remains unchanged. This study shows that, similar to grain boundaries and cracks, stacking faults can act as quick pathways for the transportation of hydrogen in the hydrogen storage materials

  3. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.F.

    1976-01-01

    For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1 : 64 compared to the 1200 MW PWR plant Biblis A. (Auth.)

  4. Investigation of the fire at the Uranium Enrichment Laboratory. Analysis of samples and pressurization experiment/analysis of container

    International Nuclear Information System (INIS)

    Akabori, Mitsuo; Minato, Kazuo; Watanabe, Kazuo

    1998-05-01

    To investigate the cause of the fire at the Uranium Enrichment Laboratory of the Tokai Research Establishment on November 20, 1997, samples of uranium metal waste and scattered residues were analyzed. At the same time the container lid that had been blown off was closely inspected, and the pressurization effects of the container were tested and analyzed. It was found that 1) the uranium metal waste mainly consisted of uranium metal, carbides and oxides, whose relative amounts were dependent on the particle size, 2) the uranium metal waste hydrolyzed to produce combustible gases such as methane and hydrogen, and 3) the lid of the outer container could be blown off by an explosive rise of the inner pressure caused by combustion of inflammable gas mixture. (author)

  5. An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

    International Nuclear Information System (INIS)

    1982-04-01

    An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

  6. Contempt-LT: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Wheat, L.L.; Wagner, R.J.; Niederauer, G.F.; Obenchain, C.F.

    1975-06-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air-vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different. CONTEMPT-LT can be used to model all current boiling water reactor pressure suppression systems, including containments with either vertical or horizontal vent systems. CONTEMPT-LT can also be used to model pressurized water reactor dry containments, subatmospheric containments, and dual volume containments with an annulus region, and can be used to describe containment responses in experimental containment systems. The program user defines which compartments are used, specifies input mass and energy additions, defines heat structure and leakage systems, and describes the time advancement and output control. CONTEMPT-LT source decks are available in double precision extended-binary-coded-decimal-interchange-code (EBCDIC) versions. Sample problems have been run on the IBM360/75 computer. (U.S.)

  7. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 1

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherlands and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  8. Proceedings of the international specialist meeting on BWR-pressure suppression containment technology. Vol. 2

    International Nuclear Information System (INIS)

    Schultheiss, G.F.

    1981-01-01

    In the frame of R + D-work for BWR-pressure suppression systems the GKSS-Forschungszentrum Geesthacht GmbH organized an international specialist meeting. All important safety relevant aspects of pressure suppression system technology have been included. About 60 experts from USA, Japan, Sweden, Italy, Netherland and the Federal Republic of Germany participated. They came from licensing authorities, vendors, research centers and universities. In 24 papers they have shown the world-wide present status of theoretical and experimental know-how on pressure suppression system behaviour. In discussions and working groups recommendations for future work have been compiled. (orig.) [de

  9. Mark II pressure suppression containment systems: an analytical model of the pool swell phenomenon

    International Nuclear Information System (INIS)

    Ernst, R.J.; Ward, M.G.

    1976-12-01

    A one-dimensional pool swell model of the dynamic and thermodynamic conditions in the suppression chamber following a postulated loss-of-coolant accident (LOCA) is described. The pool swell phenomena is approximated by a constant thickness water slug, which is accelerated upward by the difference between the air bubble pressure acting below the pool and the wetwell air space pressure acting above the pool surface. The transient bubble pressure is computed using the known drywell pressure history and a quasi-steady compressible vent flow model. Comparisons of model predictions with pool swell experimental data are favorable and show the model is based on a conservative interpretation of the physical phenomena involved

  10. Evaluation of Ultimate Pressure Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    Energy Technology Data Exchange (ETDEWEB)

    Choun, Youngsun; Hahm, Daegi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Fiber reinforced concrete (FRC) includes thousands of small fibers that are distributed randomly in the concrete. Fibers resist the growth of cracks in concrete through their bridging at the cracks. Therefore, FRC fails in tension only when the fibers break or are pulled out of the cement matrix. For this reason, the addition of fibers in concrete mixing increases the tensile toughness of concrete and enhances the post-cracking behavior. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity of a prestressed concrete containment building (PCCB). In this study, the effects of steel or polyamide fiber reinforcement on the ultimate pressure capacity of a PCCB are evaluated. When R-SFRC contains hooked steel fibers in a volume fraction of 1.0%, the ultimate pressure capacity of a PCCB can be improved by 17%. When R-PFRC contains polyamide fibers in a volume fraction of 1.5%, the ultimate pressure capacity of a PCCB can be enhanced by 10%. Further studies are needed to determine the strain limits acceptable for PCCBs reinforced with fibers.

  11. Evaluation of Ultimate Pressure Capacity of a Prestressed Concrete Containment Building with Steel or Polyamide Fiber Reinforcement

    International Nuclear Information System (INIS)

    Choun, Youngsun; Hahm, Daegi

    2014-01-01

    Fiber reinforced concrete (FRC) includes thousands of small fibers that are distributed randomly in the concrete. Fibers resist the growth of cracks in concrete through their bridging at the cracks. Therefore, FRC fails in tension only when the fibers break or are pulled out of the cement matrix. For this reason, the addition of fibers in concrete mixing increases the tensile toughness of concrete and enhances the post-cracking behavior. A prevention of through-wall cracks and an increase of the post-cracking ductility will improve the ultimate internal pressure capacity of a prestressed concrete containment building (PCCB). In this study, the effects of steel or polyamide fiber reinforcement on the ultimate pressure capacity of a PCCB are evaluated. When R-SFRC contains hooked steel fibers in a volume fraction of 1.0%, the ultimate pressure capacity of a PCCB can be improved by 17%. When R-PFRC contains polyamide fibers in a volume fraction of 1.5%, the ultimate pressure capacity of a PCCB can be enhanced by 10%. Further studies are needed to determine the strain limits acceptable for PCCBs reinforced with fibers

  12. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    Energy Technology Data Exchange (ETDEWEB)

    Kukreja, Mukesh [Reactor Safety Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400085 (India)]. E-mail: mrkukreja@yahoo.com

    2005-08-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India.

  13. Damage evaluation of 500 MWe Indian Pressurized Heavy Water Reactor nuclear containment for aircraft impact

    International Nuclear Information System (INIS)

    Kukreja, Mukesh

    2005-01-01

    Safety assessment of Indian nuclear containments has been carried out for aircraft impact. The loading time history for Boeing and Airbus categories of aircrafts is generated based on the principle of momentum transfer of crushable aircrafts. The case studies include the analysis of BWR Mark III containment as a benchmark problem and analyses of Pressurised Heavy Water Reactor containment (inner and outer containment) for impulsive loading due to aircraft impact. Initially, the load is applied on outer containment wall model and subsequently the load is transferred to inner containment after the local perforation of the outer containment wall is noticed in the transient simulation. The analysis methodology evolved in the present work would be useful for studying the behavior of double containment walls and multi barrier structural configurations for aircraft impact with higher energies. The present analysis illustrates that with the provision of double containments for Indian nuclear power plants, adequate reserve strength is available for the case of an extremely low probability event of missile impact generated due to the commercial aircrafts operated in India

  14. Lattice Boltzmann simulations of pressure-driven flows in microchannels using Navier–Maxwell slip boundary conditions

    KAUST Repository

    Reis, Tim; Dellar, Paul J.

    2012-01-01

    lattice Boltzmann formulations cannot capture Knudsen boundary layers, we replace the usual discrete analogs of the specular diffuse reflection conditions from continuous kinetic theory with a moment-based implementation of the first-order Navier

  15. MDEP Technical Report TR-CSWG-01. Technical Report: Regulatory Frameworks for the Use of Nuclear Pressure Boundary Codes and Standards in MDEP Countries

    International Nuclear Information System (INIS)

    2013-01-01

    The Codes and Standards Working Group (CSWG) is one of the issue-specific working groups that the MDEP members are undertaking; its long term goal is harmonisation of regulatory and code requirements for design and construction of pressure-retaining components in order to improve the effectiveness and efficiency of the regulatory design reviews, increase quality of safety assessments, and to enable each regulator to become stronger in its ability to make safety decisions. The CSWG has interacted closely with the Standards Development Organisations (SDOs) and CORDEL in code comparison and code convergence. The Code Comparison Report STP-NU-051 has been issued by SDO members to identify the extent of similarities and differences amongst the pressure-boundary codes and standards used in various countries. Besides the differences in codes and standards, the way how the codes and standards are applied to systems, structures and components also affects the design and construction of nuclear power plant. Therefore, to accomplish the goal of potential harmonisation, it is also vital that the regulators learn about each other's procedures, processes, and regulations. To facilitate the learning process, the CSWG meets regularly to discuss issues relevant to licensing new reactors and using codes and standards in licensing safety reviews. The CSWG communicates very frequently with the SDOs to discuss similarities and differences among the various codes and how to proceed with potential harmonisation. It should be noted that the IAEA is invited to all of the issue-specific working groups within MDEP to ensure consistency with IAEA standards. The primary focus of this technical report is to consolidate information shared and accomplishments achieved by the member countries. This report seeks to document how each MDEP regulator utilises national or regional mechanical codes and standards in its safety reviews and licensing of new reactors. The preparation of this report

  16. Safety conditions of using structural steels under high temperature and pressures in hydrogen containing environment

    International Nuclear Information System (INIS)

    Asviyan, M.B.

    1984-01-01

    The method for establishing full-strength conditions was suggested on the base of results of creep-rupture test of tube samples under hydrogen pressure and according to permissible stresses in neutral medium. Applicability of the method was considered taking St3 and 12KhM steels as examples. It was shown that the use of suggested dependences and special efficiency factors enables to forecast endurance limit for the given steel grade and assigned partial hydrogen pressure without labour-intensive test conducting

  17. Regulations for pressurized equipment in the European Single Market - construction of steam boilers, containers and pipelines

    International Nuclear Information System (INIS)

    Grassmuck, J.

    1992-01-01

    The impulses produced by the data of the standardized EC Single Market have now reached pressurized equipment in the field of EC Guidelines and European standardisation. This must be regarded as a great challenge to the interested and concerned parties. All efforts to represent the interested parties in European Committees must be made. In order to reach the goal quickly and successfully, a considerable readiness to compromise is, however, necessary. At the end of the development process, a comprehensible, standardized set of regulations will be available for pressurized equipment throughout Europe. The regulations will consist of national ones converted into European Guidelines and Standards. (orig.) [de

  18. Condensation phenomena in BWR-pressure suppression containments under LOCA conditions

    International Nuclear Information System (INIS)

    Aust, E.; McCauley, E.W.; Niemann, H.R.

    1983-01-01

    Experimental studies on condensation phenomena in pressure suppression systems (PSS) have shown, that chugging produces the major dynamic loads in a PSS. Time correlation of digital and visual data have produced understanding of the essential physics of this phenomenon: chugging events are characterized by pipe outside and pipe inside condensation. Pipe outside condensation is smooth, sometimes accompanied by vent pipe acoustic frequency. Pipe inside condensation is ring-like and induces a strong pressure pulse with ringdown frequency. The steam ring is caused by the retreating steam front in the pipe exit, which acts as a BORDA-mouth. (orig.) [de

  19. Process and device for reducing the pressure in the saftey containment of a nuclear reactor plant

    International Nuclear Information System (INIS)

    Stiefel, M.

    1984-01-01

    Part of the gaseous contents of the safety containment are drawn off. Hydrogen up to a maximum of 3.5% by volume is added to this gas. Part of the oxygen content of the gas is burnt with the hydrogen in the well-known way. The gas reduced in oxygen content is returned to the safety containment. The water produced in the reaction is taken back with the gas to the safety containment in the form of steam and is condensed there. (orig./HP) [de

  20. ICECON: a computer program used to calculate containment back pressure for LOCA analysis (including ice condenser plants)

    International Nuclear Information System (INIS)

    1976-07-01

    The ICECON computer code provides a method for conservatively calculating the long term back pressure transient in the containment resulting from a hypothetical Loss-of-Coolant Accident (LOCA) for PWR plants including ice condenser containment systems. The ICECON computer code was developed from the CONTEMPT/LT-022 code. A brief discussion of the salient features of a typical ice condenser containment is presented. Details of the ice condenser models are explained. The corrections and improvements made to CONTEMPT/LT-022 are included. The organization of the code, including the calculational procedure, is outlined. The user's manual, to be used in conjunction with the CONTEMPT/LT-022 user's manual, a sample problem, a time-step study (solution convergence) and a comparison of ICECON results with the results of the NSSS vendor are presented. In general, containment pressure calculated with the ICECON code agree with those calculated by the NSSS vendor using the same mass and energy release rates to the containment

  1. Nonlinear transient dynamic response of pressure relief valves for a negative containment system

    International Nuclear Information System (INIS)

    Aziz, T.S.; Duff, C.G.; Tang, J.H.K.

    1979-01-01

    The response of the piston for the postulated simultaneous effect of pressure and an earthquake is obtained for different parameters and accident conditions. Response quantities such as accelerations, displacements, rotations, diaphragm forces as well as opening time during a design basis earthquake are obtained. The results of the different analyses, as related to the functional operability of the valves, are evaluated and discussed. (orig.)

  2. Railroad Rails Containing Electrode-Induced Pitting from Pressure Electric Welding

    Science.gov (United States)

    2018-04-18

    This paper describes the forensic evaluations of three railroad rails containing electrode-induced pitting. These evaluations include: magnetic particle inspection to nondestructively detect cracks emanating from the pitting; fractography to study th...

  3. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    International Nuclear Information System (INIS)

    Horschel, D.S.

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission's program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix

  4. Experimental results from pressure testing a 1:6-scale nuclear power plant containment

    Energy Technology Data Exchange (ETDEWEB)

    Horschel, D.S. [Sandia National Labs., Albuquerque, NM (United States)

    1992-01-01

    This report discusses the testing of a 1:6-scale, reinforced-concrete containment building at Sandia National Laboratories, in Albuquerque, New Mexico. The scale-model, Light Water Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc., and was instrumented with over 1200 transducers to prepare for the test. The containment model was tested to failure to determine its response to static internal overpressurization. As part of the US Nuclear Regulatory Commission`s program on containment integrity, the test results will be used to assess the capability of analytical methods to predict the performance of containments under severe-accident loads. The scaled dimensions of the cylindrical wall and hemispherical dome were typical of a full-size containment. Other typical features included in the heavily reinforced model were equipment hatches, personnel air locks, several small piping penetrations, and a ihin steel liner that was attached to the concrete by headed studs. In addition to the transducers attached to the model, an acoustic detection system and several video and still cameras were used during testing to gather data and to aid in the conduct of the test. The model and its instrumentation are briefly discussed, and is followed by the testing procedures and measured response of the containment model. A summary discussion is included to aid in understanding the significance of the test as it applies to real world reinforced concrete containment structures. The data gathered during SIT and overpressure testing are included as an appendix.

  5. Accidental sequences associated with the containment of the pressurized water nuclear installation - INAP

    International Nuclear Information System (INIS)

    Natacci, Faustina Beatriz; Correa, Francisco

    2002-01-01

    The analysis of accidental sequences associated with the Containment is one of the most important tasks during the development of the Probabilistic Safety Assessment (PSA) of nuclear plants mainly because of its importance on the mitigation of consequences of severe postulated accident initiating events. This paper presents a first approach of the Containment analysis of the INAP identifying failures and events that can compromise its performance, and outlining accidental sequences and Containment end states. The initial plant damage states, which are the input for this study, are based on the event trees developed in the PSA level 1 for the INAP. It should be emphasized that since this PSA is still in a preliminary stage it is subjected to further completion. Consequently, the Containment analysis shall also be revised in order to incorporate, in an extension as complete as possible, all initial plant damage states, the corresponding event trees, and the related Containment end states. Finally, it can be concluded that the evaluation of the qualitative analysis presented herein allows a concise and broad knowledge of the qualitative analysis presented herein allows a concise and broad knowledge of the development of accidental sequences related to the Containment of the INAP. (author)

  6. Effect of Reynolds Number and Periodic Unsteady Wake Flow Condition on Boundary Layer Development, Separation, and Intermittency Behavior Along the Suction Surface of a Low Pressure Turbine Blade

    Science.gov (United States)

    Schobeiri, M. T.; Ozturk, B.; Ashpis, David E.

    2007-01-01

    The paper experimentally studies the effects of periodic unsteady wake flow and different Reynolds numbers on boundary layer development, separation and re-attachment along the suction surface of a low pressure turbine blade. The experimental investigations were performed on a large scale, subsonic unsteady turbine cascade research facility at Turbomachinery Performance and Flow Research Laboratory (TPFL) of Texas A&M University. The experiments were carried out at Reynolds numbers of 110,000 and 150,000 (based on suction surface length and exit velocity). One steady and two different unsteady inlet flow conditions with the corresponding passing frequencies, wake velocities, and turbulence intensities were investigated. The reduced frequencies chosen cover the operating range of LP turbines. In addition to the unsteady boundary layer measurements, surface pressure measurements were performed. The inception, onset, and the extent of the separation bubble information collected from the pressure measurements were compared with the hot wire measurements. The results presented in ensemble-averaged, and the contour plot forms help to understand the physics of the separation phenomenon under periodic unsteady wake flow and different Reynolds number. It was found that the suction surface displayed a strong separation bubble for these three different reduced frequencies. For each condition, the locations defining the separation bubble were determined carefully analyzing and examining the pressure and mean velocity profile data. The location of the boundary layer separation was dependent of the Reynolds number. It is observed that starting point of the separation bubble and the re-attachment point move further downstream by increasing Reynolds number from 110,000 to 150,000. Also, the size of the separation bubble is smaller when compared to that for Re=110,000.

  7. Material properties for reactor pressure vessels and containment shells under dynamic loading

    International Nuclear Information System (INIS)

    Albertini, C.

    1997-01-01

    The effects of high strain rate, dynamic biaxial loading and deformation mode (tension, shear) on the mechanical properties of AISI 316 austenitic stainless steel in as-received and pre-damaged (creep, LCF) conditions are reported. This research was conducted to assess the performances of the containment shell of fast breeder reactors. The results of this research have been utilized to prepare similar investigations for SA 537 Class 1 ferritic steel used for the containment shell of LWR. The first results of these investigations are reported. A programme to study the mechanical properties of plain concrete with real size aggregate at high strain rate is described. (orig.)

  8. The EPR (European Pressurized Water Reactor) containment - concept, testing of leakage behaviour, FRP liner

    Energy Technology Data Exchange (ETDEWEB)

    Touret, J.P. [EDF SEPTEN, Villeurbanne (France); Liersch, G. [Bayernwerk Kerenergie GmbH, Muenchen (Germany); Danisch, R. [Siemens AG, KWU NAD, Erlangen (Germany)

    2001-07-01

    The Basic Design of the EPR has now been completed. The containment plays a major safety-related role with respect to protection of the environment against radioactive releases. The EPR features a double (steel-reinforced concrete/prestressed concrete) containment design, with the inner containment coated additionally with a fibreglass-reinforced plastic (FRP) liner in certain areas. This means that containment leaktightness is provided mainly by the prestressed concrete and the FRP liner in the event of a postulated accident. The numerous findings of the tests carried out so far in both France and Germany are summarized. (orig.) [German] Das Basic Design fuer den EPR ist fertiggestellt. Entscheidend fuer eine Realisierung wird neben der politischen Akzeptanz vor allem die Wettbewerbsfaehigkeit mit anderen Energietraegern sein. Im EPR-Projekt wird der hohe Sicherheitsstandard der heutigen Kernkraftwerke in Deutschland und Frankreich ergaenzt, indem zusaetzlich technische Massnahmen ergriffen werden, um die Konsequenzen beim unterstellten Versagen aller sicherheitstechnischen Einrichtungen mit der Folge eines postulierten Niederschmelzen des Kerns technisch zu beherrschen. (orig.)

  9. The effects of high pressure treatments on C. jejuni in ground poultry products containing polyphosphate additives

    Science.gov (United States)

    Marinades containing polyphosphates have been previously implicated in the enhanced survival of Campylobacter spp. in poultry product exudates. The enhanced Campylobacter survival was attributed primarily to the ability of some polyphosphates to change the pH of the exudate to one more amenable to ...

  10. Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Kanzleiter, T.

    1975-10-01

    The phenomena occuring within a containment during a LOCA are currently investigated through experiments with a modelcontainment at Battelle-Institut Frankfurt on behalf of the Bundesministerium fuer Forschung und Technologie, Bonn. The experimental results are to be compared with the results of model calculations in order to improve the calculational methods. An experimental facility was built, consisting of a primary coolant circuit and a special model-containment. The model-containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m 3 and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross section. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiment a PWR-configuration with nine compartments has been istalled. The model scale of the compartment volumes and the overflow areas are about 1:64 compared to the 1,200-MW-PWR-plant Biblis A. Later investigations will also include BWR-experiments and experiments leading to an extremely high load on special containment structures. (orig.) [de

  11. Pressure-induced structural change in MgSiO3 glass at pressures near the Earth's core-mantle boundary.

    Science.gov (United States)

    Kono, Yoshio; Shibazaki, Yuki; Kenney-Benson, Curtis; Wang, Yanbin; Shen, Guoyin

    2018-02-20

    Knowledge of the structure and properties of silicate magma under extreme pressure plays an important role in understanding the nature and evolution of Earth's deep interior. Here we report the structure of MgSiO 3 glass, considered an analog of silicate melts, up to 111 GPa. The first (r1) and second (r2) neighbor distances in the pair distribution function change rapidly, with r1 increasing and r2 decreasing with pressure. At 53-62 GPa, the observed r1 and r2 distances are similar to the Si-O and Si-Si distances, respectively, of crystalline MgSiO 3 akimotoite with edge-sharing SiO 6 structural motifs. Above 62 GPa, r1 decreases, and r2 remains constant, with increasing pressure until 88 GPa. Above this pressure, r1 remains more or less constant, and r2 begins decreasing again. These observations suggest an ultrahigh-pressure structural change around 88 GPa. The structure above 88 GPa is interpreted as having the closest edge-shared SiO 6 structural motifs similar to those of the crystalline postperovskite, with densely packed oxygen atoms. The pressure of the structural change is broadly consistent with or slightly lower than that of the bridgmanite-to-postperovskite transition in crystalline MgSiO 3 These results suggest that a structural change may occur in MgSiO 3 melt under pressure conditions corresponding to the deep lower mantle.

  12. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    International Nuclear Information System (INIS)

    Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

    2006-01-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  13. Pressurized and radioactivity-containing components of systems outside the primary circuit. Pt. 1

    International Nuclear Information System (INIS)

    1990-01-01

    This document lays down the requirements on: a) Organisations involved in manufacturing; b) manufacture of materials and moulds, their chemical composition, mechanical and engineering properties, physical properties, heat treatment and further machining; c) verification procedures and controls for the achievement and maintenance of specified quality standards for materials and moulds, plus destructive and non-destructive testing; d) the provision of documents for the documentary files on test results. The following components fall under the remit of this rule: a) pressure vessel, b) piping and pipe sections/fittings, c) pumps, d) valves. (orig./HP) [de

  14. Pressurized and radioactivity-containing components of systems outside the primary circuit. Pt. 1

    International Nuclear Information System (INIS)

    1991-01-01

    This document lays down the requirements on: a) Organisations involved in manufacturing; b) Manufacture of materials and moulds, their chemical composition, mechanical and engineering properties, physical properties, heat treatment and further machining; c) Verification procedures and controls for the achievement and maintenance of specified quality standards for materials and moulds, plus destructive and non-destructive testing; d) The provision of documents for the documentary files on test results. The following components fall under the remit of this rule: a) pressure vessel, b) piping and pipe sections/fittings, c) pumps, d) valves. (orig.) [de

  15. Decontamination of objects in a sealed container by means of atmospheric pressure plasmas

    DEFF Research Database (Denmark)

    Leipold, Frank; Schultz-Jensen, Nadja; Kusano, Yukihiro

    2011-01-01

    . The ambient atmosphere was air at atmospheric pressure. A plasma is generated inside the bag forming ozone from the oxygen. The maximum ozone concentration in the bag was found to be 140 ppm. A log 6 reduction of L. innocua is obtained after 15 min of exposure time. The temperature of the slides after...... for the experiments. Glass slides were inoculated with L. innocua. The slides were placed inside a low density polyethylene (LDPE) bag. The bag was filled with a gas mixture of 97.5 Vol% Ar and 2.5 Vol% O2 and subsequently sealed. The bag was placed between the electrodes of a dielectric barrier discharge...

  16. Containment bellows testing under extreme loads

    International Nuclear Information System (INIS)

    Splezter, B.L.; Lambert, L.D.; Parks, M.B.

    1993-01-01

    Sandia National Laboratories (SNL) is conducting several research programs to help develop validated methods for the prediction of the ultimate pressure capacity, at elevated temperatures, of light water reactor (LWR) containment structures. To help understand the ultimate pressure of the entire containment pressure boundary, each component must be evaluated. The containment pressure boundary consists of the containment shell and many access, piping, and electrical penetrations. The focus of the current research program is to study the ultimate behavior of flexible metal bellows that are used at piping penetrations. Bellows are commonly used at piping penetrations in steel containments; however, they have very few applications in concrete (reinforced or prestressed) containments. The purpose of piping bellows is to provide a soft connection between the containment shell and the pipe are attached while maintaining the containment pressure boundary. In this way, piping loads caused by differential movement between the piping and the containment shell are minimized. SNL is conducting a test program to determine the leaktight capacity of containment bellows when subjected to postulated severe accident conditions. If the test results indicate that containment bellows could be a possible failure mode of the containment pressure boundary, then methods will be developed to predict the deformation, pressure, and temperature conditions that would likely cause a bellows failure. Results from the test program would be used to validate the prediction methods. This paper provides a description of the use and design of bellows in containment piping penetrations, the types of possible bellows loadings during a severe accident, and an overview of the test program, including available test results at the time of writing

  17. Free convection boundary layer flow past a horizontal flat plate embedded in porous medium filled by nano-fluid containing gyro-tactic microorganisms

    Energy Technology Data Exchange (ETDEWEB)

    Aziz, A. [Department of Mechanical Engineering, School of Engineering and Applied Science, Gonzaga University, Spokane, WA 99258 (United States); Khan, W.A. [Department of Engineering Sciences, National University of Sciences and Technology, Karachi 75350 (Pakistan); Pop, I. [Department of Applied Mathematics, Babes-Bolyai University, Cluj-Napoca (Romania)

    2012-06-15

    The steady boundary layer free convection flow past a horizontal flat plate embedded in a porous medium filled by a water-based nano-fluid containing gyro-tactic microorganisms is investigated. The Oberbeck-Boussinesq approximation is assumed in the analysis. The effects of bio-convection parameters on the dimensionless velocity, temperature, nano-particle concentration and density of motile microorganisms as well as on the local Nusselt, Sherwood and motile microorganism numbers are investigated and presented graphically. In the absence of bio-convection, the results are compared with the existing data in the open literature and found to be in good agreement. The bio-convection parameters strongly influence the heat, mass, and motile microorganism transport rates. (authors)

  18. Administrative Area Boundaries 2 (State Boundaries), Region 9, 2010, NAVTEQ

    Data.gov (United States)

    U.S. Environmental Protection Agency — NAVTEQ Administrative Area Boundaries 2 (State Boundaries) for Region 9. There are five Administrative Area Boundaries layers (1, 2, 3, 4, 5). These layers contain...

  19. Administrative Area Boundaries 4 (City Boundaries), Region 9, 2010, NAVTEQ

    Data.gov (United States)

    U.S. Environmental Protection Agency — NAVTEQ Administrative Area Boundaries 4 (City Boundaries) for Region 9. There are five Administrative Area Boundaries layers (1, 2, 3, 4, 5). These layers contain...

  20. Survey of neutrons inside the containment of a pressurized water reactor

    International Nuclear Information System (INIS)

    Hankins, D.E; Griffith, R.V.

    1978-01-01

    A neutron survey was made inside the containment of the Farley Nuclear Plant, Alabama Power and Light Company, Dothan, Alabama, in November 1977. The survey was made to determine the spectra of leakage neutrons and to evaluate the accuracy of albedo neutron dosimeters and a 9-in.-diameter sphere rem meter. The survey also covered variations in the neutron spectra, the ratio of gamma-to-neutron dose rates, and the thermal neutron component of the neutron dose

  1. Neutron dosimetry in containment of a pressurized water reactor utilizing the Panasonic UD-802 dosimetry system

    International Nuclear Information System (INIS)

    Kralick, S.C.

    1984-01-01

    The Panasonic UD-802 dosimeter was evaluated as a potential neutron dosimeter for use in containment of a PWR. The Panasonic UD-802 dosimeter, although designed as a beta and gamma dosimeter, is also sensitive to neutrons. UD-802 dosimeters were mounted on polyethylene phantoms and irradiated to known doses at selected locations in containment. The known neutron dose equivalents were determined based on remmeter dose rate measurements and stay times. The thermoluminescent response of the dosimeters and the known neutron dose equivalents were used to obtain a calibration factor at each location. The average calibration factor was 3.7 (unit of dosimeter response per mrem) and all calibration factors were within +-30% of this mean value. The dosimeter distance from the phantom was found to have minimal effect on the response but the system was directionally dependent, necessitating a correction in the calibration factor. The minimum significant dosimeter response was determined independent of any calibration factor. The minimum significant response of the UD-802 to neutrons is a function of the corresponding gamma exposure rate. It is concluded that the Panasonic UD-802 dosimeter can be used for neutron dosimetry in PWR containment

  2. A reformulated synthetic turbulence generation method for a zonal RANS–LES method and its application to zero-pressure gradient boundary layers

    International Nuclear Information System (INIS)

    Roidl, B.; Meinke, M.; Schröder, W.

    2013-01-01

    Highlights: • A synthetic turbulence generation method (STGM) is presented. • STGM is applied to sub and supersonic flows at low and moderate Reynolds numbers. • STGM shows a convincing quality in zonal RANS–LES for flat-plate boundary layers (BLs). • A good agreement with the pure LES and reference DNS findings is obtained. • RANS-to-LES transition length is reduced to less than four boundary-layer thicknesses. -- Abstract: A synthetic turbulence generation (STG) method for subsonic and supersonic flows at low and moderate Reynolds numbers to provide inflow distributions of zonal Reynolds-averaged Navier–Stokes (RANS) – large-eddy simulation (LES) methods is presented. The STG method splits the LES inflow region into three planes where a local velocity signal is decomposed from the turbulent flow properties of the upstream RANS solution. Based on the wall-normal position and the local flow Reynolds number, specific length and velocity scales with different vorticity content are imposed at the inlet plane of the boundary layer. The quality of the STG method for incompressible and compressible zero-pressure gradient boundary layers is shown by comparing the zonal RANS–LES data with pure LES, pure RANS, and direct numerical simulation (DNS) solutions. The distributions of the time and spanwise wall-shear stress, Reynolds stress distributions, and two point correlations of the zonal RANS–LES simulations are smooth in the transition region and in good agreement with the pure LES and reference DNS findings. The STG approach reduces the RANS-to-LES transition length to less than four boundary-layer thicknesses

  3. Validation of Heat Transfer Thermal Decomposition and Container Pressurization of Polyurethane Foam.

    Energy Technology Data Exchange (ETDEWEB)

    Scott, Sarah Nicole; Dodd, Amanda B.; Larsen, Marvin E.; Suo-Anttila, Jill M.; Erickson, Kenneth L

    2014-09-01

    Polymer foam encapsulants provide mechanical, electrical, and thermal isolation in engineered systems. In fire environments, gas pressure from thermal decomposition of polymers can cause mechanical failure of sealed systems. In this work, a detailed uncertainty quantification study of PMDI-based polyurethane foam is presented to assess the validity of the computational model. Both experimental measurement uncertainty and model prediction uncertainty are examined and compared. Both the mean value method and Latin hypercube sampling approach are used to propagate the uncertainty through the model. In addition to comparing computational and experimental results, the importance of each input parameter on the simulation result is also investigated. These results show that further development in the physics model of the foam and appropriate associated material testing are necessary to improve model accuracy.

  4. Flexible Pressure Sensor Based on PVDF Nanocomposites Containing Reduced Graphene Oxide-Titania Hybrid Nanolayers

    Directory of Open Access Journals (Sweden)

    Aisha Al-Saygh

    2017-01-01

    Full Text Available A novel flexible nanocomposite pressure sensor with a tensile strength of about 47 MPa is fabricated in this work. Nanolayers of titanium dioxide (titania nanolayers, TNL synthesized by hydrothermal method are used to reinforce the polyvinylidene fluoride (PVDF by simple solution mixing. A hybrid composite is prepared by incorporating the TNL (2.5 wt % with reduced graphene oxide (rGO (2.5 wt % synthesized by improved graphene oxide synthesis to form a PVDF/rGO-TNL composite. A comparison between PVDF, PVDF/rGO (5 wt %, PVDF/TNL (5 wt % and PVDF/rGO-TNL (total additives 5 wt % samples are analyzed for their sensing, thermal and dielectric characteristics. The new shape of additives (with sharp morphology, good interaction and well distributed hybrid additives in the matrix increased the sensitivity by 333.46% at 5 kPa, 200.7% at 10.7 kPa and 246.7% at 17.6 kPa compared to the individual PVDF composite of TNL, confirming its possible application in fabricating low cost and light weight pressure sensing devices and electronic devices with reduced quantity of metal oxides. Increase in the β crystallinity percentage and removal of α phase for PVDF was detected for the hybrid composite and linked to the improvement in the mechanical properties. Tensile strength for the hybrid composite (46.91 MPa was 115% higher than that of the neat polymer matrix. Improvement in the wettability and less roughness in the hybrid composites were observed, which can prevent fouling, a major disadvantage in many sensor applications.

  5. Application of Atmospheric Pressure Photoionization H/D-exchange Mass Spectrometry for Speciation of Sulfur-containing Compounds.

    Science.gov (United States)

    Acter, Thamina; Kim, Donghwi; Ahmed, Arif; Ha, Ji-Hyoung; Kim, Sunghwan

    2017-08-01

    Herein we report the observation of atmospheric pressure in-source hydrogen-deuterium exchange (HDX) of thiol group for the first time. The HDX for thiol group was optimized for positive atmospheric pressure photoionization (APPI) mass spectrometry (MS). The optimized HDX-MS was applied for 31 model compounds (thiols, thiophenes, and sulfides) to demonstrate that exchanged peaks were observed only for thiols. The optimized method has been successfully applied to the isolated fractions of sulfur-rich oil samples. The exchange of one and two thiol hydrogens with deuterium was observed in the thiol fraction; no HDX was observed in the other fractions. Thus, the results presented in this study demonstrate that the HDX-MS method using APPI ionization source can be effective for speciation of sulfur compounds. This method has the potential to be used to access corrosion problems caused by thiol-containing compounds. Graphical Abstract ᅟ.

  6. Microstructure and Properties of Cobalt-and Zinc-Containing Magnetic Magnesium Alloys Processed by High-Pressure Die Casting

    Science.gov (United States)

    Klose, Christian; Demminger, Christian; Maier, Hans Jürgen

    The inherent magnetic properties of lightweight alloys based on magnesium and cobalt offer a novel way in order to measure mechanical loads throughout the entire structural component using the magnetoelastic effect. Because the solubility of cobalt in the magnesium matrix is negligible, the magnetic properties mainly originate from Co-rich precipitates. Thus, the size and distribution of Co-containing phases within the alloy's microstructure wields a major influence on the amplitude of the load-sensitive properties which can be measured by employing the harmonic analysis of eddy-current signals. In this study, Mg-Co-based alloys are produced by several casting methods which allow the application of different cooling rates, e.g. gravity die casting and high-pressure die casting. The differences between the manufactured alloys' micro- and phase structures are compared depending on the applied cooling rate and the superior magnetic and mechanical properties of the high-pressure die cast material are demonstrated.

  7. Relation between the Fluctuating Wall Pressure and the Turbulent Structure of a Boundary Layer on a Cylinder in Axial Flow

    Science.gov (United States)

    1993-08-12

    Rlain in . power spectral density of the fluctuating wall pressure on the cylinder, boldine . fractional contribution to the total wall pressure energy...or repeated sequences of events are responsible for the production of turbulence in the near- wall region and the desire to extract their...signals over a prespecified window centered about the event detection times to extract the individual events. I 3.) Ensemble average the individual

  8. Stability of spatially developing boundary layers

    Science.gov (United States)

    Govindarajan, Rama

    1993-07-01

    A new formulation of the stability of boundary-layer flows in pressure gradients is presented, taking into account the spatial development of the flow. The formulation assumes that disturbance wavelength and eigenfunction vary downstream no more rapidly than the boundary-layer thickness, and includes all terms of O(1) and O(R(exp -1)) in the boundary-layer Reynolds number R. Although containing the Orr-Sommerfeld operator, the present approach does not yield the Orr-Sommerfeld equation in any rational limit. In Blasius flow, the present stability equation is consistent with that of Bertolotti et al. (1992) to terms of O(R(exp -1)). For the Falkner-Skan similarity solutions neutral boundaries are computed without the necessity of having to march in space. Results show that the effects of spatial growth are striking in flows subjected to adverse pressure gradients.

  9. Stresses and strains in the steel containment resulting from transient pressure and temperature loading during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gruner, P.; Kuntze, W.M.; Jansky, J.

    1985-01-01

    Posttest calculations of stresses and strains in the steel containment of the German research reactor HDR were performed for a simulated LOCA. The results of the theoretical investigations are presented and compared to experimental findings. The pressure and temperature loading of the shell was determined with the thermodynamic code COFLOW on the basis of a multi-compartment model. Using a three-dimensional finite element model the temporal behaviour of the containment was calculated employing the structural mechanics code ASKA. Global bending deformations and local negative straining of the steel shell is discussed. Theoretical and experimental results agree in most cases rather well. Reasons for deviations will be discussed. The specific behaviour of strains found in the vicinity of locally heated areas will be explained by means of analytical considerations. (orig.)

  10. Shakedown boundary determination of a 90° back-to-back pipe bend subjected to steady internal pressures and cyclic in-plane bending moments

    International Nuclear Information System (INIS)

    Abdalla, Hany F.

    2014-01-01

    No experimental data exist within open literature, to the best knowledge of the author, for determining shakedown boundaries of 90° back-to-back pipe bends. Ninety degree back-to-back pipe bends are extensively utilized within piping networks of nuclear submarines and modern turbofan aero-engines where space limitation is considered a paramount concern. In the current research, the 90° back-to-back pipe bend setup analyzed is subjected to a spectrum of steady internal pressures and cyclic in-plane bending moments. A previously developed direct non-cyclic simplified technique for determining elastic shakedown limit loads is utilized to generate the elastic shakedown boundary of the analyzed structure. The simplified technique outcomes showed excellent correlation with the results of full elastic–plastic cyclic loading finite element simulations. - Highlights: • No shakedown experimental data exist for 90° back-to-back pipe bends. • A non-cyclic technique is utilized to generate the elastic shakedown boundary. • The non-cyclic technique succeeded in generating the structure's Bree diagram. • The non-cyclic technique correlated well with full cyclic loading FE simulations

  11. Thermodynamic investigation of the phase equilibrium boundary between TiO2 rutile and its α-PbO2-type high-pressure polymorph

    Science.gov (United States)

    Kojitani, Hiroshi; Yamazaki, Monami; Kojima, Meiko; Inaguma, Yoshiyuki; Mori, Daisuke; Akaogi, Masaki

    2018-06-01

    Heat capacity (C P) of rutile and α-PbO2 type TiO2 (TiO2-II) were measured by the differential scanning calorimetry and thermal relaxation method. Using the results, standard entropies at 1 atm and 298.15 K of rutile and TiO2-II were determined to be 50.04(4) and 46.54(2) J/mol K, respectively. Furthermore, thermal expansivity (α) determined by high-temperature X-ray diffraction measurement and mode Grüneisen parameters obtained by high-pressure Raman spectroscopy suggested the thermal Grüneisen parameter (γ th) for TiO2-II of 1.7(1). By applying the obtained low-temperature C P and γ th, the measured C P and α data of TiO2-II were extrapolated to higher temperature region using a lattice vibrational model calculation, as well as rutile. Internally consistent thermodynamic data sets of both rutile and TiO2-II assessed in this study were used to thermodynamically calculate the rutile‒TiO2-II phase equilibrium boundary. The most plausible boundary was obtained to be P (GPa) = 0.0074T (K) - 1.7. Our boundary suggests that the crystal growth of TiO2-II observed below 5.5 GPa and 900 K in previous studies advanced in its stability field. The phase boundary calculation also suggested small, exothermic phase transition enthalpy from rutile to TiO2-II at 1 atm and 298.15 K of - 0.5 to - 1.1 kJ/mol. This implies that the thermodynamic stability of rutile at 1 atm above room temperature is due to larger contribution of entropy term.

  12. The influence of selected containment structures on debris dispersal and transport following high pressure melt ejection from the reactor vessel

    International Nuclear Information System (INIS)

    Pilch, M.; Tarbell, W.W.; Brockmann, J.E.

    1988-09-01

    High pressure expulsion of molten core debris from the reactor pressure vessel may result in dispersal of the debris from the reactor cavity. In most plants, the cavity exits into the containment such that the debris impinges on structures. Retention of the debris on the structures may affect the further transport of the debris throughout the containment. Two tests were done with scaled structural shapes placed at the exit of 1:10 linear scale models of the Zion cavity. The results show that the debris does not adhere significantly to structures. The lack of retention is attributed to splashing from the surface and reentrainment in the gas flowing over the surface. These processes are shown to be applicable to reactor scale. A third experiment was done to simulate the annular gap between the reactor vessel and cavity wall. Debris collection showed that the fraction of debris exiting through the gap was greater than the gap-to-total flow area ratio. Film records indicate that dispersal was primarily by entrainment of the molten debris in the cavity. 29 refs., 36 figs., 11 tabs

  13. Test results on direct containment heating by high-pressure melt ejection into the Surtsey vessel: The TDS test series

    International Nuclear Information System (INIS)

    Allen, M.D.; Blanchat, T.K.; Pilch, M.M.

    1994-08-01

    The Technology Development and Scoping (TDS) test series was conducted to test and develop instrumentation and procedures for performing steam-driven, high-pressure melt ejection (HPME) experiments at the Surtsey Test Facility to investigate direct containment heating (DCH). Seven experiments, designated TDS-1 through TDS-7, were performed in this test series. These experiments were conducted using similar initial conditions; the primary variable was the initial pressure in the Surtsey vessel. All experiments in this test series were performed with a steam driving gas pressure of ≅ 4 MPa, 80 kg of lumina/iron/chromium thermite melt simulant, an initial hole diameter of 4.8 cm (which ablated to a final hole diameter of ≅ 6 cm), and a 1/10th linear scale model of the Surry reactor cavity. The Surtsey vessel was purged with argon ( 2 ) to limit the recombination of hydrogen and oxygen, and gas grab samples were taken to measure the amount of hydrogen produced

  14. Round-robin pretest analyses of a 1:6-scale reinforced concrete containment model subject to static internal pressurization

    International Nuclear Information System (INIS)

    Clauss, D.B.

    1987-05-01

    Analyses of a 1:6-scale reinforced concrete containment model that will be tested to failure at Sandia National Laboratories in the spring of 1987 were conducted by the following organizations in the United States and Europe: Sandia National Laboratories (USA), Argonne National Laboratory (USA), Electric Power Research Institute (USA), Commissariat a L'Energie Atomique (France), HM Nuclear Installations Inspectorate (UK), Comitato Nazionale per la ricerca e per lo sviluppo dell'Energia Nucleare e delle Energie Alternative (Italy), UK Atomic Energy Authority, Safety and Reliability Directorate (UK), Gesellschaft fuer Reaktorsicherheit (FRG), Brookhaven National Laboratory (USA), and Central Electricity Generating Board (UK). Each organization was supplied with a standard information package, which included construction drawings and actual material properties for most of the materials used in the model. Each organization worked independently using their own analytical methods. This report includes descriptions of the various analytical approaches and pretest predictions submitted by each organization. Significant milestones that occur with increasing pressure, such as damage to the concrete (cracking and crushing) and yielding of the steel components, and the failure pressure (capacity) and failure mechanism are described. Analytical predictions for pressure histories of strain in the liner and rebar and displacements are compared at locations where experimental results will be available after the test. Thus, these predictions can be compared to one another and to experimental results after the test

  15. Effect of Spray System on Fission Product Distribution in Containment During a Severe Accident in a Two-Loop Pressurized Water Reactor

    Directory of Open Access Journals (Sweden)

    Mehdi Dehjourian

    2016-08-01

    Full Text Available The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

  16. A study on the pressurized water reactor (PWR) containment response analysis methodologies for postulated severe accident

    International Nuclear Information System (INIS)

    Ahn, Kwang Il

    1992-02-01

    The present study contains two major parts: one is the treatment of uncertainties involved in the current APET and the other is the importance analysis of the APET uncertainty inputs. A clear disadvantage of the expert opinion polling process approach for uncertainty analysis of the current probabilistic risk assessment (PRA) is that the sufficient robustness in the final results may not be attained against the ambiguity of the information upon which the experts base their judgement or the judgmental uncertainty arising under various imprecise and incomplete information. For the treatment of such type of uncertainty, a new approach based on fuzzy set theory is proposed. Then its potential use to the uncertainty analysis of the current PRA is proved through an analysis of accident progression event tree (APET). As a product, a formal procedure with computational algorithms suitable for application of the fuzzy set theory to the APET analysis is provided. Comparing with the uncertainty analysis results obtained by the statistical approach currently used in PRA, the present approach has several major advantages: Firstly, it greatly enhances the robustness in the final results of APET uncertainty analysis by modeling the judgmental uncertainty that arises in the probabilistic quantification of APET top events. Secondly, the modeling of APET uncertainty analysis is far more convenient because of the nonprobabilistic features of fuzzy probabilities used for uncertainty quantification of the APET top events. Thirdly, the APET model can easily be operated by means of a well defined formal propagation logic of fuzzy set theory without going through a tedious sampling procedure. Finally, the fuzzy outcomes provide at least as much information as the existing methods based on the statistical approach. Thus, the present approach can be used as a valuable alternative approach to uncertainty analysis used in the current PRA. Two importance measures for the importance analysis of

  17. Design, Modeling and Optimization of a Piezoelectric Pressure Sensor based on a Thin-Film PZT Membrane Containing Nanocrystalline Powders

    Directory of Open Access Journals (Sweden)

    Vahid MOHAMMADI

    2009-11-01

    Full Text Available In this paper fabrication of a 0-3 ceramic/ceramic composite lead zirconate titanate, Pb(Zr0.52Ti0.48O3 thin film has been presented and then a pressure sensor based on multilayer thin-film PZT diaphragm contain of Lead Zirconate Titanate nanocrystalline powders was designed, modeled and optimized. Dynamics characteristics of this multilayer diaphragm have been investigated by ANSYS® FE software. By this simulation the effective parameters of the multilayer PZT diaphragm for improving the performance of a pressure sensor in different ranges of pressure are optimized. The optimized thickness ratio of PZT layer to SiO2 was given in the paper to obtain the maximum deflection of the multilayer thin-film PZT diaphragm. A 0-3 ceramic/ceramic composite lead zirconate titanate, Pb(Zr0.52Ti0.48O3 film has been developed to fabricate the pressure sensor by a hybrid sol gel process. PZT nanopowders fabricated via conventional sol gel method and uniformly dispersed in PZT precursor solution by an attrition mill. XRD analysis shows that perovskite structure would be formed due to the presence of a significant amount of ceramic nanopowders. This texture has a good effect on piezoelectric properties of perovskite structure. The film forms a strongly bonded network and less shrinkage occurs, so the films do not crack during process. Also the aspect ratio through this process would be increased. SEM micrographs indicated that PZT films were uniform, crack free and have a composite microstructure and a piezoelectric coefficient d31 of -40 pC.N-1 and d33 ranged from 50pm.N-1 to 60pm.N-1.

  18. Interaction between a normal shock wave and a turbulent boundary layer at high transonic speeds. Part 1: Pressure distribution. Part 2: Wall shear stress. Part 3: Simplified formulas for the prediction of surface pressures and skin friction

    Science.gov (United States)

    Adamson, T. C., Jr.; Liou, M. S.; Messiter, A. F.

    1980-01-01

    An asymptotic description is derived for the interaction between a shock wave and a turbulent boundary layer in transonic flow, for a particular limiting case. The dimensionless difference between the external flow velocity and critical sound speed is taken to be much smaller than one, but large in comparison with the dimensionless friction velocity. The basic results are derived for a flat plate, and corrections for longitudinal wall curvature and for flow in a circular pipe are also shown. Solutions are given for the wall pressure distribution and the shape of the shock wave. Solutions for the wall shear stress are obtained, and a criterion for incipient separation is derived. Simplified solutions for both the wall pressure and skin friction distributions in the interaction region are given. These results are presented in a form suitable for use in computer programs.

  19. Thermodynamic model of a containment with pressure suppression pool for parametric studies to support the conceptual design

    International Nuclear Information System (INIS)

    Mueller, Pablo

    2004-01-01

    The aim of this work was to develop a model to simulate the evolution of the thermodynamic variables in a nuclear reactor containment with pressure suppression pool under accidental transients.We wanted a program able to give fast results, to facilitate the physical interpretation of the phenomena involved, and to make parametric studies.We did not pretend to get a precise result of a particular case.The program was made to be used as a design tool for the containment and to solve the interactions with the primary cooling system and the other security systems of the reactor, on a conceptual design context.The model consists on energy and mass balances on control volumes with liquid water, steam and a non-condensable gas like air.The dynamics of the system is shown with a base case during a loss of coolant accident.Sensibility and effects of varying some important parameters like volumes and heat and mass transfer coefficients are studied.Finally the results for the CAREM-25 reactor are compared with the codes CORAN, MELCOR 1.8.4 and CONTAIN 2.0 [es

  20. Boundaries of mantle–lithosphere domains in the Bohemian Massif as extinct exhumation channels for high-pressure rocks

    Czech Academy of Sciences Publication Activity Database

    Babuška, Vladislav; Plomerová, Jaroslava

    2013-01-01

    Roč. 23, č. 3 (2013), s. 973-987 ISSN 1342-937X R&D Projects: GA ČR GA205/07/1088; GA ČR GAP210/12/2381; GA AV ČR IAA300120709 Institutional research plan: CEZ:AV0Z30120515 Keywords : Bohemian Massif * mantle lithosphere domains * fossil olivine fabric * high pressure Subject RIV: DC - Siesmology, Volcanology, Earth Structure Impact factor: 8.122, year: 2013

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-07-01

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

  2. Brain stem slice conditioned medium contains endogenous BDNF and GDNF that affect neural crest boundary cap cells in co-culture.

    Science.gov (United States)

    Kaiser, Andreas; Kale, Ajay; Novozhilova, Ekaterina; Siratirakun, Piyaporn; Aquino, Jorge B; Thonabulsombat, Charoensri; Ernfors, Patrik; Olivius, Petri

    2014-05-30

    Conditioned medium (CM), made by collecting medium after a few days in cell culture and then re-using it to further stimulate other cells, is a known experimental concept since the 1950s. Our group has explored this technique to stimulate the performance of cells in culture in general, and to evaluate stem- and progenitor cell aptitude for auditory nerve repair enhancement in particular. As compared to other mediums, all primary endpoints in our published experimental settings have weighed in favor of conditioned culture medium, where we have shown that conditioned culture medium has a stimulatory effect on cell survival. In order to explore the reasons for this improved survival we set out to analyze the conditioned culture medium. We utilized ELISA kits to investigate whether brain stem (BS) slice CM contains any significant amounts of brain-derived neurotrophic factor (BDNF) and glial cell derived neurotrophic factor (GDNF). We further looked for a donor cell with progenitor characteristics that would be receptive to BDNF and GDNF. We chose the well-documented boundary cap (BC) progenitor cells to be tested in our in vitro co-culture setting together with cochlear nucleus (CN) of the BS. The results show that BS CM contains BDNF and GDNF and that survival of BC cells, as well as BC cell differentiation into neurons, were enhanced when BS CM were used. Altogether, we conclude that BC cells transplanted into a BDNF and GDNF rich environment could be suitable for treatment of a traumatized or degenerated auditory nerve. Copyright © 2014 Elsevier B.V. All rights reserved.

  3. Planar time-resolved PIV for velocity and pressure retrieval in atmospheric boundary layer over surface waves.

    Science.gov (United States)

    Troitskaya, Yuliya; Kandaurov, Alexander; Sergeev, Daniil; Bopp, Maximilian; Caulliez, Guillemette

    2017-04-01

    Air-sea coupling in general is important for weather, climate, fluxes. Wind wave source is crucially important for surface waves' modeling. But the wind-wave growth rate is strongly uncertain. Using direct measurements of pressure by wave-following Elliott probe [1] showed, weak and indefinite dependence of wind-wave growth rate on the wave steepness, while Grare et.al. [2] discuss the limitations of direct measurements of pressure associated with the inability to measure the pressure close to the surface by contact methods. Recently non-invasive methods for determining the pressure on the basis of technology of time-resolved PIV are actively developed [3]. Retrieving air flow velocities by 2D PIV techniques was started from Reul et al [4]. The first attempt for retrieving wind pressure field of waves in the laboratory tank from the time-resolved PIV measurements was done in [5]. The experiments were performed at the Large Air-Sea Interaction Facility (LASIF) - MIO/Luminy (length 40 m, cross section of air channel 3.2 x 1.6 m). For 18 regimes with wind speed up to 14 m/s including presence of puddle waves, a combination of time resolved PIV technique and optical measurements of water surface form was applied to detailed investigation of the characteristics of the wind flow over the water surface. Ammonium chloride smoke was used for flow visualization illuminated by two 6 Wt blue diode lasers combined into a vertical laser plane. Particle movement was captured with high-speed camera using Scheimpflug technique (up to 20 kHz frame rate with 4-frame bursts, spatial resolution about 190 μm, field of view 314x12 mm). Velocity air flow field was retrieved by PIV images processing with adaptive cross-correlation method on the curvilinear grid following surface wave form. The resulting time resolved instantaneous velocity fields on regular grid allowed us to obtain momentum fluxes directly from measured air velocity fluctuations. The average wind velocity patterns were

  4. Assessment of the Internal Pressure Fragility of the Hanul NPP Units 3 and 4 Containment Building Using a Nonlinear Finite Element Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hyung Kui; Hahm, Dea Gi; Choi, In Kil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    The sensitivity of the concrete strength is relatively higher compared to that of the steel strength. According to changes in the structure of the material, about 6-10% ultimate internal pressure differences occurred. Thirty sets of an FE model considering the material uncertainty of concrete and steel were composed for the internal pressure fragility assessment. From the internal pressure fragility assessment of the target containment building, the median capacity of liner leakage is estimated to be 116 psi. As can be seen from the Fukushima nuclear power plant accident, the containment building is the final protecting shield to prevent radiation leakage. Thus, a structural soundness evaluation for the containment pressure loads owing to a severe accident is very important. Recently, a probabilistic safety assessment has been commonly used to take into account the possible factors of uncertainty in a structural system. An assessment of the internal pressure fragility of the CANDU type containment buildings considering the correlation of structural material variables, and an assessment of the internal pressure fragility of the CANDU type containment buildings using a nonlinear finite element analysis, were also performed. However, for PWR type containment buildings, a fragility assessment has not been performed yet using a nonlinear finite element model (FEM) analysis. In this study, for the Hanul NPP units 3 and 4 containment building, the internal pressure fragility assessment was established using an FEM analysis. To do this, a three-dimensional finite element model, material property values, and a sensitive analysis were developed. A nonlinear finite element analysis of the Hanul NPP units 3 and 4 containment building was performed for a material sensitivity analysis and internal pressure fragility assessment.

  5. Assessment of the Internal Pressure Fragility of the Hanul NPP Units 3 and 4 Containment Building Using a Nonlinear Finite Element Analysis

    International Nuclear Information System (INIS)

    Park, Hyung Kui; Hahm, Dea Gi; Choi, In Kil

    2013-01-01

    The sensitivity of the concrete strength is relatively higher compared to that of the steel strength. According to changes in the structure of the material, about 6-10% ultimate internal pressure differences occurred. Thirty sets of an FE model considering the material uncertainty of concrete and steel were composed for the internal pressure fragility assessment. From the internal pressure fragility assessment of the target containment building, the median capacity of liner leakage is estimated to be 116 psi. As can be seen from the Fukushima nuclear power plant accident, the containment building is the final protecting shield to prevent radiation leakage. Thus, a structural soundness evaluation for the containment pressure loads owing to a severe accident is very important. Recently, a probabilistic safety assessment has been commonly used to take into account the possible factors of uncertainty in a structural system. An assessment of the internal pressure fragility of the CANDU type containment buildings considering the correlation of structural material variables, and an assessment of the internal pressure fragility of the CANDU type containment buildings using a nonlinear finite element analysis, were also performed. However, for PWR type containment buildings, a fragility assessment has not been performed yet using a nonlinear finite element model (FEM) analysis. In this study, for the Hanul NPP units 3 and 4 containment building, the internal pressure fragility assessment was established using an FEM analysis. To do this, a three-dimensional finite element model, material property values, and a sensitive analysis were developed. A nonlinear finite element analysis of the Hanul NPP units 3 and 4 containment building was performed for a material sensitivity analysis and internal pressure fragility assessment

  6. Comparing systolic and diastolic Blood pressure changes and heartbeat rate following administration of anesthetics containing epinephrine and felypressin

    Directory of Open Access Journals (Sweden)

    M. Jafari

    1998-05-01

    Full Text Available   Complex mechanisms have been known for keeping blood pressure in normal level. In fact, these mechanisms have inter-related functions and can be dysregulated by both internal and external stimuli while cardiovascular system functions to minimize these changes. Vasoconstrictors can cause clinical and hemodynamical changes as 1-2 cartridges of epinephrine containing lidocaine can has no considerable effects in a normal individual ( unless administered IV but 3 cartridges can bring about some clinical symptoms, according to a number of investigations. In current study, epinephrine’s effect on heartbeat rate was found more potent than felypressin which is considered as a disadvantage. on the other hand, epinephrine acts on arteries and can cause less bleeding, less drug toxicity and deeper and longer anesthesia. Therefore, it is preferred to felypressin due to its better action. It should be noted that the changes resulted by epinephrine and felypressin are of no significant importance in healthy individuals.

  7. Development of Probability Evaluation Methodology for High Pressure/Temperature Gas Induced RCS Boundary Failure and SG Creep Rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Hong, Soon Joon; Lee, Jin Yong; Lee, Kyung Jin; Lee, Kuh Hyung [FNC Tech. Co., Seoul (Korea, Republic of)

    2008-04-15

    Existing MELCOR 1.8.5 model was improved in view of severe accident natural circulation and MELCOR 1.8.6 input model was developed and calculation sheets for detailed MELCOR 1.8.6 model were produced. Effects of natural circulation modeling were found by simulating SBO accident by comparing existing model with detailed model. Major phenomenon and system operations which affect on natural circulation by high temperature and high pressure gas were investigated and representative accident sequences for creep rupture model of RCS pipeline and SG tube were selected.

  8. Interaction between a normal shock wave and a turbulent boundary layer at high transonic speeds. Part 1: Pressure distribution

    Science.gov (United States)

    Messiter, A. F.

    1979-01-01

    Analytical solutions are derived which incorporate additional physical effects as higher order terms for the case when the sonic line is very close to the wall. The functional form used for the undisturbed velocity profile is described to indicate how various parameters will be calculated for later comparison with experiment. The basic solutions for the pressure distribution are derived. Corrections are added for flow along a wall having longitudinal curvature and for flow in a circular pipe, and comparisons with available experimental data are shown.

  9. Analysis study on change of tendon behavior during pressurization process of Pre-stressed Concrete Containment Vessel

    International Nuclear Information System (INIS)

    Kashiwase, Takako; Nagasaka, Hideo

    1999-01-01

    NUPEC has been planning the ultimate strength test of Pre-stressed Concrete Containment Vessel (PCCV). The test model is 1/4 uniform scale model of Japan actual PCCV. It involves an equipment hatch, several penetrations and liner with T-anchors. The ancillary test for the PCCV test was conducted, in which friction coefficient of hoop tendon was evaluated by tensile force distribution using the same tendon as that of 1/4 PCCV model. Tendon will be in plastic region under internal pressure above 3.5 times design pressure (Pd) and surface characteristic of tendon and the resultant friction coefficient will be changed. In the present paper, tendon friction coefficient in the plastic region was obtained by evaluating plastic region data of tendon in the ancillary test. The validity of the obtained friction coefficient was confirmed by the tendon elongation data. In addition to the formally developed elastic region friction coefficient, the obtained plastic region correlation was incorporated into ABAQUS Ver. 5.6. The effect of tendon tensile force distribution change on structural behavior up to 3.8 Pd was evaluated. (author)

  10. Fluorescent silica hybrid materials containing benzimidazole dyes obtained by sol-gel method and high pressure processing

    International Nuclear Information System (INIS)

    Hoffmann, Helena Sofia; Stefani, Valter; Benvenutti, Edilson Valmir; Costa, Tania Maria Haas; Gallas, Marcia Russman

    2011-01-01

    Research highlights: → Sol-gel technique was used to obtain silica based hybrid materials containing benzimidazole dyes. → The sol-gel catalysts, HF and NaF, produce xerogels with different optical and textural characteristics. → High pressure technique (6.0 GPa) was used to produce fluorescent and transparent silica compacts with the dyes entrapped in closed pores, maintaining their optical properties. → The excited state intramolecular proton transfer (ESIPT) mechanism of benzimidazole dyes was studied by steady-state fluorescence spectroscopy for the monoliths, powders, and compacts. - Abstract: New silica hybrid materials were obtained by incorporation of two benzimidazole dyes in the silica network by sol-gel technique, using tetraethylorthosilicate (TEOS) as inorganic precursor. Several syntheses were performed with two catalysts (HF and NaF) producing powders and monoliths with different characteristics. The dye 2-(2'-hydroxy-5'-aminophenyl)benzimidazole was dispersed and physically adsorbed in the matrix, and the dye 2'(5'-N-(3-triethoxysilyl)propylurea-2'-hydroxyphenyl)benzimidazole was silylated, becoming chemically bonded to the silica network. High pressure technique was used to produce fluorescent and transparent silica compacts with the silylated and incorporated dye, at 6.0 GPa and room temperature. The excited state intramolecular proton transfer (ESIPT) mechanism of benzimidazole dyes was studied by steady-state fluorescence spectroscopy for the monoliths, powders, and compacts. The influence of the syntheses conditions was investigated by textural analysis using nitrogen adsorption isotherms.

  11. Analyses on interaction of internal and external surface cracks in a pressurized cylinder by hybrid boundary element method

    International Nuclear Information System (INIS)

    Chai Guozhong; Fang Zhimin; Jiang Xianfeng; Li Gan

    2004-01-01

    This paper presents a comprehensive range of analyses on the interaction of two identical semi-elliptical surface cracks at the internal and external surfaces of a pressurized cylinder. The considered ratios of the crack depth to crack length are b/a=0.25, 0.5, 0.75 and 1.0; the ratios of the crack depth to wall thickness of the cylinder are 2b/t=0.2, 0.4, 0.6, 0.7 and 0.8. Forty crack configurations are analyzed and the stress intensity factors along the crack front are presented. The numerical results show that for 2b/t<0.7, the interaction leads to a decrease in the stress intensity factors for both internal and external surface cracks, compared with a single internal or external surface crack. Thus for fracture analysis of a practical pressurized cylinder with two identical semi-elliptical surface cracks at its internal and external surfaces, a conservative result is obtained by ignoring the interaction

  12. Evaluation of containment peak pressure and structural response for a large-break loss-of-coolant accident in a VVER-440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Sienicki, J.J.; Kulak, R.F.; Pfeiffer, P.A. [Argonne National Lab., IL (United States); Voeroess, L.; Techy, Z. [VEIKI Inst. for Electric Power Research, Budapest (Hungary); Katona, T. [Paks Nuclear Power Plant (Hungary)

    1998-07-01

    A collaborative effort between US and Hungarian specialists was undertaken to investigate the response of a VVER-440/213-type NPP to a maximum design-basis accident, defined as a guillotine rupture with double-ended flow from the largest pipe (500 mm) in the reactor coolant system. Analyses were performed to evaluate the magnitude of the peak containment pressure and temperature for this event; additional analyses were performed to evaluate the ultimate strength capability of the containment. Separate cases were evaluated assuming 100% effectiveness of the bubbler-condenser pressure suppression system as well as zero effectiveness. The pipe break energy release conditions were evaluated from three sources: (1) FSAR release rate based on Soviet safety calculations, (2) RETRAN-03 analysis and (3) ATHLET analysis. The findings indicated that for 100% bubbler-condenser effectiveness the peak containment pressures were less than the containment design pressure of 0.25 MPa. For the BDBA case of zero effectiveness of the bubbler-condenser system, the peak pressures were less than the calculated containment failure pressure of 0.40 MPa absolute.

  13. Penetration of gas into concrete during a leakage rate test of reactor containments and its significance for the drop in pressure

    Directory of Open Access Journals (Sweden)

    Nilsson L.-O.

    2011-04-01

    Full Text Available The objective of the project described in the paper was to develop a simulation model that describes transient air pressure distribution in concrete in order to see if the leakage rates obtained from the Containment Integrated Leakage Rate Tests can be explained by the transient air pressurization of concrete pores inside the steel liner. A partial differential equation was derived which describes transient air pressure distribution in concrete pores. The model was validated against experimental results. The simulation model shows that there are significant air fluxes into the concrete structures that can explain the pressure drop during a leakage test.

  14. Effect of oxygen partial pressure on the density of antiphase boundaries in Fe3O4 thin films on Si(100)

    Science.gov (United States)

    Singh, Suraj Kumar; Husain, Sajid; Kumar, Ankit; Chaudhary, Sujeet

    2018-02-01

    Polycrystalline Fe3O4 thin films were grown on Si(100) substrate by reactive DC sputtering at different oxygen partial pressures PO2 for controlling the growth associated density of antiphase boundaries (APBs). The micro-Raman analyses were performed to study the structural and electronic properties in these films. The growth linked changes in the APBs density are probed by electron-phonon coupling strength (λ) and isothermal magnetization measurements. The estimated values of λ are found to vary from 0.39 to 0.56 with the increase in PO2 from 2.2 × 10-5 to 3.0 × 10-5 Torr, respectively. The saturation magnetization (saturation field) values are found to increase (decrease) from 394 (5.9) to 439 (3.0) emu/cm3 (kOe) with the increase in PO2 . The sharp Verwey transition (∼120 K), low saturation field, high saturation magnetization and low value of λ (comparable to the bulk value ∼0.51) clearly affirm the negligible amount of APBs in the high oxygen partial pressure deposited thin films.

  15. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    International Nuclear Information System (INIS)

    Ritter, S.; Seifert, H.P.; Devrient, B.; Roth, A.; Ehrnsten, U.; Ernestova, M.; Zamboch, M.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  16. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part I: BWR/NWC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWR) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It was focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of EAC crack growth behaviour/mechanism of LAS in high-temperature water under steady-state power operation (constant load) and transient operating conditions (e.g., start-up/shut-down, transients in water chemistry and load). Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurised water reactor (VVER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (VVER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarises the most important crack growth results obtained under simulated BWR/NWC conditions. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  17. Storage of hydrogen in advanced high pressure container. Final report for PSO projekt; Lagring af brint i avancerede hoejtryksbeholdere. Slutrapport for PSO-projekt

    Energy Technology Data Exchange (ETDEWEB)

    Christiansen, Jens

    2006-04-15

    The objective of the project has been to study barriers for a production of advanced high pressure containers especially suitable for hydrogen, in order to create a basis for a container production in Denmark. The project has primarily focused on future Danish need for hydrogen storage in the MWh area. One task has been to examine requirement specifications for pressure tanks that can be expected in connection with these stores. Six potential storage needs have been identified: (1) Buffer in connection with start-up/regulation on the power grid. (2) Hydrogen and oxygen production. (3) Buffer store in connection with VEnzin vision. (4) Storage tanks on hydrogen filling stations. (5) Hydrogen for the transport sector from 1 TWh surplus power. (6) Tanker transport of hydrogen. Requirements for pressure containers for the above mentioned use have been examined. The connection between stored energy amount, pressure and volume compared to liquid hydrogen and oil has been stated in tables. As starting point for production technological considerations and economic calculations of various container concepts, an estimation of laminate thickness in glass-fibre reinforced containers with different diameters and design print has been made, for a 'pure' fibre composite container and a metal/fibre composite container respectively. (BA)

  18. Reverse water-in-fluorocarbon emulsions for use in pressurized metered-dose inhalers containing hydrofluoroalkane propellants.

    Science.gov (United States)

    Butz, N; Porté, C; Courrier, H; Krafft, M P; Vandamme, Th F

    2002-05-15

    Pulmonary administration of drugs has demonstrated numerous advantages in the treatment of pulmonary diseases due to direct targeting to the respiratory tract. It enables avoiding the first pass effect, reduces the amount of drugs administered, targets drugs to specific sites and reduces their side effects. Reverse water-in-fluorocarbon (FC) emulsions are potential drug delivery systems for pulmonary administration using pressurized metered-dose inhalers (pMDI). The external phase of these emulsions consists of perfluorooctyl bromide (PFOB, perflubron), whereas their internal phase contains the drugs solubilized or dispersed in water. These emulsions are stabilized by a perfluoroalkylated dimorpholinophosphate (F8H11DMP), i.e. a fluorinated surfactant. This study demonstrates the possibility of delivering a reverse fluorocarbon emulsion via the pulmonary route using a CFC-free pMDI. Two hydrofluoroalkanes (HFAs) (Solkane(R) 134a and Solkane(R) 227) were used as propellants, and various solution (or emulsion)/propellant ratios (1/3, 1/2, 2/3, 1/1, 3/2, 3/1 v/v) were investigated. The insolubility of water (with or without the fluorinated surfactant F8H11DMP) in both HFA 227 and HFA 134a was demonstrated. PFOB and the reverse emulsion were totally soluble or dispersible in all proportions in both propellants. This study demonstrated also that the reverse FC emulsion can be successfully used to deliver caffeine in a homogeneous and reproducible way. The mean diameter of the emulsion water droplets in the pressured canister was investigated immediately after packaging and after 1 week of storage at room temperature. Best results were obtained with emulsion/propellant ratios comprised between 2/3 and 3/2, and with HFA 227 as propellant.

  19. Monitoring localized cracks on under pressure concrete nuclear containment wall using linear and nonlinear ultrasonic coda wave interferometry

    Science.gov (United States)

    Legland, J.-B.; Abraham, O.; Durand, O.; Henault, J.-M.

    2018-04-01

    Civil engineering is constantly demanding new methods for evaluation and non-destructive testing (NDT), particularly to prevent and monitor serious damage to concrete structures. Tn this work, experimental results are presented on the detection and characterization of cracks using nonlinear modulation of coda waves interferometry (NCWT) [1]. This method consists in mixing high-amplitude low-frequency acoustic waves with multi-scattered probe waves (coda) and analyzing their effects by interferometry. Unlike the classic method of coda analysis (CWT), the NCWT does not require the recording of a coda as a reference before damage to the structure. Tn the framework of the PTA-ENDE project, a 1/3 model of a preconstrained concrete containment (EDF VeRCoRs mock-up) is placed under pressure to study the leakage of the structure. During this evaluation protocol, specific areas are monitored by the NCWT (during 5 days, which correspond to the protocol of nuclear power plant pressurization under maintenance test). The acoustic nonlinear response due to the high amplitude of the acoustic modulation gives pertinent information about the elastic and dissipative nonlinearities of the concrete. Tts effective level is evaluated by two nonlinear observables extracted from the interferometry. The increase of nonlinearities is in agreement with the creation of a crack with a network of microcracks located at its base; however, a change in the dynamics of the evolution of the nonlinearities may indicate the opening of a through crack. Tn addition, as during the experimental campaign, reference codas have been recorded. We used CWT to follow the stress evolution and the gas leaks ratio of the structure. Both CWT and NCWT results are presented in this paper.

  20. Dissolved organic matter cycling in eastern Mediterranean rivers experiencing multiple pressures. The case of the trans-boundary Evros River

    Directory of Open Access Journals (Sweden)

    E. PITTA

    2014-07-01

    Full Text Available The objective of our study was to provide a comprehensive evaluation on C, N, P cycling in medium sized Mediterranean rivers, such as the Evros, experiencing multiple pressures (intensive agriculture, industrial activities, population density. Our work aims also to contribute to the development of integrated management policies. Dissolved organic matter (DOM cycling were investigated, during a one-year study. It was shown that the organic component of N and P was comparable to those of large Mediterranean rivers (Rhone, Po. In the lower parts of the river where all point and non-point inputs converge, the high inorganic N input favour elevated assimilation rates by phytoplankton and result in increased chl-a concentrations and autochthonous dissolved organic matter (DOM production during the dry season with limited water flow. Moreover, carbohydrate distribution revealed that there is a constant background of soil derived mono-saccharides on top of which are superimposed impulses of poly-saccharides during blooms. During the dry season, inorganic nutrients and DOM are trapped in the lower parts of the river, whereas during high flow conditions DOM is flushed towards the sea and organic nitrogen forms can become an important TDN constituent (at least 40% transported to shelf waters. The co-existence of terrigenous material with autochthonous and some anthropogenic is supported by the relatively low DOC:DON and DOC:DOP ratios, the positive correlation of DOC vs chl-a and the decoupling between DOC and DON. Overall, this study showed that in medium size Mediterranean rivers, such as the Evros, intensive agriculture and pollution sources in combination with water management practices and climatic variability are important factors determining C, N, P dynamics and export to coastal seas. Also, it highlights the importance of the organic fraction of N and P when considering management practices.

  1. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    Energy Technology Data Exchange (ETDEWEB)

    Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

    2004-07-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  2. Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

    International Nuclear Information System (INIS)

    Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

    2004-01-01

    One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

  3. Simulation of a scenario of total loss of external and internal power (Sbo) for different vent pressures of the containment of a BWR-5

    International Nuclear Information System (INIS)

    Cardenas V, J.; Mugica R, C. A.; Godinez S, V.

    2014-10-01

    The simulation of a Station Black Out (Sbo) was realized with intervention of the vent containment by means of a rigid vent coming from the dry-well and that discharges directly to the atmosphere, with the MELCOR code version 2.1. This scenario was carried out for a BWR-5 and containment type Mark II, with a thermal power of 2317 MWt similar to the reactor of nuclear power plant of Laguna Verde. For this scenario was considered as only available system for coolant injection to the reactor to the Reactor Core Isolation Cooling (Rcic), which remained operating 4 hours with batteries bank. The Security and Relief Valves (SR V) were considered functional (by simplicity) and that they mechanically do not exceed their capacity to liberate pressure due to the performances in their safety way. The operator maneuver to perform the SR V and to de pressurize the vessel until the pressure (13 kg/cm 2 ) to operate the low pressure systems was modeled. The results cover approximately 48 hours (172000 seconds), time in which was observed the behavior of the level and pressure in the vessel. Also the scenario evolution was analyzed to different vent pressures of the primary containment (2.0, 3.0, 4.5, 6.0, and 10.0 kg/cm 2 ), the temperature profiles of the dry-well, the hydrogen accumulation in the containment, the radio-nuclides liberation through rigid vent to the atmosphere and the inventory of these. In this work an analysis of the pressure behavior in the primary containment is presented, with the purpose of minimizing liberated fission products to the environment. (Author)

  4. Containment integrity analysis under accidents

    International Nuclear Information System (INIS)

    Lin Chengge; Zhao Ruichang; Liu Zhitao

    2010-01-01

    Containment integrity analyses for current nuclear power plants (NPPs) mainly focus on the internal pressure caused by design basis accidents (DBAs). In addition to the analyses of containment pressure response caused by DBAs, the behavior of containment during severe accidents (SAs) are also evaluated for AP1000 NPP. Since the conservatism remains in the assumptions,boundary conditions and codes, margin of the results of containment integrity analyses may be overestimated. Along with the improvements of the knowledge to the phenomena and process of relevant accidents, the margin overrated can be appropriately reduced by using the best estimate codes combined with the uncertainty methods, which could be beneficial to the containment design and construction of large passive plants (LPP) in China. (authors)

  5. Development of boundary layers

    International Nuclear Information System (INIS)

    Herbst, R.

    1980-01-01

    Boundary layers develop along the blade surfaces on both the pressure and the suction side in a non-stationary flow field. This is due to the fact that there is a strongly fluctuating flow on the downstream blade row, especially as a result of the wakes of the upstream blade row. The author investigates the formation of boundary layers under non-stationary flow conditions and tries to establish a model describing the non-stationary boundary layer. For this purpose, plate boundary layers are measured, at constant flow rates but different interferent frequency and variable pressure gradients. By introducing the sample technique, measurements of the non-stationary boundary layer become possible, and the flow rate fluctuation can be divided in its components, i.e. stochastic turbulence and periodical fluctuation. (GL) [de

  6. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    International Nuclear Information System (INIS)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W.; Nichols, R.T.; Sweet, D.W.

    1991-08-01

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs

  7. Experimental results of direct containment heating by high-pressure melt ejection into the Surtsey vessel: The DCH-3 and DCH-4 tests

    Energy Technology Data Exchange (ETDEWEB)

    Allen, M.D.; Pilch, M.; Brockmann, J.E.; Tarbell, W.W. (Sandia National Labs., Albuquerque, NM (United States)); Nichols, R.T. (Ktech Corp., Albuquerque, NM (United States)); Sweet, D.W. (AEA Technology, Winfrith (United Kingdom))

    1991-08-01

    Two experiments, DCH-3 and DCH-4, were performed at the Surtsey test facility to investigate phenomena associated with a high-pressure melt ejection (HPME) reactor accident sequence resulting in direct containment heating (DCH). These experiments were performed using the same experimental apparatus with identical initial conditions, except that the Surtsey test vessel contained air in DCH-3 and argon in DCH-4. Inerting the vessel with argon eliminated chemical reactions between metallic debris and oxygen. Thus, a comparison of the pressure response in DCH-3 and DCH-4 gave an indication of the DCH contribution due to metal/oxygen reactions. 44 refs., 110 figs., 43 tabs.

  8. CFD Simulations for the Effect of Unsteady Wakes on the Boundary Layer of a Highly Loaded Low-Pressure Turbine Airfoil (L1A)

    Science.gov (United States)

    Vinci, Samuel, J.

    2012-01-01

    This report is the third part of a three-part final report of research performed under an NRA cooperative Agreement contract. The first part was published as NASA/CR-2012-217415. The second part was published as NASA/CR-2012-217416. The study of the very high lift low-pressure turbine airfoil L1A in the presence of unsteady wakes was performed computationally and compared against experimental results. The experiments were conducted in a low speed wind tunnel under high (4.9%) and then low (0.6%) freestream turbulence intensity for Reynolds number equal to 25,000 and 50,000. The experimental and computational data have shown that in cases without wakes, the boundary layer separated without reattachment. The CFD was done with LES and URANS utilizing the finite-volume code ANSYS Fluent (ANSYS, Inc.) under the same freestream turbulence and Reynolds number conditions as the experiment but only at a rod to blade spacing of 1. With wakes, separation was largely suppressed, particularly if the wake passing frequency was sufficiently high. This was validated in the 3D CFD efforts by comparing the experimental results for the pressure coefficients and velocity profiles, which were reasonable for all cases examined. The 2D CFD efforts failed to capture the three dimensionality effects of the wake and thus were less consistent with the experimental data. The effect of the freestream turbulence intensity levels also showed a little more consistency with the experimental data at higher intensities when compared with the low intensity cases. Additional cases with higher wake passing frequencies which were not run experimentally were simulated. The results showed that an initial 25% increase from the experimental wake passing greatly reduced the size of the separation bubble, nearly completely suppressing it.

  9. Containment for low temperature district nuclear-heating reactor

    International Nuclear Information System (INIS)

    He Shuyan; Dong Duo

    1992-03-01

    Integral arrangement is adopted for Low Temperature District Nuclear-heating Reactor. Primary heat exchangers, control rod drives and spent fuel elements are put in the reactor pressure vessel together with reactor core. Primary coolant flows through reactor core and primary heat exchangers in natural circulation. Primary coolant pipes penetrating the wall of reactor pressure vessel are all of small diameters. The reactor vessel constitutes the main part of pressure boundary of primary coolant. Therefore the small sized metallic containment closed to the wall of reactor vessel can be used for the reactor. Design principles and functions of the containment are as same as the containment for PWR. But the adoption of small sized containment brings about some benefits such as short period of manufacturing, relatively low cost, and easy for sealing. Loss of primary coolant accident would not be happened during the rupture accident of primary coolant pressure boundary inside the containment owing to its intrinsic safety

  10. Dermal application of nitric oxide releasing acidified nitrite-containing liniments significantly reduces blood pressure in humans.

    Science.gov (United States)

    Opländer, Christian; Volkmar, Christine M; Paunel-Görgülü, Adnana; Fritsch, Thomas; van Faassen, Ernst E; Mürtz, Manfred; Grieb, Gerrit; Bozkurt, Ahmet; Hemmrich, Karsten; Windolf, Joachim; Suschek, Christoph V

    2012-02-15

    Vascular ischemic diseases, hypertension, and other systemic hemodynamic and vascular disorders may be the result of impaired bioavailability of nitric oxide (NO). NO but also its active derivates like nitrite or nitroso compounds are important effector and signal molecules with vasodilating properties. Our previous findings point to a therapeutical potential of cutaneous administration of NO in the treatment of systemic hemodynamic disorders. Unfortunately, no reliable data are available on the mechanisms, kinetics and biological responses of dermal application of nitric oxide in humans in vivo. The aim of the study was to close this gap and to explore the therapeutical potential of dermal nitric oxide application. We characterized with human skin in vitro and in vivo the capacity of NO, applied in a NO-releasing acidified form of nitrite-containing liniments, to penetrate the epidermis and to influence local as well as systemic hemodynamic parameters. We found that dermal application of NO led to a very rapid and significant transepidermal translocation of NO into the underlying tissue. Depending on the size of treated skin area, this translocation manifests itself through a significant systemic increase of the NO derivates nitrite and nitroso compounds, respectively. In parallel, this translocation was accompanied by an increased systemic vasodilatation and blood flow as well as reduced blood pressure. We here give evidence that in humans dermal application of NO has a therapeutic potential for systemic hemodynamic disorders that might arise from local or systemic insufficient availability of NO or its bio-active NO derivates, respectively. Copyright © 2012 Elsevier Inc. All rights reserved.

  11. Allegheny County Parcel Boundaries

    Data.gov (United States)

    Allegheny County / City of Pittsburgh / Western PA Regional Data Center — This dataset contains parcel boundaries attributed with county block and lot number. Use the Property Information Extractor for more control downloading a filtered...

  12. Matrimid®/polysulfone blend mixed matrix membranes containing ZIF-8 nanoparticles for high pressure stability in natural gas separation

    NARCIS (Netherlands)

    Shahid, S.; Nijmeijer, K.

    2017-01-01

    Plasticization is of important concern in high pressure natural gas separation. Majority of the pure polymers and MOF-MMM systems suffer from plasticization at low pressures. Combination of polymer blending and MMM approach could lead to plasticization resistant membranes with improved membrane

  13. Hierarchical structures and phase nucleation and growth during pressure-induced crystallization of polypropylene containing dispersion of nanoclay: The impact on physical and mechanical properties

    International Nuclear Information System (INIS)

    Misra, R.D.K.; Yuan, Q.; Chen, J.; Yang, Y.

    2010-01-01

    The objective of this study is to describe the evolution of structure and phases during pressure-induced crystallization of polymers containing dispersion of nanoparticles, in the pressure range of 0.1-200 MPa. The model material for nanoparticles is nanoclay and the model polymer is polypropylene, which can potentially form several crystalline phases. While the phase selection in polypropylene is dictated by pressure and temperature, however, the introduction of nanoparticles alters the nucleation and growth of phases via nanoparticle interface driven evolution. To delineate and separate the effects of applied crystallization pressure from nanoparticle effects, a relative comparison is made between neat polypropylene and polypropylene containing dispersion of nanoclay under similar experimental conditions. The significant finding is that nanoclay interacts with the host polypropylene in a manner such that it alters the structural morphology of α- and γ-crystals of polypropylene. Furthermore, nanoclay promotes the formation of γ-phase at ambient pressure suggesting its role as structure and morphology director in the stabilization of the less accessible γ-phase, and with the possibility of epitaxial growth that enhances toughness. The equilibrium melting point measurements point to thermodynamic interaction between nanoclay and polypropylene, which is supported by the change in glass transition temperature. Thus, the two components, nanoclay and pressure, together provide a unique opportunity to tune hierarchical structures and phase evolution, which has significant implication on physico-chemical and mechanical properties.

  14. A container

    DEFF Research Database (Denmark)

    2012-01-01

    A container assembly for the containment of fluids or solids under a pressure different from the ambient pressure comprising a container (2) comprising an opening and an annular sealing, a lid (3) comprising a central portion (5) and engagement means (7) for engaging the annular flange, and sealing...... means (10) wherein the engagement means (7) is adapted, via the sealing means, to seal the opening when the pressure of the container assembly differs from the ambient pressure in such a way that the central portion (5) flexes in the axial direction which leads to a radial tightening of the engagement...... means (7) to the container, wherein the container further comprises locking means (12) that can be positioned so that the central portion is hindered from flexing in at least one direction....

  15. Mosquito-Producing Containers, Spatial Distribution, and Relationship between Aedes aegypti Population Indices on the Southern Boundary of its Distribution in South America (Salto, Uruguay)

    Science.gov (United States)

    Basso, César; Caffera, Ruben M.; García da Rosa, Elsa; Lairihoy, Rosario; González, Cristina; Norbis, Walter; Roche, Ingrid

    2012-01-01

    A study was conducted in the city of Salto, Uruguay, to identify mosquito-producing containers, the spatial distribution of mosquitoes and the relationship between the different population indices of Aedes aegypti. On each of 312 premises visited, water-filled containers and immature Ae. aegypti mosquitoes were identified. The containers were counted and classified into six categories. Pupae per person and Stegomyia indices were calculated. Pupae per person were represented spatially. The number of each type of container and number of mosquitoes in each were analyzed and compared, and their spatial distribution was analyzed. No significant differences in the number of the different types of containers with mosquitoes or in the number of mosquitoes in each were found. The distribution of the containers with mosquito was random and the distribution of mosquitoes by type of container was aggregated or highly aggregated. PMID:23128295

  16. Mosquito-producing containers, spatial distribution, and relationship between Aedes aegypti population indices on the southern boundary of its distribution in South America (Salto, Uruguay).

    Science.gov (United States)

    Basso, César; Caffera, Ruben M; García da Rosa, Elsa; Lairihoy, Rosario; González, Cristina; Norbis, Walter; Roche, Ingrid

    2012-12-01

    A study was conducted in the city of Salto, Uruguay, to identify mosquito-producing containers, the spatial distribution of mosquitoes and the relationship between the different population indices of Aedes aegypti. On each of 312 premises visited, water-filled containers and immature Ae. aegypti mosquitoes were identified. The containers were counted and classified into six categories. Pupae per person and Stegomyia indices were calculated. Pupae per person were represented spatially. The number of each type of container and number of mosquitoes in each were analyzed and compared, and their spatial distribution was analyzed. No significant differences in the number of the different types of containers with mosquitoes or in the number of mosquitoes in each were found. The distribution of the containers with mosquito was random and the distribution of mosquitoes by type of container was aggregated or highly aggregated.

  17. Closure Welding of Plutonium Bearing Storage Containers

    International Nuclear Information System (INIS)

    Cannell, G.R.

    2002-01-01

    A key element in the Department of Energy (DOE) strategy for the stabilization, packaging and storage of plutonium-bearing materials involves closure welding of DOE-STD-3013 Outer Containers (3013 container). The 3013 container provides the primary barrier and pressure boundary preventing release of plutonium-bearing materials to the environment. The final closure (closure weld) of the 3013 container must be leaktight, structurally sound and meet DOE STD 3013 specified criteria. This paper focuses on the development, qualification and demonstration of the welding process for the closure welding of Hanford PFP 3013 outer containers

  18. CONTEMPT-LT/028: a computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Hargroves, D.W.; Metcalfe, L.J.; Wheat, L.L.; Niederauer, G.F.; Obenchain, C.F.

    1979-03-01

    CONTEMPT-LT is a digital computer program, written in FORTRAN IV, developed to describe the long-term behavior of water-cooled nuclear reactor containment systems subjected to postulated loss-of-coolant accident (LOCA) conditions. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and energy exchange with adjacent compartments. The program is capable of describing the effects of leakage on containment response. Models are provided to describe fan cooler and cooling spray engineered safety systems. An annular fan model is also provided to model pressure control in the annular region of dual containment systems. Up to four compartments can be modeled with CONTEMPT-LT, and any compartment except the reactor system may have both a liquid pool region and an air--vapor atmosphere region above the pool. Each region is assumed to have a uniform temperature, but the temperatures of the two regions may be different

  19. Reactor containment

    International Nuclear Information System (INIS)

    Kawabe, Ryuhei; Yamaki, Rika.

    1990-01-01

    A water vessel is disposed and the gas phase portion of the water vessel is connected to a reactor container by a pipeline having a valve disposed at the midway thereof. A pipe in communication with external air is extended upwardly from the liquid phase portion to a considerable height so as to resist against the back pressure by a waterhead in the pipeline. Accordingly, when the pressure in the container is reduced to a negative level, air passes through the pipeline and uprises through the liquid phase portion in the water vessel in the form of bubbles and then flows into the reactor container. When the pressure inside of the reactor goes higher, since the liquid surface in the water vessel is forced down, water is pushed up into the pipeline. Since the waterhead pressure of a column of water in the pipeline and the pressure of the reactor container are well-balanced, gases in the reactor container are not leaked to the outside. Further, in a case if a great positive pressure is formed in the reactor container, the inner pressure overcomes the waterhead of the column of water, so that the gases containing radioactive aerosol uprise in the pipeline. Since water and the gases flow being in contact with each other, this can provide the effect of removing aerosol. (T.M.)

  20. Analysis of some antecipated transients without scram for a pressurized water cooled reactor (PWR) using coupling of the containment code CORAN to the system model code ALMOD

    International Nuclear Information System (INIS)

    Carvalho, F. de A.T. de.

    1985-01-01

    Some antecipated transients without scram (ATWS) for a pressurized water cooled reactor, model KWU 1300 MWe, are studied using coupling of the containment code CORAN to the system model code ALMOD, under severe random conditions. This coupling has the objective of including containment model as part of a unified code system. These severe conditions include failure of reactor scram, following a station black-out and emergency power initiation for the burn-up status at the beginning and end of the cycle. Furthermore, for the burn-up status at the end of the cycle a failure in the closure of the pressurizer relief valve was also investigated. For the beginning of the cycle, the containment participates actively during the transient. It is noted that the effect of the burn-up in the fuel is to reduce the seriousness of these transients. On the other hand, the failure in the closure of the pressurized relief valve makes this transients more severe. Moreover, the containment safety or radiological public safety is not affected in any of the cases. (Author) [pt

  1. Note: implementation of a cold spot setup for controlled variation of vapor pressures and its application to an InBr containing discharge lamp.

    Science.gov (United States)

    Briefi, S

    2013-02-01

    In order to allow for a systematic investigation of the plasma properties of discharges containing indium halides, which are proposed as an efficient alternative for mercury based low pressure discharge lamps, a controlled variation of the indium halide density is mandatory. This can be achieved by applying a newly designed setup in which a well-defined cold spot location is implemented and the cold spot temperature can be adjusted between 50 and 350 °C without influencing the gas temperature. The performance of the setup has been proved by comparing the calculated evaporated InBr density (using the vapor pressure curve) with the one measured via white light absorption spectroscopy.

  2. White-Beam X-ray Diffraction and Radiography Studies on High-Boron Containing Borosilicate Glass at High Pressures

    Science.gov (United States)

    Ham, Kathryn; Vohra, Yogesh; Kono, Yoshio; Wereszczak, Andrew; Patel, Parimal

    Multi-angle energy-dispersive x-ray diffraction studies and white-beam x-ray radiography were conducted with a cylindrically shaped (1 mm diameter and 0.7 mm high) high-boron content borosilicate glass sample (17.6% B2O3) to a pressure of 13.7 GPa using a Paris-Edinburgh (PE) press at Beamline 16-BM-B, HPCAT of the Advanced Photon Source. The measured structure factor S(q) to large q = 19 Å-1, is used to determine information about the internuclear bond distances between various species of atoms within the glass sample. Sample pressure was determined with gold as a pressure standard. The sample height as measured by radiography showed an overall uniaxial compression of 22.5 % at 13.7 GPa with 10.6% permanent compaction after decompression to ambient conditions. The reduced pair distribution function G(r) was extracted and Si-O, O-O, and Si-Si bond distances were measured as a function of pressure. Raman spectroscopy of pressure recovered sample as compared to starting material showed blue-shift and changes in intensity and widths of Raman bands associated with silicate and B3O6 boroxol rings. US Army Research Office under Grant No. W911NF-15-1-0614.

  3. Process for producing curved surface of membrane rings for large containers, particulary for prestressed concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1977-01-01

    Membrane rings for large pressure vessels, particularly for prestressed-concrete pressure vessels, often have curved surfaces. The invention describes a process of producing these at site, which is particularly advantageous as the forming and installation of the vessel component coincide. According to the invention, the originally flat membrane ring is set in a predetermined position, is then pressed in sections by a forming tool (with a preformed support ring as opposite tool), and shaped. After this, the shaped parts are welded to the ring-shaped wall parts of the large vessel. The manufacture of single and double membrane rings arrangements is described. (HP) [de

  4. Method of producing the arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels of nuclear reactors

    International Nuclear Information System (INIS)

    Kumpf, H.

    1976-01-01

    In producing arched surfaces of diaphragm rings for large containers, especially for prestressed-concrete pressure vessels for nuclear power plants, it is of advantage to manufacture these directly on the construction site. According to the invention the, at first level, diaphragm ring is put on the predetermined place, sectionally pressed against and shaped by a shaping tool - with a profiled supporting ring as a counter-acting tool - and afterwards welded together with the annular wall sections of the large container along the shaped parts. The manufacture of single and double configurations of diaphragm rings is described. It is of advantage if shaping and mounting position coincide. (UWI) [de

  5. Operation method for wall surface of pressure suppression chamber of reactor container and floating scaffold used for the method

    International Nuclear Information System (INIS)

    Matsuzaki, Tetsuo; Kounomaru, Toshimi; Saito, Koichi.

    1996-01-01

    A floating scaffold is provisionally disposed in adjacent with the wall surface of pool water of a pressure suppression chamber while being floated on the surface of the pool water before the drainage of the pool water from the pressure vessel. The floating scaffold has guide rollers sandwiching a bent tube of an existent facility so that the horizontal movement is restrained, and is movable only in a vertical direction depending on the change of water level of the pool water. In addition, a handrail for preventing dropping, and a provisional illumination light are disposed. When pool water in the pressure suppression chamber is drained, the water level of the pool water is lowered in accordance with the amount of drained water. The floating scaffold floating on the water surface is lowered while being guided by the bent tube, and the operation position is lowered. An operator riding on the floating scaffold inspects the wall surfaces of the pressure chamber and conducts optional repair and painting. (I.N.)

  6. Procedure for qualification of electric equipment installed in containments for pressurized water reactors subject to accident conditions

    International Nuclear Information System (INIS)

    1991-11-01

    This generic norm is usable for electrical equipment installed in containment building of PWR subject to accidental conditions. She defines the qualification methods and the general rules usable for the test specifications of qualification for these materials

  7. Benefits of oxygen in CuInSe{sub 2} and CuGaSe{sub 2} containing Se-rich grain boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Feng, Chunbao, E-mail: chunbaofeng@126.com [Department of Mathematics and Physics, Chongqing University of Posts and Telecommunications, Chongqing, 400065 (China); Luo, Min; Li, Bolin; Li, Dengfeng [Department of Mathematics and Physics, Chongqing University of Posts and Telecommunications, Chongqing, 400065 (China); Nie, Jinlan [Department of Applied Physics, University of Electronic Science and Technology of China, Chengdu, 610054 (China); Dong, Huining [Department of Mathematics and Physics, Chongqing University of Posts and Telecommunications, Chongqing, 400065 (China)

    2014-05-01

    Using density functional theory calculation, we show that oxygen (O) exhibits an interesting effect in CuInSe{sub 2} and CuGaSe{sub 2}. The Se atoms with dangling bonds in a Se-rich Σ3 (114) grain boundary (GB) create deep gap states due to strong interaction between Se atoms. However, when such a Se atom is substituted by an O atom, the deep gap states can be shifted into valence band, making the site no longer a harmful non-radiative recombination center. We find that O atoms prefer energetically to substitute these Se atoms and induce significant lattice relaxation due to their smaller atomic size and stronger electronegativity, which effectively reduces the anion–anion interaction. Consequently, the deep gap states are shifted to lower energy regions close or even below the top of the valence band.

  8. Diets containing salmon fillet delay development of high blood pressure and hyperfusion damage in kidneys in obese Zucker fa/fa rats.

    Science.gov (United States)

    Vikøren, Linn A; Drotningsvik, Aslaug; Mwakimonga, Angela; Leh, Sabine; Mellgren, Gunnar; Gudbrandsen, Oddrun A

    2018-04-01

    Hypertension is the leading risk factor for cardiovascular and chronic renal diseases, affecting more than 1 billion people. Fish intake is inversely correlated with the prevalence of hypertension in several, but not all, studies, and intake of fish oil and fish proteins has shown promising potential to delay development of high blood pressure in rats. The effects of baked and raw salmon fillet intake on blood pressure and renal function were investigated in obese Zucker fa/fa rats, which spontaneously develop hypertension with proteinuria and renal failure. Rats were fed diets containing baked or raw salmon fillet in an amount corresponding to 25% of total protein from salmon and 75% of protein from casein, or casein as the sole protein source (control group) for 4 weeks. Results show lower blood pressure and lower urine concentrations of albumin and cystatin C (relative to creatinine) in salmon diet groups when compared to control group. Morphological examinations revealed less prominent hyperfusion damage in podocytes from rats fed diets containing baked or raw salmon when compared to control rats. In conclusion, diets containing baked or raw salmon fillet delayed the development of hypertension and protected against podocyte damage in obese Zucker fa/fa rats. Copyright © 2018 American Heart Association. Published by Elsevier Inc. All rights reserved.

  9. Overpressurization performance of containment structures

    International Nuclear Information System (INIS)

    Barr, P.; Bleackley, M.; Harrop, L.P.; Hargreaves, J.; Jowett, J.; Phillips, D.W.

    1987-01-01

    The containment building of a PWR is the outermost engineered barrier between the reactor and the environment. The most important element of such a containment system is the pressure boundary structure and its associated seals and penetrations. This containment structure is designed deterministically to withstand a number of loads and load combinations of which the dominant one is generally the internal pressure due to the double-ended guillotine break in one of the primary circuit loops. Typically, the design basis large LOCA produces a peak pressure increase in the region of 0.3 MPa in some 10 seconds and with a duration of up to a few tens of seconds. The assessment of overpressure performance of the containment structure is a key component of the PWR safety case, and is usually carried out by estimating a static factor of safety to some failure limit state. These estimates can be made using simple force-balance calculations or complicated finite element calculations, and both approaches have merit. In this paper we examine these approaches and discuss their value in estimating failure pressures and failure modes for a variety of internal pressurization transients. This discussion covers both general design and risk considerations and is illustrated by numerical examples taken from previous and on-going analysis

  10. Structural modification of swai-fish (Pangasius hypophthalmus)-based emulsions containing non-meat protein additives by ultra-high pressure and thermal treatments

    Science.gov (United States)

    Techarang, Jiranat; Apichartsrangkoon, Arunee; Phanchaisri, Boonrak; Pathomrungsiyoungkul, Pattavara; Sriwattana, Sujinda

    2017-07-01

    Swai-fish emulsions containing fermented soybeans (thua nao and rice-koji miso) were pressurized at 600 MPa for 20 min or heated at 72°C for 30 min. The fish batters were blended with soy protein isolate (SPI) or whey protein concentrate (WPC) to stabilize the emulsions. The processed fish emulsions were then subjected to physical, chemical and microbiological examinations. The results of gel strength and water-holding potential showed that SPI addition yielded higher impact on these properties than WPC addition, which was also confirmed by the interactions between SPI and native fish proteins depicted by electrophoregrams. The frequency profiles suggested that the heated gels had a greater storage and loss moduli than pressurized gels, while pressurized WPC set-gel displayed larger loss tangent (the predominance of viscous moiety) than those pressurized SPI set-gel. High bacteria and spore counts of B. subtilis (residual of the thua nao) were observed in both pressurized and heated fish-based emulsions.

  11. General design and main problems of a gas-heavy-water power reactor contained in a pressure vessel

    International Nuclear Information System (INIS)

    Roche, R.; Gaudez, J.C.

    1964-01-01

    In the framework of research carried out on a CO 2 -cooled power reactor moderated by heavy water, the so-called 'pressure vessel' solution involves the total integration of the core, of the primary circuit (exchanges and blowers) and of the fuel handling machine inside a single, strong, sealed vessel made of pre-stressed concrete. A vertical design has been chosen: the handling 'attic' is placed above the core, the exchanges being underneath. This solution makes it possible to standardize the type of reactor which is moderated by heavy-water or graphite and cooled by a downward stream of carbon dioxide gas; it has certain advantages and disadvantages with respect to the pressure tube solution and these are considered in detail in this report. Extrapolation presents in particular.problems due specifically to the heavy water (for example its cooling,its purification, the balancing of the pressures of the heavy water and of the gas, the assembling of the internal structures, the height of the attic, etc. (authors) [fr

  12. Reactor container

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu; Saba, Kazuhisa.

    1979-01-01

    Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

  13. Optimization and application of atmospheric pressure chemical and photoionization hydrogen-deuterium exchange mass spectrometry for speciation of oxygen-containing compounds.

    Science.gov (United States)

    Acter, Thamina; Kim, Donghwi; Ahmed, Arif; Jin, Jang Mi; Yim, Un Hyuk; Shim, Won Joon; Kim, Young Hwan; Kim, Sunghwan

    2016-05-01

    This paper presents a detailed investigation of the feasibility of optimized positive and negative atmospheric pressure chemical ionization (APCI) mass spectrometry (MS) and atmospheric pressure photoionization (APPI) MS coupled to hydrogen-deuterium exchange (HDX) for structural assignment of diverse oxygen-containing compounds. The important parameters for optimization of HDX MS were characterized. The optimized techniques employed in the positive and negative modes showed satisfactory HDX product ions for the model compounds when dichloromethane and toluene were employed as a co-solvent in APCI- and APPI-HDX, respectively. The evaluation of the mass spectra obtained from 38 oxygen-containing compounds demonstrated that the extent of the HDX of the ions was structure-dependent. The combination of information provided by different ionization techniques could be used for better speciation of oxygen-containing compounds. For example, (+) APPI-HDX is sensitive to compounds with alcohol, ketone, or aldehyde substituents, while (-) APPI-HDX is sensitive to compounds with carboxylic functional groups. In addition, the compounds with alcohol can be distinguished from other compounds by the presence of exchanged peaks. The combined information was applied to study chemical compositions of degraded oils. The HDX pattern, double bond equivalent (DBE) distribution, and previously reported oxidation products were combined to predict structures of the compounds produced from oxidation of oil. Overall, this study shows that APCI- and APPI-HDX MS are useful experimental techniques that can be applied for the structural analysis of oxygen-containing compounds.

  14. An analysis, using the CLAPTRAP code, of the pressure transients developed in the Carolinas Virginia Tube Reactor during containment performance tests

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1982-11-01

    To check containment performance of the CVTR, steam was injected above the operating floor through a 10 foot pipe cap containing the 1 inch diameter holes, at a steady rate of 102.8 lb/sec for a period of 166 seconds. This steam had an enthalpy of 1195 Btu/lb and was therefore not entirely typical of the much wetter material which would be rejected for the greater part of a true breached circuit accident. Pressure transients measured experimentally within the containment were compared with results calculated by the American code CONTEMPT and these results in turn have allowed the Winfrith code CLAPTRAP to be tested for consistency and to establish that the use of this code would have led to similar conclusions about the heat transfer coefficients at the heat absorbent surfaces. (U.K.)

  15. Comparative studies of the pressure - and temperature temporal behavior in the Angra I containment when submitted to the design basic accident

    International Nuclear Information System (INIS)

    Costa, J.R.

    1980-12-01

    A computer code - CONDRU 4 - was brought from Germany, that is being used for the determination of pressure - and temperature temporal behavior that occurs inside the metallic containment of PWR type reactors before the loss of coolant accident (LOCA). Simulation for Angra-1 reactor was made, considering the ocurrence of the worst postulated accident for the containment integrity. The results obtained with CONDRU 4 computer code were compared with those obtained by the CONTEMPT-LT-and COCO computer code for the same nuclear power plant. The discrepancy found among the results were due mainly to the different modes adopted in the several codes for the steam-water separation of coolant injected in the containment. (Author) [pt

  16. Combined effect of smear layer characteristics and hydrostatic pulpal pressure on dentine bond strength of HEMA-free and HEMA-containing adhesives.

    Science.gov (United States)

    Mahdan, Mohd Haidil Akmal; Nakajima, Masatoshi; Foxton, Richard M; Tagami, Junji

    2013-10-01

    This study evaluated the combined effect of smear layer characteristics with hydrostatic pulpal pressure (PP) on bond strength and nanoleakage expression of HEMA-free and -containing self-etch adhesives. Flat dentine surfaces were obtained from extracted human molars. Smear layers were created by grinding with #180- or #600-SiC paper. Three HEMA-free adhesives (Xeno V, G Bond Plus, Beautibond Multi) and two HEMA-containing adhesives (Bond Force, Tri-S Bond) were applied to the dentine surfaces under hydrostatic PP or none. Dentine bond strengths were determined using the microtensile bond test (μTBS). Data were statistically analyzed using three- and two-way ANOVA with Tukey post hoc comparison test. Nanoleakage evaluation was carried out under a scanning electron microscope (SEM). Coarse smear layer preparation and hydrostatic PP negatively affected the μTBS of HEMA-free and -containing adhesives, but there were no significant differences. The combined experimental condition significantly reduced μTBS of the HEMA-free adhesives, while the HEMA-containing adhesives exhibited no significant differences. Two-way ANOVA indicated that for HEMA-free adhesives, there were significant interactions in μTBS between smear layer characteristics and pulpal pressure, while for HEMA-containing adhesives, there were no significant interactions between them. Nanoleakage formation within the adhesive layers of both adhesive systems distinctly increased in the combined experimental group. The combined effect of coarse smear layer preparation with hydrostatic PP significantly reduced the μTBS of HEMA-free adhesives, while in HEMA-containing adhesives, these effects were not obvious. Smear layer characteristics and hydrostatic PP would additively compromise dentine bonding of self-etch adhesives, especially HEMA-free adhesives. Copyright © 2013 Elsevier Ltd. All rights reserved.

  17. High-pressure vapor-liquid equilibria of systems containing ethylene glycol, water and methane - Experimental measurements and modeling

    DEFF Research Database (Denmark)

    Folas, Georgios; Berg, Ole J.; Solbraa, Even

    2007-01-01

    This work presents new experimental phase equilibrium measurements of the binary MEG-methane and the ternary MEG-water-methane system at low temperatures and high pressures which are of interest to applications related to natural gas processing. Emphasis is given to MEG and water solubility...... measurements in the gas phase. The CPA and SRK EoS, the latter using either conventional or EoS/G(E) mixing rules are used to predict the solubility of the heavy components in the gas phase. It is concluded that CPA and SRK using the Huron-Vidal mixing rule perform equally satisfactory, while CPA requires...

  18. Reactor container

    International Nuclear Information System (INIS)

    Furukawa, Hideyasu; Oyamada, Osamu; Uozumi, Hiroto.

    1976-01-01

    Purpose: To provide a container for a reactor provided with a pressure suppressing chamber pool which can prevent bubble vibrating load, particularly negative pressure generated at the time of starting to release exhaust from a main steam escape-safety valve from being transmitted to a lower liner plate of the container. Constitution: This arrangement is characterized in that a safety valve exhaust pool for main steam escape, in which a pressure suppressing chamber pool is separated and intercepted from pool water in the pressure suppressing chamber pool, a safety valve exhaust pipe is open into said safety valve exhaust pool, and an isolator member, which isolates the bottom liner plate in the pressure suppressing chamber pool from the pool water, is disposed on the bottom of the safety valve exhaust pool. (Nakamura, S.)

  19. Emerging boundaries

    DEFF Research Database (Denmark)

    Løvschal, Mette

    2014-01-01

    of temporal and material variables have been applied as a means of exploring the processes leading to their socioconceptual anchorage. The outcome of this analysis is a series of interrelated, generative boundary principles, including boundaries as markers, articulations, process-related devices, and fixation...

  20. Changing Boundaries

    DEFF Research Database (Denmark)

    Brodkin, Evelyn; Larsen, Flemming

    2013-01-01

    project that is altering the boundary between the democratic welfare state and the market economy. We see workfare policies as boundary-changing with potentially profound implications both for individuals disadvantaged by market arrangements and for societies seeking to grapple with the increasing...

  1. Negotiating boundaries

    DEFF Research Database (Denmark)

    Aarhus, Rikke; Ballegaard, Stinne Aaløkke

    2010-01-01

    to maintain the order of the home when managing disease and adopting new healthcare technology. In our analysis we relate this boundary work to two continuums of visibility-invisibility and integration-segmentation in disease management. We explore five factors that affect the boundary work: objects......, activities, places, character of disease, and collaboration. Furthermore, the processes are explored of how boundary objects move between social worlds pushing and shaping boundaries. From this we discuss design implications for future healthcare technologies for the home.......To move treatment successfully from the hospital to that of technology assisted self-care at home, it is vital in the design of such technologies to understand the setting in which the health IT should be used. Based on qualitative studies we find that people engage in elaborate boundary work...

  2. Vapour pressures, osmotic and activity coefficients for binary mixtures containing (1-ethylpyridinium ethylsulfate + several alcohols) at T = 323.15 K

    International Nuclear Information System (INIS)

    Calvar, Noelia; Gomez, Elena; Dominguez, Angeles; Macedo, Eugenia A.

    2010-01-01

    Osmotic coefficients of binary mixtures containing several primary and secondary alcohols (1-propanol, 2-propanol, 1-butanol, 2-butanol, and 1-pentanol) and the pyridinium-based ionic liquid 1-ethylpyridinium ethylsulfate were determined at T = 323.15 K using the vapour pressure osmometry technique. From the experimental results, vapour pressure and activity coefficients can be determined. For the correlation of osmotic coefficients, the extended Pitzer model modified by Archer, and the modified NRTL (MNRTL) model were used, obtaining deviations lower than 0.017 and 0.047, respectively. The mean molal activity coefficients and the excess Gibbs free energy for the binary mixtures studied were determined from the parameters obtained with the extended Pitzer model modified by Archer.

  3. Long term integrity of reactor pressure vessel and primary containment vessel after the severe accidents in Fukushima Daiichi Nuclear Power Station. Leaching property of spent oxide fuel segment and corrosion property of a carbon steel under artificial seawater immersion

    International Nuclear Information System (INIS)

    2014-06-01

    Primary containment vessel (PCV), reactor pressure vessel and pedestal in Fukushima Daiichi Nuclear power station units 1 through 3 have been exposed to severe thermal, chemical and mechanical conditions due to core meltdown events and seawater injections for emergent core cooling. These components will be immersed in diluted seawater with dissolved fission products under irradiation until the end of debris removal. Fresh water injected into the cores contacts with debris to cool, dissolves or erodes their constituents, mixed with retained water, and becomes 'accumulated water' with radioactive nuclides. We have focused the leaching of fission products into the accumulated water under lower temperature (323 K). FUGEN spent oxide fuel segments were immersed to determine the leaching factor of fission product and actinide elements. Since PCV made from carbon steel is one of the most important boundaries to prevent from fission products release, corrosion behavior has been paid attention to evaluate their integrity. Carbon steel specimens were immersion- and electrochemical-tested in diluted seawater with simulants of the accumulated water at 323 K in order to evaluate the effect of fission products in particular cesium and radiation. (author)

  4. Reactor container

    International Nuclear Information System (INIS)

    Kato, Masami; Nishio, Masahide.

    1987-01-01

    Purpose: To prevent the rupture of the dry well even when the melted reactor core drops into a reactor pedestal cavity. Constitution: In a reactor container in which a dry well disposed above the reactor pedestal cavity for containing the reactor pressure vessel and a torus type suppression chamber for containing pressure suppression water are connected with each other, the pedestal cavity and the suppression chamber are disposed such that the flow level of the pedestal cavity is lower than the level of the pressure suppression water. Further, a pressure suppression water introduction pipeway for introducing the pressure suppression water into the reactor pedestal cavity is disposed by way of an ON-OFF valve. In case if the melted reactor core should fall into the pedestal cavity, the ON-OFF valve for the pressure suppression water introduction pipeway is opened to introduce the pressure suppression water in the suppression chamber into the pedestal cavity to cool the melted reactor core. (Ikeda, J.)

  5. On the use of expert judgments to estimate the pressure increment in the Sequoyah containment at vessel breach

    International Nuclear Information System (INIS)

    Chhibber, S.; Apostolakis, G.E.; Okrent, D.

    1994-01-01

    The use of expert judgments in probabilistic risk assessments has become common. Simple aggregation methods have often been used with the result that expert biases and interexpert dependence are often neglected. Sophisticated theoretical models for the use of expert opinions have been proposed that offer ways of incorporating expert biases and dependence, but they have not found wide acceptance because of the difficulty and rigor of these methods. Practical guidance on the use of the versatile Bayesian expert judgment aggregation model is provided. In particular, the case study of pressure increment due to vessel breach in the Sequoyah nuclear power plant is chosen to illustrate how phenomenological uncertainty can be addressed by using the Bayesian aggregation model. The results indicate that the Bayesian aggregation model is a suitable candidate model for aggregating expert judgments, especially if there is phenomenological uncertainty. Phenomenological uncertainty can be represented through the dependence parameter of the Bayesian model. This is because the sharing of assumptions by the experts tends to introduce dependence between the experts. The extent of commonality in the experts' beliefs can be characterized by assessing their interdependence. The results indicate that uncertainty is possibly underestimated by ignoring dependence

  6. Study of Gallium Arsenide Etching in a DC Discharge in Low-Pressure HCl-Containing Mixtures

    Science.gov (United States)

    Dunaev, A. V.; Murin, D. B.

    2018-04-01

    Halogen-containing plasmas are often used to form topological structures on semiconductor surfaces; therefore, spectral monitoring of the etching process is an important diagnostic tool in modern electronics. In this work, the emission spectra of gas discharges in mixtures of hydrogen chloride with argon, chlorine, and hydrogen in the presence of a semiconducting gallium arsenide plate were studied. Spectral lines and bands of the GaAs etching products appropriate for monitoring the etching rate were determined. It is shown that the emission intensity of the etching products is proportional to the GaAs etching rate in plasmas of HCl mixtures with Ar and Cl2, which makes it possible to monitor the etching process in real time by means of spectral methods.

  7. Mathematical simulation of gas pressure in fibre-reinforced concrete container at radiation and biological decomposition of cellulose, bituminized and concrete radwastes

    International Nuclear Information System (INIS)

    Kuruc, J.; Kvito, P.

    2005-01-01

    Fibre-reinforced concrete container (FRCC) are used for long-time repository of radioactive wastes. Low- and middle-active radwastes from operation of the NPPs V-1, V-2 Jaslovske Bohunice, Mochovce NPP and from decommissioned NPP A-1 (Jaslovske Bohunice) are treated in the plant SE-VYZ in Jaslovske Bohunice and after immobilisation are deposited in National Radwaste Repository Mochovce (RU RAO). After filling of the RU RAO, FRCC will be stored during 300 years. During this time the integrity of the FRCC must be guaranteed. By the influence of autoradiolysis of the cellulose and bituminized radwastes as well as in cement grout the gases are formed, mainly the hydrogen, methane and carbon dioxide. In the case of presence of available water (a w ≥ 0.63) and in presence of microbes and moulds at appropriate conditions the biological decomposition of cellulose materials may proceed with formation of H 2 , CH 4 a CO 2 . With increasing of developed gases may increase pressure in FRCC, that may initiate the loss of integrity of the FRCC with following endangering of radiation safety of the RU RAO, respectively of the territory over the repository.Authors developed the new mathematical model of pressure of gases in FRCC and in deposited barrels with cellulose and bituminized radwastes. The mathematical model is based on biological decomposition of cellulose materials as well as on radiation decomposition of cellulose, bitumen and concrete. In this mathematical model the diffusion through the walls of FRCC is the main process responsible for decreasing of the pressure. This model was developed in two basic variants: (1) Mathematical model of gas pressure in FRCC as function of dose; (2) Mathematical model of gas pressure in FRCC as function of mass of cellulose

  8. Reactor container

    International Nuclear Information System (INIS)

    Hidaka, Masataka; Hatamiya, Shigeo; Kawasaki, Terufumi; Fukui, Toru; Suzuki, Hiroaki; Kataoka, Yoshiyuki; Kawabe, Ryuhei; Murase, Michio; Naito, Masanori.

    1990-01-01

    In order to suppress the pressure elevation in a reactor container due to high temperature and high pressure steams jetted out upon pipeway rupture accidents in the reactor container, the steams are introduced to a pressure suppression chamber for condensating them in stored coolants. However, the ability for suppressing the pressure elevation and steam coagulation are deteriorated due to the presence of inactive incondensible gases. Then, there are disposed a vent channel for introducing the steams in a dry well to a pressure suppression chamber in the reactor pressure vessel, a closed space disposed at the position lower than a usual liquid level, a first channel having an inlet in the pressure suppression chamber and an exit in the closed space and a second means connected by way of a backflow checking means for preventing the flow directing to the closed space. The first paths are present by plurality, a portion of which constitutes a syphon. The incondensible gases and the steams are discharged to the dry well at high pressure by using the difference of the water head for a long cooling time after the pipeway rupture accident. Then, safety can be improved without using dynamic equipments as driving source. (N.H.)

  9. Boundary Spanning

    DEFF Research Database (Denmark)

    Zølner, Mette

    The paper explores how locals span boundaries between corporate and local levels. The aim is to better comprehend potentialities and challenges when MNCs draws on locals’ culture specific knowledge. The study is based on an in-depth, interpretive case study of boundary spanning by local actors in...... approach with pattern matching is a way to shed light on the tacit local knowledge that organizational actors cannot articulate and that an exclusively inductive research is not likely to unveil....

  10. Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

    International Nuclear Information System (INIS)

    Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

    1999-01-01

    In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

  11. The interior regularity of pressure associated with a weak solution to the Navier-Stokes equations with the Navier-type boundary conditions

    Czech Academy of Sciences Publication Activity Database

    Neustupa, Jiří; Al Baba, Hind

    2018-01-01

    Roč. 463, č. 1 (2018), s. 222-234 ISSN 0022-247X R&D Projects: GA ČR(CZ) GA17-01747S Institutional support: RVO:67985840 Keywords : Navier-Stokes equation * Navier-type boundary conditions * interior regularity Subject RIV: BA - General Mathematics OBOR OECD: Pure mathematics Impact factor: 1.064, year: 2016 https://www. science direct.com/ science /article/pii/S0022247X18302233?via%3Dihub

  12. The interior regularity of pressure associated with a weak solution to the Navier-Stokes equations with the Navier-type boundary conditions

    Czech Academy of Sciences Publication Activity Database

    Neustupa, Jiří; Al Baba, Hind

    2018-01-01

    Roč. 463, č. 1 (2018), s. 222-234 ISSN 0022-247X R&D Projects: GA ČR(CZ) GA17-01747S Institutional support: RVO:67985840 Keywords : Navier-Stokes equation * Navier-type boundary conditions * interior regularity Subject RIV: BA - General Mathematics OBOR OECD: Pure mathematics Impact factor: 1.064, year: 2016 https://www.sciencedirect.com/science/article/pii/S0022247X18302233?via%3Dihub

  13. Application of the High Gradient hydrodynamics code to simulations of a two-dimensional zero-pressure-gradient turbulent boundary layer over a flat plate

    Science.gov (United States)

    Kaiser, Bryan E.; Poroseva, Svetlana V.; Canfield, Jesse M.; Sauer, Jeremy A.; Linn, Rodman R.

    2013-11-01

    The High Gradient hydrodynamics (HIGRAD) code is an atmospheric computational fluid dynamics code created by Los Alamos National Laboratory to accurately represent flows characterized by sharp gradients in velocity, concentration, and temperature. HIGRAD uses a fully compressible finite-volume formulation for explicit Large Eddy Simulation (LES) and features an advection scheme that is second-order accurate in time and space. In the current study, boundary conditions implemented in HIGRAD are varied to find those that better reproduce the reduced physics of a flat plate boundary layer to compare with complex physics of the atmospheric boundary layer. Numerical predictions are compared with available DNS, experimental, and LES data obtained by other researchers. High-order turbulence statistics are collected. The Reynolds number based on the free-stream velocity and the momentum thickness is 120 at the inflow and the Mach number for the flow is 0.2. Results are compared at Reynolds numbers of 670 and 1410. A part of the material is based upon work supported by NASA under award NNX12AJ61A and by the Junior Faculty UNM-LANL Collaborative Research Grant.

  14. Temperature and pressure determination of the tin melt boundary from a combination of pyrometry, spectral reflectance, and velocity measurements along release paths

    Science.gov (United States)

    La Lone, Brandon; Asimow, Paul; Fatyanov, Oleg; Hixson, Robert; Stevens, Gerald

    2017-06-01

    Plate impact experiments were conducted on tin samples backed by LiF windows to determine the tin melt curve. Thin copper flyers were used so that a release wave followed the 30-40 GPa shock wave in the tin. The release wave at the tin-LiF interface was about 300 ns long. Two sets of experiments were conducted. In one set, spectral emissivity was measured at six wavelengths using a flashlamp illuminated integrating sphere. In the other set, thermal radiance was measured at two wavelengths. The emissivity and thermal radiance measurements were combined to obtain temperature histories of the tin-LiF interface during the release. PDV was used to obtain stress histories. All measurements were combined to obtain temperature vs. stress release paths. A kink or steepening in the release paths indicate where the releases merge onto the melt boundary, and release paths originating from different shock stresses overlap on the melt boundary. Our temperature-stress release path measurements provide a continuous segment of the tin melt boundary that is in good agreement with some of the published melt curves. This work was done by National Security Technologies, LLC, under Contract No. DE-AC52-06NA25946 with the U.S. Department of Energy, and supported by the Site-Directed Research and Development Program. DOE/NV/259463133.

  15. Computer aided probabilistic assessment of containment integrity

    International Nuclear Information System (INIS)

    Tsai, J.C.; Touchton, R.A.

    1984-01-01

    In the probabilistic risk assessment (PRA) of a nuclear power plant, there are three probability-based techniques which are widely used for event sequence frequency quantification (including nodal probability estimation). These three techniques are the event tree analysis, the fault tree analysis and the Bayesian approach for database development. In the barrier analysis for assessing radionuclide release to the environment in a PRA study, these techniques are employed to a greater extent in estimating conditions which could lead to failure of the fuel cladding and the reactor coolant system (RCS) pressure boundary, but to a lesser degree in the containment pressure boundary failure analysis. The main reason is that containment issues are currently still in a state of flux. In this paper, the authors describe briefly the computer programs currently used by the nuclear industry to do event tree analyses, fault tree analyses and the Bayesian update. The authors discuss how these computer aided probabilistic techniques might be adopted for failure analysis of the containment pressure boundary

  16. Integrability and boundary conditions of supersymmetric systems

    International Nuclear Information System (INIS)

    Yue Ruihong; Liang Hong

    1996-01-01

    By studying the solutions of the reflection equations, we find out a series of integrable supersymmetric systems with different boundary conditions. The Hamiltonian contains four free parameters which describe the contribution of the boundary terms

  17. Boundary of the State of Iowa

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — This coverage contains polygons representing the Iowa Boundary, it was derived from a coverage of county boundaries, called COUNTIES, of the state of Iowa. COUNTIES...

  18. Parcels and Land Ownership, This data set consists of digital map files containing parcel-level cadastral information obtained from property descriptions. Cadastral features contained in the data set include real property boundary lines, rights-of-way boundaries, property dimensions, Published in Not Provided, 1:2400 (1in=200ft) scale, Racine County Government.

    Data.gov (United States)

    NSGIC Local Govt | GIS Inventory — Parcels and Land Ownership dataset current as of unknown. This data set consists of digital map files containing parcel-level cadastral information obtained from...

  19. Reactor container

    International Nuclear Information System (INIS)

    Naruse, Yoshihiro.

    1990-01-01

    The thickness of steel shell plates in a reactor container embedded in sand cussions is monitored to recognize the corrosion of the steel shell plates. That is, the reactor pressure vessel is contained in a reactor container shell and the sand cussions are disposed on the lower outside of the reactor container shell to elastically support the shell. A pit is disposed at a position opposing to the sand cussions for measuring the thickness of the reactor container shell plates. The pit is usually closed by a closing member. In the reactor container thus constituted, the closing member can be removed upon periodical inspection to measure the thickness of the shell plates. Accordingly, the corrosion of the steel shell plates can be recognized by the change of the plate thickness. (I.S.)

  20. Blurring Boundaries

    DEFF Research Database (Denmark)

    Neergaard, Ulla; Nielsen, Ruth

    2010-01-01

    of welfare functions into EU law both from an internal market law and a constitutional law perspective. The main problem areas covered by the Blurring Boundaries project were studied in sub-projects on: 1) Internal market law and welfare services; 2) Fundamental rights and non-discrimination law aspects......; and 3) Services of general interest. In the Blurring Boundaries project, three aspects of the European Social Model have been particularly highlighted: the constitutionalisation of the European Social Model, its multi-level legal character, and the clash between market access justice at EU level...... and distributive justice at national level....

  1. A β-cyclodextrin, polyethyleneimine and silk fibroin hydrogel containing Centella asiatica extract and hydrocortisone acetate: releasing properties and in vivo efficacy for healing of pressure sores.

    Science.gov (United States)

    Lee, M S; Seo, S R; Kim, J-C

    2012-10-01

    Pressure sores are lesions caused by impaired blood flow. Conventional dressings can absorb exudates, but do not promote wound healing. A hydrogel composed of β-cyclodextrin (β-CD), polyethyleneimine (PEI) and silk fibroin (SF) was assessed for use in healing of pressure sores. The hydrogel was prepared by crosslinking β-CD-grafted PEI and SF using epichlorohydrin. The gel was then immersed in an aqueous solution of Centella asiatica extract (CAE) 0.7 mg/mL and/or hydrocortisone acetate (HCA) 0.5 mg/mL. The in vivo pressure sore-healing efficacy of the dry gel (with or without the drugs) was investigated in terms of the hyperplasia of epidermis and the number of neutrophils in the skin tissue. The specific loading of CAE was 0.0091 g/g of dry gel. The percentage of CAE released at 24 h at pH 3.0, 5.0 and 7.4 was approximately 63.9%, 55.0% and 44.4%, respectively. This pH-dependent release is possibly due to the degree of gel swelling, which decreased with increasing pH. The specific loading of HCA was 0.0050 g/g dry gel, and the percentage release of HCA at 24 h was around 20% at all three pH points. It is likely that HCA release is independent of pH. HCA is a hydrophobic compound, and therefore the release of HCA is affected by the partitioning of HCA between the β-CD cavity and the bulk water phase, but not by the degree of swelling of the hydrogel. The pressure sores treated with the hydrogel healed in 6 days, compared with 10 days for controls. In this study, a β-CD/PEI/SF hydrogel containing CAE and HCA reduced the healing time for pressure sores. © The Author(s). CED © 2012 British Association of Dermatologists.

  2. Fracture prevention and availability in the series safety of the pressure boundary of light water reactors. Materials for advanced reactor systems

    International Nuclear Information System (INIS)

    1983-01-01

    This brochure contains the full text of 23 lectures, which were given at the 9th MPA Seminar in October 1983. The main part of the work consists of investigations of fracture mechanics of reactor steel. (RW) [de

  3. Containment Modelling with the ASTEC Code

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Grgic, Davor

    2014-01-01

    ASTEC is an integral computer code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fur Anlagen-und Reaktorsicherheit (GRS, Germany) to assess the nuclear power plant behaviour during a severe accident (SA). It consists of 13 coupled modules which compute various SA phenomena in primary and secondary circuits of the nuclear power plants (NPP), and in the containment. The ASTEC code was used to model and to simulate NPP behaviour during a postulated station blackout accident in the NPP Krsko, a two-loop pressurized water reactor (PWR) plant. The primary system of the plant was modelled with 110 thermal hydraulic (TH) volumes, 113 junctions and 128 heat structures. The secondary system was modelled with 76 TH volumes, 77 junctions and 87 heat structures. The containment was modelled with 10 TH volumes by taking into account containment representation as a set of distinctive compartments, connected with 23 junctions. A total of 79 heat structures were used to simulate outer containment walls and internal steel and concrete structures. Prior to the transient calculation, a steady state analysis was performed. In order to achieve correct plant initial conditions, the operation of regulation systems was modelled. Parameters which were subjected to regulation were the pressurizer pressure, the pressurizer narrow range level and steam mass flow rates in the steam lines. The accident analysis was focused on containment behaviour, however the complete integral NPP analysis was carried out in order to provide correct boundary conditions for the containment calculation. During the accident, the containment integrity was challenged by release of reactor system coolant through degraded coolant pump seals and, later in the accident following release of the corium out of the reactor pressure vessel, by the molten corium concrete interaction and direct containment heating mechanisms. Impact of those processes on relevant

  4. Simulation of containment phenomena during the Phebus FPT1 test with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2002-01-01

    Thermal-hydraulic and aerosol phenomena which occurred in the containment vessel of the Phebus integral experimental facility during the first 30000 s of the Phebus FPT1 test were simulated with the CONTAIN thermal-hydraulic computer code. A single-cell input model of the vessel was developed, and boundary and initial conditions that were determined during the experiment were applied. The comparison of experimental and calculated results shows that, although the atmosphere temperature was well simulated, the calculated condensation rate was apparently too high, resulting in a lower pressure of the containment atmosphere. The aerosol deposition process was well simulated.(author)

  5. A boundary integral equation for boundary element applications in multigroup neutron diffusion theory

    International Nuclear Information System (INIS)

    Ozgener, B.

    1998-01-01

    A boundary integral equation (BIE) is developed for the application of the boundary element method to the multigroup neutron diffusion equations. The developed BIE contains no explicit scattering term; the scattering effects are taken into account by redefining the unknowns. Boundary elements of the linear and constant variety are utilised for validation of the developed boundary integral formulation

  6. A calculation technique to improve continuous monitoring of containment integrity

    International Nuclear Information System (INIS)

    Dick, J.E.

    1990-01-01

    The containment envelope of nuclear plants is a passive and extremely effective safety feature. World experience indicates, however, that inadvertent breaches of envelope integrity can go undetected for substantial time periods. Consequently, continuous monitoring of integrity is being closely examined by many containment designers and operators. The most promising approach is to use sensors and systems that automatically measure changes in the mass of air in containment, time integrate any known air mass flow rates across containment boundaries, and perform a mass balance to obtain the air mass leaked. As fluctuations in such measurements are typically too large to enable leakage to be calculated to the desired precision, filtering and statistical techniques must be used to filter out random and time-dependent fluctuations. Current approaches cannot easily deal with nonrandom or systematic fluctuations in the measurements, including pressure changes within the containment. As a result, sampling periods must be kept short, or data measured during periods of varying containment pressure must be discarded. The technique described allows for much longer sampling periods under conditions of fluctuating containment pressure and eliminates the invalidation of data when the containment pressure fluctuation is nonrandom. It should therefore yield a much more precise value for the containment leakage characteristic. It also promises to be able to distinguish the presence of systematic errors unrelated to systematic pressure changes and to establish whether the containment leakage characteristic is laminar or turbulent

  7. Effect of combination tablets containing amlodipine 10 mg and irbesartan 100 mg on blood pressure and cardiovascular risk factors in patients with hypertension

    Directory of Open Access Journals (Sweden)

    Yagi S

    2015-01-01

    Full Text Available Shusuke Yagi,1 Akira Takashima,1 Minoru Mitsugi,2 Toshihiro Wada,2 Junko Hotchi,1 Ken-ichi Aihara,3 Tomoya Hara,1 Masayoshi Ishida,1 Daiju Fukuda,4 Takayuki Ise,1 Koji Yamaguchi,1 Takeshi Tobiume,1 Takashi Iwase,1 Hirotsugu Yamada,1 Takeshi Soeki,1 Tetsuzo Wakatsuki,1 Michio Shimabukuro,4 Masashi Akaike,5 Masataka Sata11Department of Cardiovascular Medicine, Graduate School of Health Biosciences, University of Tokushima, Tokushima, 2Department of Internal Medicine, Shikoku Central Hospital, Shikokuchuo, 3Department of Medicine and Bioregulatory Sciences, 4Department of Cardio-Diabetes Medicine, 5Department of Medical Education, Graduate School of Health Biosciences, University of Tokushima, Tokushima, JapanBackground: Hypertension is one of the major risk factors for cardiovascular and cerebrovascular disease and mortality. Patients who receive insufficient doses of antihypertensive agents or who are poorly adherent to multidrug treatment regimens often fail to achieve adequate blood pressure (BP control. The aim of this study was to determine the efficacy of an angiotensin II receptor blocker (ARB and calcium channel blocker (CCB combination tablet containing a regular dose of irbesartan (100 mg and a high dose of amlodipine (10 mg with regard to lowering BP and other risk factors for cardiovascular disease.Methods: We retrospectively evaluated data from 68 patients with essential hypertension whose treatment regimen was changed either from combination treatment with an independent ARB and a low-dose or regular-dose CCB or from a combination tablet of ARB and a low-dose or regular-dose CCB to a combination tablet containing amlodipine 10 mg and irbesartan 100 mg, because of incomplete BP control. Previous treatments did not include irbesartan as the ARB.Results: The combination tablet decreased systolic and diastolic BP. In addition, it significantly decreased serum uric acid, low-density lipoprotein cholesterol, and increased high

  8. Improved containment isolation for CANDU plants

    Energy Technology Data Exchange (ETDEWEB)

    Stretch, A H [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    The publication of Regulatory Policy Statement R- 7 in February 1991 by the Atomic Energy Control Board imposes new requirement for the isolation of fluid piping penetrating the containment boundary. The Appendix of R-7 describes the detailed requirements for metal extensions of the containment envelope, including the code classification qf the pressure retaining portions and isolation requirements for fluid piping and tubing.The application of these new requirements to the existing CANDU 6 design resulted in a number of design changes, including the addition of manual and automatic isolation valves, changes in code classification, and leakage monitoring considerations. (author). 3 refs., 3 figs.

  9. Mean flow structure of non-equilibrium boundary layers with adverse ...

    Indian Academy of Sciences (India)

    According to them, an equilibrium boundary layer might exist if the pressure ... of adverse pressure gradient on the turbulent boundary layer at the flat plate for ..... of a constant-pressure turbulent layer to the sudden application of an sudden.

  10. Comparison of CONTAIN and TCE calculations for direct containment heating of Surry

    International Nuclear Information System (INIS)

    Washington, K.E.; Stuart, D.S.

    1996-01-01

    This paper presents the results of several CONTAIN code calculations used to model direct containment heating (DCH) loads for the Surry plant. The results of these calculations are compared with the results obtained using the two-cell equilibrium (TCE) model for the same set of initial and boundary conditions. This comparison is important because both models have been favorably validated against the available DCH database, yet there are potentially important modeling differences. The comparisons are to quantitatively assess the impact of these differences. A major conclusion of this study is that, for the accident conditions studied and for a broad range of sensitivity cases, the peak pressures predicted by both TCE and CONTAIN are well below the failure pressure for the Surry containment. (orig.)

  11. Containment heat removal system

    International Nuclear Information System (INIS)

    Wade, G.E.; Barbanti, G.; Gou, P.F.; Rao, A.S.; Hsu, L.C.

    1992-01-01

    This patent describes a nuclear system of a type including a containment having a nuclear reactor therein, the nuclear reactor including a pressure vessel and a core in the pressure vessel, the system. It comprises a gravity pool of coolant disposed at an elevation sufficient to permit a flow of coolant into the nuclear reactor pressure vessel against a predetermined pressure within the nuclear reactor pressure vessel; means for reducing a pressure of steam in the nuclear reactor pressure vessel to a value less than the predetermined pressure in the event of a nuclear accident, the means including a depressurization valve connected to the pressure vessel, the means further including steam heat dissipating means such dissipating means including a suppression pool; a supply of water in the suppression pool, there being a headspace in the suppression pool above the water supply; a substantial amount of air in the head space; means for feeding pressurized steam from the nuclear reactor pressure vessel to a location under a surface of the supply of water, the supply of water being effective to absorb heat sufficient to reduce steam pressure below the predetermined pressure; and a check valve for communicating the headspace with the containment, the check valve being oriented to vent air in the headspace to the containment when a pressure in the headspace exceeds a pressure in the containment by a predetermined pressure differential

  12. Bond durability of adhesives containing modified-monomer with/without-fluoride after aging in artificial saliva and under intrapulpal pressure simulation.

    Science.gov (United States)

    El-Deeb, H A; Al Sherbiney, H H; Mobarak, E H

    2013-01-01

    To evaluate the dentin bond strength durability of adhesives containing modified-monomer with/without-fluoride after storage in artificial saliva and under intrapulpal pressure simulation (IPPS). The occlusal enamel of 48 freshly extracted teeth was trimmed to expose midcoronal dentin. Roots were sectioned to expose the pulp chamber and to connect the specimens to the pulpal-pressure assembly. Specimens were assigned into four groups (n=12) according to adhesive system utilized: a two-step etch-and-rinse adhesive system (SB, Adper Single Bond 2, 3M ESPE), a two-step self-etch adhesive system (CSE, Clearfil SE Bond, Kuraray Medical Inc), and two single-step self-etch adhesives with the same modified monomer (bis-acrylamide)-one with fluoride (AOF, AdheSE One F, Ivoclar-Vivadent) and the other without (AO, AdheSE One, Ivoclar-Vivadent). Bonding was carried out while the specimens were subjected to 15-mm Hg IPPS. Resin composite (Valux Plus, 3M ESPE) buildups were made. After curing, specimens were aged in artificial saliva and under 20-mm Hg IPPS at 37°C in a specially constructed incubator either for 24 hours or six months prior to testing. Bonded specimens (n=6/group) were sectioned into sticks (n=24/group) with a cross section of 0.9 ± 0.01 mm(2) and subjected to microtensile bond strength (μTBS) testing using a universal testing machine. Data were statistically analyzed using two-way analysis of variance (ANOVA) with repeated measures, one-way ANOVA tests, and a t-test (partificial saliva and under IPPS, yet these values remained significantly higher than those for the other two adhesives with modified monomers. There was no significant difference in the bond strength values between fluoride-containing and fluoride-free self-etch adhesive systems (AOF and AO) after 24 hours or six months. Modes of failure were mainly adhesive and mixed. Based on the results of this study, 1) Fluoride addition did not affect dentin bond durability; and 2) despite the fact that

  13. Reactor container

    International Nuclear Information System (INIS)

    Kojima, Yoshihiro; Hosomi, Kenji; Otonari, Jun-ichiro.

    1997-01-01

    In the present invention, a catalyst for oxidizing hydrogen to be disposed in a reactor container upon rupture of pipelines of a reactor primary coolant system is prevented from deposition of water droplets formed from a reactor container spray to suppress elevation of hydrogen concentration in the reactor container. Namely, a catalytic combustion gas concentration control system comprises a catalyst for oxidizing hydrogen and a support thereof. In addition, there is also disposed a water droplet deposition-preventing means for preventing deposition of water droplets in a reactor pressure vessel on the catalyst. Then, the effect of the catalyst upon catalytic oxidation reaction of hydrogen can be kept high. The local elevation of hydrogen concentration can be prevented even upon occurrence of such a phenomenon that various kinds of mobile forces in the container such as dry well cooling system are lost. (I.S.)

  14. Performance of Sequoyah Containment Anchorage System

    International Nuclear Information System (INIS)

    Fanous, F.; Greimann, L.; Wassef, W.; Bluhm, D.

    1993-01-01

    Deformation of a steel containment anchorage system during a severe accident may result in a leakage path at the containment boundaries. Current design criteria are based on either ductile or brittle failure modes of headed bolts that do not account for factors such as cracking of the containment basemat or deformation of the anchor bolt that may affect the behavior of the containment anchorage system. The purpose of this study was to investigate the performance of a typical ice condenser containment's anchorage system. This was accomplished by analyzing the Sequoyah Containment Anchorage System. Based on a strength of materials approach and assuming that the anchor bolts are resisting the uplift caused by the internal pressure, one can estimate that the failure of the anchor bolts would occur at a containment pressure of 79 psig. To verify these results and to calibrate the strength of materials equation, the Sequoyah containment anchorage system was analyzed with the ABAQUS program using a three-dimensional, finite-element model. The model included portions of the steel containment building, shield building, anchor bolt assembly, reinforced concrete mat and soil foundation material

  15. Boundary issues

    Science.gov (United States)

    Townsend, Alan R.; Porder, Stephen

    2011-03-01

    What is our point of no return? Caesar proclaimed 'the die is cast' while crossing the Rubicon, but rarely does modern society find so visible a threshold in our continued degradation of ecosystems and the services they provide. Humans have always used their surroundings to make a living— sometimes successfully, sometimes not (Diamond 2005)—and we intuitively know that there are boundaries to our exploitation. But defining these boundaries has been a challenge since Malthus first prophesied that nature would limit the human population (Malthus 1798). In 2009, Rockström and colleagues tried to quantify what the 6.8 billion (and counting) of us could continue to get away with, and what we couldn't (Rockström et al 2009). In selecting ten 'planetary boundaries', the authors contend that a sustainable human enterprise requires treating a number of environmental thresholds as points of no return. They suggest we breach these Rubicons at our own peril, and that we've already crossed three: biodiversity loss, atmospheric CO2, and disruption of the global nitrogen (N) cycle. As they clearly hoped, the very act of setting targets has provoked scientific inquiry about their accuracy, and about the value of hard targets in the first place (Schlesinger 2009). Such debate is a good thing. Despite recent emphasis on the science of human-ecosystem interactions, understanding of our planetary boundaries is still in its infancy, and controversy can speed scientific progress (Engelhardt and Caplan 1987). A few weeks ago in this journal, Carpenter and Bennett (2011) took aim at one of the more controversial boundaries in the Rockström analysis: that for human alteration of the global phosphorus (P) cycle. Rockström's group chose riverine P export as the key indicator, suggesting that humans should not exceed a value that could trigger widespread marine anoxic events—and asserting that we have not yet crossed this threshold. There are defensible reasons for a marine

  16. boundary dissipation

    Directory of Open Access Journals (Sweden)

    Mehmet Camurdan

    1998-01-01

    are coupled by appropriate trace operators. This overall model differs from those previously studied in the literature in that the elastic chamber floor is here more realistically modeled by a hyperbolic Kirchoff equation, rather than by a parabolic Euler-Bernoulli equation with Kelvin-Voight structural damping, as in past literature. Thus, the hyperbolic/parabolic coupled system of past literature is replaced here by a hyperbolic/hyperbolic coupled model. The main result of this paper is a uniform stabilization of the coupled PDE system by a (physically appealing boundary dissipation.

  17. Requirements to be taken into account when designing safety-related mechanical components conveying or containing pressurized fluid and classified as level 2 or 3

    International Nuclear Information System (INIS)

    1984-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The purpose of this RFS is to specify the requirements to be taken into account when designing mechanical components conveying or containing pressurized fluid and which are in safety class 2 or 3

  18. Review on experiments relating to primary containment vessel failure

    International Nuclear Information System (INIS)

    Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

    2015-01-01

    Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

  19. Nuclear reactor containment device

    International Nuclear Information System (INIS)

    Ichiki, Tadaharu.

    1980-01-01

    Purpose: To reduce the volume of a containment shell and decrease the size of a containment equipment for BWR type reactors by connecting the containment shell and a suppression pool with slanted vent tubes to thereby shorten the vent tubes. Constitution: A pressure vessel containing a reactor core is installed at the center of a building and a containment vessel for the nuclear reactor that contains the pressure vessel forms a cabin. To a building situated below the containment shell, is provided a suppression chamber in which cooling water is charged to form a suppression pool. The suppression pool is communicated with vent tubes that pass through the partition wall of the containment vessel. The vent tubes are slanted and their lower openings are immersed in coolants. Therefore, if accident is resulted and fluid at high temperature and high pressure is jetted from the pressure vessel, the jetting fluid is injected and condensated in the cooling water. (Moriyama, K.)

  20. Analysis of the microbiota of refrigerated chopped parsley after treatments with a coating containing enterocin AS-48 or by high-hydrostatic pressure.

    Science.gov (United States)

    Grande Burgos, María José; López Aguayo, María Del Carmen; Pérez Pulido, Rubén; Galvez, Antonio; Lucas, Rosario

    2017-09-01

    Parsley can be implicated in foodborne illness, yet chopped parsley is used as an ingredient or garnish for multiple dishes. The aim of the present study was to determine the effect of two different treatments on the bacterial diversity of parsley: (i) coating with a pectin-EDTA solution containing the circular bacteriocin enterocin AS-48, and (ii) treatment by high hydrostatic pressure (HHP) at 600MPa for 8min. Control and treated parsley were stored in trays at 5°C for 10days. Both treatments reduced viable counts by 3.7 log cycles and retarded growth of survivors during storage. The bacterial diversity of the chopped parsley was studied by high throughput sequencing (Illumina Miseq). Bacterial diversity of control samples mainly consists of Proteobacteria (96.87%) belonging to genera Pseudomonas (69.12%), Rheinheimera (8.56%) and Pantoea (6.91%) among others. During storage, the relative abundance of Bacteroidetes (mainly Flavobacterium and Sphingobacterium) increased to 26.66%. Application of the pectin-bacteriocin-EDTA coating reduced the relative abundance of Proteobacteria (63.75%) and increased that of Firmicutes (34.70%). However, the relative abundances of certain groups such as Salmonella, Shigella and Acinetobacter increased at early storage times. Late storage was characterized by an increase in the relative abundance of Proteobacteria, mainly Pseudomonas. Upon application of HHP treatment, the relative abundance of Proteobacteria was reduced (85.88%) while Actinobacteria increased (8.01%). During early storage of HHP-treated samples, the relative abundance of Firmicutes increased. Potentially-pathogenic bacteria (Shigella) only increased in relative abundance by the end of storage. Results of the present study indicate that the two treatments had different effects on the bacterial diversity of parsley. The HHP treatment provided a safer product, since no potentially-pathogenic bacteria were detected until the end of the storage period. Copyright

  1. Wipe selection for the analysis of surface materials containing chemical warfare agent nitrogen mustard degradation products by ultra-high pressure liquid chromatography-tandem mass spectrometry.

    Science.gov (United States)

    Willison, Stuart A

    2012-12-28

    Degradation products arising from nitrogen mustard chemical warfare agent were deposited on common urban surfaces and determined via surface wiping, wipe extraction, and liquid chromatography–tandem mass spectrometry detection. Wipes investigated included cotton gauze, glass fiber filter, non-woven polyester fiber and filter paper, and surfaces included several porous (vinyl tile, painted drywall, wood) and mostly non-porous (laminate, galvanized steel, glass) surfaces. Wipe extracts were analyzed by ultra-high pressure liquid chromatography–tandem mass spectrometry (UPLC–MS/MS) and compared with high performance liquid chromatography–tandem mass spectrometry (HPLC–MS/MS) results. An evaluation of both techniques suggests UPLC–MS/MS provides a quick and sensitive analysis of targeted degradation products in addition to being nearly four times faster than a single HPLC run, allowing for greater throughput during a wide-spread release concerning large-scale contamination and subsequent remediation events. Based on the overall performance of all tested wipes, filter paper wipes were selected over other wipes because they did not contain interferences or native species (TEA and DEA) associated with the target analytes, resulting in high percent recoveries and low background levels during sample analysis. Other wipes, including cotton gauze, would require a pre-cleaning step due to the presence of large quantities of native species or interferences of the targeted analytes. Percent recoveries obtained from a laminate surface were 47–99% for all nitrogen mustard degradation products. The resulting detection limits achieved from wipes were 0.2 ng/cm(2) for triethanolamine (TEA), 0.03 ng/cm(2) for N-ethyldiethanolamine (EDEA), 0.1 ng/cm(2) for N-methyldiethanolamine (MDEA), and 0.1 ng/cm(2) for diethanolamine (DEA).

  2. How to determine the pressure of a methane-containing gas mixture by means of two weak Raman bands, v(3) and 2v(2)

    DEFF Research Database (Denmark)

    Hansen, Susanne Brunsgaard; Berg, Rolf W.; Stenby, Erling Halfdan

    2002-01-01

    . Surprisingly it is observed that the ratio at a fixed pressure is independent of the composition and thereby of the surroundings in which the methane molecule is vibrating. A model function to predict the pressure is given. From a practical point of view, the present results could be useful for determining...... directly the total pressure in methane mixtures the composition of which is not known.......Raman spectra of a pure CH4 sample, two CH4-C2H6 mixtures and a CH4-N2 mixture were obtained as a function of pressure at pressures up to 39.6 MPaA (MPa absolute). These spectra are presented in the region 3120-2980 cm-1. A clear pressure dependence of the area ratio between two weak methane bands...

  3. Grain boundaries in Ni3Al. 2

    International Nuclear Information System (INIS)

    Kung, H.; Sass, S.L.

    1992-01-01

    This paper discusses the dislocation structure of small angle tilt and twist boundaries in ordered Ni 3 Al, with and without boron, investigated using transmission electron microscopy. Dislocation with Burgers vectors that correspond to anti-phase boundary (APB)-coupled superpartials were found in small angle twist boundaries in both boron-free and boron-doped Ni 3 Al, and a small angle tilt boundary in boron-doped Ni 3 Al. The boundary structures are in agreement with theoretical models proposed by Marcinkowski and co-workers. The APB energy determined from the dissociation of the grain boundary dislocations was lower than values reported for isolated APBs in Ni 3 Al. For small angle twist boundaries the presence of boron reduced the APB energy at the interface until it approached zero. This is consistent with the structure of these boundaries containing small regions of increased compositional disorder in the first atomic plane next to the interface

  4. Properties of the quantum Hall effect of the two-dimensional electron gas in the n-inversion layer of InSb grain boundaries under high hydrostatic pressure

    International Nuclear Information System (INIS)

    Kraak, W.; Nachtwei, G.; Herrmann, R.; Glinski, M.

    1988-01-01

    The magnetotransport properties of the two-dimensional electron gas (2DEG) confined at the interface of the grain boundary in p-type InSb bicrystals are investigated. Under high hydrostatic pressures and in high magnetic fields (B > 5 T) the integral quantum Hall regime is reached, where the Hall resistance ρ xy is quantized to h/e 2 j (j is the number of filled Landau levels of the 2DEG). In this high field regime detailed measurements are given of the resistivity ρ xx and the Hall resistance ρ xy as function of temperature T and current density j x . An unexpected high accuracy of the Hall resistance ρ xy at magnetic field values close to a fully occupied Landau level is found, despite the high value of the diagonal resistivity ρ xx . At high current densities j x in the quantum Hall regime (j = 1) a sudden breakdown of the quantized resistance value associated with a jump-like switching to the next lower quantized value h/2e 2 is observed. A simple macroscopic picture is proposed to account for these novel transport properties associated with the quantum Hall effect. (author)

  5. Subatmospheric double containment system

    International Nuclear Information System (INIS)

    Gans, D. Jr.; Noble, J.H.

    1978-01-01

    A reinforced concrete double wall nuclear containment structure with each wall including an essentially impervious membrane or liner and porous concrete filling the annulus between the two walls is described. The interior of the structure is maintained at subatmospheric pressure, and the annulus between the two walls is maintained at a subatmospheric pressure intermediate between that of the interior and the surrounding atmospheric pressure, during normal operation. In the event of an accident within the containment structure the interior pressure may exceed atmospheric pressure, but leakage from the interior to the annulus between the double walls will not result in the pressure of the annulus exceeding atmospheric pressure so that there is no net outleakage from the containment structure

  6. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  7. Containment vessel drain system

    Science.gov (United States)

    Harris, Scott G.

    2018-01-30

    A system for draining a containment vessel may include a drain inlet located in a lower portion of the containment vessel. The containment vessel may be at least partially filled with a liquid, and the drain inlet may be located below a surface of the liquid. The system may further comprise an inlet located in an upper portion of the containment vessel. The inlet may be configured to insert pressurized gas into the containment vessel to form a pressurized region above the surface of the liquid, and the pressurized region may operate to apply a surface pressure that forces the liquid into the drain inlet. Additionally, a fluid separation device may be operatively connected to the drain inlet. The fluid separation device may be configured to separate the liquid from the pressurized gas that enters the drain inlet after the surface of the liquid falls below the drain inlet.

  8. Grain boundary corrosion of copper canister weld material

    International Nuclear Information System (INIS)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes

    2006-01-01

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository

  9. Grain boundary corrosion of copper canister weld material

    Energy Technology Data Exchange (ETDEWEB)

    Gubner, Rolf; Andersson, Urban; Linder, Mats; Nazarov, Andrej; Taxen, Claes [Corrosion and Metals Research Inst. (KIMAB), Stockholm (Sweden)

    2006-01-15

    The proposed design for a final repository for spent fuel and other long-lived residues in Sweden is based on the multi-barrier principle. The waste will be encapsulated in sealed cylindrical canisters, which will then be placed in granite bedrock and surrounded by compacted bentonite clay. The canister design is based on a thick cast inner container fitted inside a corrosion-resistant copper canister. During fabrication of the outer copper canisters there will be some unavoidable grain growth in the welded areas. As grains grow, they will tend to concentrate impurities within the copper at the new grain boundaries. The work described in this report was undertaken to determine whether there is any possibility of enhanced corrosion at grain boundaries within the copper canister, based on the recommendations of the report SKB-TR--01-09 (INIS ref. 32025363). Grain boundary corrosion of copper is not expected to be a problem for the copper canisters in a repository. However, as one step in the experimental verification it is necessary to study grain boundary corrosion of copper in an environment where it may occur. A literature study aimed to find one or several solutions that are aggressive with respect to grain boundary corrosion of copper. Copper specimens cut from welds of real copper canisters where exposed to aerated ammonium hydroxide solution for a period of 14 days at 80 degrees C and 10 bar pressure. The samples were investigated prior to exposure using the scanning Kelvin probe technique to characterize anodic and cathodic areas on the samples. The degree of corrosion was determined by optical microscopy. No grain boundary corrosion could be observed in the autoclave experiments, however, a higher rate of corrosion was observed for the weld material compared to the base material. The work suggests that grain boundary corrosion of copper weld material is most unlikely to adversely affect SKB's copper canisters under the conditions in the repository.

  10. Technology for Boundaries

    DEFF Research Database (Denmark)

    Bødker, Susanne; Kristensen, Jannie Friis; Nielsen, Christina

    2003-01-01

    .After analysing the history and the current boundary work, the paper will propose new technological support for boundary work. In particular the paper will suggest means of supporting boundaries when these are productive and for changing boundaries when this seems more appropriate. In total, flexible technologies......This paper presents a study of an organisation, which is undergoing a process transforming organisational and technological boundaries. In particular, we shall look at three kinds of boundaries: the work to maintain and change the boundary between the organisation and its customers; boundaries...... seem a core issue when dealing with technology for boundaries....

  11. County and Parish Boundaries - COUNTY_GOVERNMENT_BOUNDARIES_IDHS_IN: Governmental Boundaries Maintained by County Agencies in Indiana (Indiana Department of Homeland Security, Polygon feature class)

    Data.gov (United States)

    NSGIC State | GIS Inventory — COUNTY_GOVERNMENT_BOUNDARIES_IDHS_IN is a polygon feature class that contains governmental boundaries maintained by county agencies in Indiana, provided by personnel...

  12. Polyurethane film dressings and ceramide 2-containing hydrocolloid dressing reduce the risk of pressure ulcer development in high-risk patients undergoing surgery: a matched case-control study

    Directory of Open Access Journals (Sweden)

    Kohta M

    2015-02-01

    Full Text Available Masushi Kohta,1 Kazumi Sakamoto,2 Tsunao Oh-i31Medical Engineering Laboratory, ALCARE Co, Ltd, Sumida-ku, Tokyo, 2Department of Nursing, 3Department of Dermatology, Tokyo Medical University Ibaraki Medical Center, Ami, Ibaraki, JapanBackground: Numerous clinical challenges regarding adhesive dressings have shown that using an adhesive dressing could minimize or prevent superficial skin loss in patients at risk of developing pressure ulcers. However, evidence that polyurethane film dressings and ceramide 2-containing hydrocolloid dressing can reduce the risk of pressure ulcer development in high-risk patients undergoing surgery is limited. Therefore, we assessed the effects of application of these dressings for reducing the risk of pressure ulcer development in these patients and identified other risk factors.Methods: A matched case-control study was conducted involving 254 patients at high risk for pressure ulcer development at one acute care hospital in Japan. No patients in this study had a pressure ulcer at the start of the study. Thirty-one patients developed a pressure ulcer during surgery, and these patients were defined as cases. Controls were randomly matched for sex and age (±4 years, from which 62 patients were selected. Medical records were obtained for preoperative factors, including age, sex, body mass index, diabetes mellitus, albumin, total protein, C-reactive protein, white cell count, red cell count, and hemoglobin, and for intraoperative factors, including dressing application, operation time, body position, and surgery type. The odds ratio (OR and 95% confidence interval (CI were determined to identify risk factors for pressure ulcer development in patients undergoing surgery.Results: By multiple logistic regression analysis, there was a significantly reduced risk of pressure ulcer development for patients who had dressing applications as compared with those without dressing applications (OR 0.063; 95% CI 0.012–0.343; P=0

  13. Estimation of maximum pressure in small containments of PWR reactors due to loss of coolant accident in primary circuit; Estimativa da pressao maxima em contencoes de reatores PWR de pequeno porte devido a um acidente de perda de refrigerante no circuito primario

    Energy Technology Data Exchange (ETDEWEB)

    Mendes Neto, Teofilo [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil); Moreira, Joao Manoel Losada [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

    2000-07-01

    This work studies the problem of containment pressurization after a LOCA in reactors with small containment free volumes. The relationship between the reactor power and the containment free volume is described with the ratio between the volumes of the primary circuit and of the containment. The maximum pressure in a containment, following a LOCA, obtained after a correlation based on large containment PWR, is around 185 psia for a primary circuit and containment volumes ratio of 0.025. For the same problem, calculations with the CONTEMPT-LT code produced a maximum pressure of 162 psia. The behavior of the temperature after a LOCA to the containment, as a function of the ratio between the primary circuit and containment volume, is such that it increases reaching asymptotically to a maximum; differently, the pressure increases almost linearly with the ratio of volumes. (author)

  14. Plan on test to failure of a prestressed concrete containment vessel model

    International Nuclear Information System (INIS)

    Takumi, K.; Nonaka, A.; Umeki, K.; Nagata, K.; Soejima, M.; Yamaura, Y.; Costello, J.F.; Riesemann, W.A. von.; Parks, M.B.; Horschel, D.S.

    1992-01-01

    A summary of the plans to test a prestressed concrete containment vessel (PCCV) model to failure is provided in this paper. The test will be conducted as a part of a joint research program between the Nuclear Power Engineering Corporation (NUPEC), the United States Nuclear Regulatory Commission (NRC), and Sandia National Laboratories (SNL). The containment model will be a scaled representation of a PCCV for a pressurized water reactor (PWR). During the test, the model will be slowly pressurized internally until failure of the containment pressure boundary occurs. The objectives of the test are to measure the failure pressure, to observe the mode of failure, and to record the containment structural response up to failure. Pre- and posttest analyses will be conducted to forecast and evaluate the test results. Based on these results, a validated method for evaluating the structural behavior of an actual PWR PCCV will be developed. The concepts to design the PCCV model are also described in the paper

  15. Containment structure optimization

    International Nuclear Information System (INIS)

    Putman, S.; Walser, A.

    1979-01-01

    The major design features investigated are: dome shape, the prestress level provided to counteract accident pressure, the effect of diameter variation, and the design pressure used to size the containment. The optimum dome shape and optimum prestress level are used to investigate the effect of variations in diameter and design pressure on containment cost. The containment internal diameter is fixed at 150 feet for investigation of dome shape, prestress level and design prestress. A hemispherical dome containment with a prestress level of 1.25 P/sub a/ is recommended regardless of design pressure selected. A design pressure of 60 psi is recommended. No significant cost penalty is associated with diameter variation in the range of 145 to 155 feet

  16. Passive containment system

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1977-01-01

    Disclosed is a containment system that provides complete protection entirely by passive means for the loss of coolant accident in a nuclear power plant and wherein all stored energy released in the coolant blowdown is contained and absorbed while the nuclear fuel is prevented from over-heating by a high containment back-pressure and a reactor vessel refill system. The primary containment vessel is restored to a high sub-atmospheric pressure within a few minutes after accident initiation and the decay heat is safely transferred to the environment while radiolytic hydrogen is contained by passive means. 20 claims, 14 figures

  17. Aging of steel containments and liners in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Oland, C.B.; Ellingwood, B.; Norris, W.E.

    1998-02-01

    Aging of the containment pressure boundary in light water reactor plants is being addressed to understand the significant factors relating occurrence of corrosion efficacy of inspection and structural capacity reduction of steel containments and liners of concrete containments. and to make recommendations on use of risk models in regulatory decisions. Current regulatory in-service inspection requirements are reviewed and a summary of containment related degradation experience is presented. Current and emerging nondestructive examination techniques and a degradation assessment methodology for characterizing and quantifying the amount of damage present are described. Quantitative tools for condition assessment of aging structures using time dependent structural reliability analysis methods are summarized. Such methods provide a framework for addressing the uncertainties attendant to aging in the decision process. Results of this research provide a means for establishing current and estimating future structural capacity margins of containments, and to address the significance of incidences of reported containment degradation

  18. High-Pressure Phase Equilibria in Systems Containing CO2 and Ionic Liquid of the [Cnmim][Tf2N] Type

    OpenAIRE

    Sedláková, Z. (Zuzana); Wagner, Z. (Zdeněk)

    2012-01-01

    In this review, we present a comparison of the high-pressure phase behaviour of binary systems constituted of CO2 and ionic liquids of the [Cn(m)mim][Tf2N] type. The comparative study shows that the solubility of CO2 in ionic liquids of the [Cnmim][Tf2N] type generally increases with increasing pressure and decreasing temperature, but some peculiarities have been observed. The solubility of CO2 in ionic liquid solvents was correlated using the Soave–Redlich–Kwong equation of state. The result...

  19. Rough-wall turbulent boundary layers with constant skin friction

    KAUST Repository

    Sridhar, A.

    2017-03-28

    A semi-empirical model is presented that describes the development of a fully developed turbulent boundary layer in the presence of surface roughness with length scale ks that varies with streamwise distance x . Interest is centred on flows for which all terms of the von Kármán integral relation, including the ratio of outer velocity to friction velocity U+∞≡U∞/uτ , are streamwise constant. For Rex assumed large, use is made of a simple log-wake model of the local turbulent mean-velocity profile that contains a standard mean-velocity correction for the asymptotic fully rough regime and with assumed constant parameter values. It is then shown that, for a general power-law external velocity variation U∞∼xm , all measures of the boundary-layer thickness must be proportional to x and that the surface sand-grain roughness scale variation must be the linear form ks(x)=αx , where x is the distance from the boundary layer of zero thickness and α is a dimensionless constant. This is shown to give a two-parameter (m,α) family of solutions, for which U+∞ (or equivalently Cf ) and boundary-layer thicknesses can be simply calculated. These correspond to perfectly self-similar boundary-layer growth in the streamwise direction with similarity variable z/(αx) , where z is the wall-normal coordinate. Results from this model over a range of α are discussed for several cases, including the zero-pressure-gradient ( m=0 ) and sink-flow ( m=−1 ) boundary layers. Trends observed in the model are supported by wall-modelled large-eddy simulation of the zero-pressure-gradient case for Rex in the range 108−1010 and for four values of α . Linear streamwise growth of the displacement, momentum and nominal boundary-layer thicknesses is confirmed, while, for each α , the mean-velocity profiles and streamwise turbulent variances are found to collapse reasonably well onto z/(αx) . For given α , calculations of U+∞ obtained from large-eddy simulations are streamwise

  20. Some basic thermohydraulic calculation methods for the analysis of pressure transients in a multicompartment total containment enclosing a breached water reactor circuit

    International Nuclear Information System (INIS)

    Porter, W.H.L.

    1976-05-01

    This paper gives an appreciation and commentary of the basic calculation methods under development at AEE Winfrith for the analysis of multicompartment total containments. The assumptions introduced and the effects of their variation are important in establishing a parametric survey of the range of possible conditions which the containment may be required to meet. These aspects of the performance will be discussed as each individual factor in the train of events is examined in turn. (U.K.)

  1. Chromoproteinoids and their ability to form boundary

    International Nuclear Information System (INIS)

    Heinz, B.

    1992-01-01

    Model systems for boundary structures and cellular systems, particularly when they are a result of natural simulation experiments, are always valuable for the study of the ''Origins of Life''. Lyophilization of chromoproteinoids - peptide like molecules containing prosthetic groups - leads to the formation of boundary structures

  2. Landfills - LANDFILL_BOUNDARIES_IDEM_IN: Waste Site Boundaries in Indiana (Indiana Department of Environmental Management, Polygon Shapefile)

    Data.gov (United States)

    NSGIC State | GIS Inventory — LANDFILL_BOUNDARIES_IDEM_IN.SHP is a polygon shapefile that contains boundaries for open dump sites, approved landfills, and permitted landfills in Indiana, provided...

  3. Kuosheng BWR/6 containment safety analysis with gothic code

    International Nuclear Information System (INIS)

    Lin Ansheng; Wang Jongrong; Yuann Rueyyng; Shih Chunkuan

    2011-01-01

    Kuosheng Nuclear Power Plant in Taiwan is a GE-designed twin-unit BWR/6 plant, each unit rated at 2894 MWt. In this study, we presented the calculated results of the containment pressure and temperature responses after the main steam line break accident, which is the design basis for the containment system. During the simulation, a power of SPU range (105.1%) was used and a model of the Mark III type containment was built using the containment thermal-hydraulic program GOTHIC. The simulation consists of short and long-term responses. The drywell pressure and temperature responses which display the maximum values in the early state of the LOCA were investigated in the short-term response; the primary containment pressure and temperature responses in the long-term response. The blowdown flow was provided by FSAR and used as boundary conditions in the short-term model; in the long-term model, the blowdown flow was calculated using a GOTHIC built-in homogeneous equilibrium model. In the long-term analysis, a simplifier RPV model was employed to calculate the blowdown flow. Finally, the calculated results, similar to the FSAR results, indicate the GOTHIC code has the capability to simulate the pressure/temperature response of Mark III containment to the main steam line break LOCA. (author)

  4. Definition of containment failure

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1982-01-01

    Core meltdown accidents of the types considered in probabilistic risk assessments (PRA's) have been predicted to lead to pressures that will challenge the integrity of containment structures. Review of a number of PRA's indicates considerable variation in the predicted probability of containment failure as a function of pressure. Since the results of PRA's are sensitive to the prediction of the occurrence and the timing of containment failure, better understanding of realistic containment capabilities and a more consistent approach to the definition of containment failure pressures are required. Additionally, since the size and location of the failure can also significantly influence the prediction of reactor accident risk, further understanding of likely failure modes is required. The thresholds and modes of containment failure may not be independent

  5. Public Land Survey Township Boundaries of Iowa

    Data.gov (United States)

    Iowa State University GIS Support and Research Facility — This coverage contains polygons representing the PLSS township boundaries of the state of Iowa. TOWNSHIP was developed from a set of 99 individual county coverages...

  6. State Wildlife Management Area Boundaries - Publicly Accessible

    Data.gov (United States)

    Minnesota Department of Natural Resources — This polygon theme contains boundaries for approximately 1392 Wildlife Management Areas (WMAs) across the state covering nearly 1,288,000 acres. WMAs are part of the...

  7. Reactor container

    International Nuclear Information System (INIS)

    Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

    1979-01-01

    Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

  8. Analytical model for computing transient pressures and forces in the safety/relief valve discharge line. Mark I Containment Program, task number 7.1.2

    International Nuclear Information System (INIS)

    Wheeler, A.J.

    1978-02-01

    An analytical model is described that computes the transient pressures, velocities and forces in the safety/relief valve discharge line immediately after safety/relief valve opening. Equations of motion are defined for the gas-flow and water-flow models. Results are not only verified by comparing them with an earlier version of the model, but also with Quad Cities and Monticello plant data. The model shows reasonable agreement with the earlier model and the plant data

  9. Rigid supersymmetry with boundaries

    Energy Technology Data Exchange (ETDEWEB)

    Belyaev, D.V. [Deutsches Elektronen-Synchrotron (DESY), Hamburg (Germany); Van Nieuwenhuizen, P. [State Univ. of New York, Stony Brook, NY (United States). C.N. Yang Inst. for Theoretical Physics

    2008-01-15

    We construct rigidly supersymmetric bulk-plus-boundary actions, both in x-space and in superspace. For each standard supersymmetric bulk action a minimal supersymmetric bulk-plus-boundary action follows from an extended F- or D-term formula. Additional separately supersymmetric boundary actions can be systematically constructed using co-dimension one multiplets (boundary superfields). We also discuss the orbit of boundary conditions which follow from the Euler-Lagrange variational principle. (orig.)

  10. Improvements in the technique of vascular perfusion-fixation employing a fluorocarbon-containing perfusate and a peristaltic pump controlled by pressure feedback

    DEFF Research Database (Denmark)

    Rostgaard, J; Qvortrup, Klaus; Poulsen, Steen Seier

    1993-01-01

    A new improved technique for whole-body perfusion-fixation of rats and other small animals is described. The driving force is a peristaltic pump which is feedback regulated by a pressure transducer that monitors the blood-perfusion pressure in the left ventricle of the heart. The primary perfusate...... to cannulate the heart; the outer and inner barrels of the cannula are connected to the peristaltic pump and to the pressure transducer, respectively. The tissue oxygen tension in the rat is monitored by a subcutaneous oxygen electrode. Measurements showed that tissue hypoxia/anoxia did not develop before......-fixative is composed of a blood substitute--13.3% oxygenated fluorocarbon FC-75--in 0.05 M cacodylate buffer (pH 7.4) with a 2% glutaraldehyde. The secondary perfusate-fixative is composed of 2% glutaraldehyde in 0.05 M cacodylate buffer (pH 7.4) with 20 mM CaCl2. A double-barrelled, self-holding cannula is used...

  11. Fission reactor container

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1991-01-01

    Cooling water is sent without using dynamic equipments upon loss of coolants accident in a pressure vessel by improving an arrangement of a nuclear reactor pressure vessel. That is, a containing space is formed at the center of a suppression chamber for storing cooling water while being partitioned with each other, in which the pressure vessel is placed. Further, a water reservoir is formed above the pressure vessel. Then a water discharge pipe is connected to the reservoir for submerging the stored water over the pressure vessel upon occurrence of loss of coolants accident. Further, a water injection pipe is disposed between the pressure suppression chamber and the pressure vessel for injecting the cooling water in the pressure suppression chamber to the reactor core of the pressure vessel by the difference of a water head upon loss of coolants accident. With such a constitution, the pressure vessel has high earthquake proofness. Further, upon loss of coolants accident of the pressure vessel, the cooling water in the reservoir is discharged to submerge and cool the pressure vessel efficiently. Further, the reactor core of the pressure vessel can certainly be cooled by the cooling water of the pressure suppression chamber without relying on dynamic equipments. (I.S.)

  12. A single-center, prospective, randomized, open-label, clinical trial of ceramide 2-containing hydrocolloid dressings versus polyurethane film dressings for pressure ulcer prevention in high-risk surgical patients

    Directory of Open Access Journals (Sweden)

    Kohta M

    2015-11-01

    Full Text Available Masushi Kohta,1 Kazumi Sakamoto,2 Yasuhiro Kawachi,3 Tsunao Oh-i4 1Medical Engineering Laboratory, ALCARE Co, Ltd, Tokyo, 2Department of Nursing, 3Department of Dermatology, Tokyo Medical University Ibaraki Medical Center, Ibaraki, 4Department of Dermatology, Tokyo Medical University Hachioji Medical Center, Tokyo, Japan Purpose: There have been previous clinical studies regarding the impact of dressings on the prevention of pressure ulcer development. However, it remains unclear whether one type of dressing is better than any other type for preventing ulcer development during surgery. Therefore, we compared the effects of ceramide 2-containing hydrocolloid dressing with film dressings in high-risk patients with regard to reducing the incidence of pressure ulcer development during surgery. Patients and methods: A prospective, randomized, open-label, clinical trial was conducted involving patients who were at a high risk of developing pressure ulcers at a Japanese hospital. The intervention group received ceramide 2-containing hydrocolloid dressings (n=66, and the control group received film dressings (n=64. The primary end point was the incidence rate of pressure ulcer development in both groups; skin damage, such as blanchable erythema, skin discoloration, contact dermatitis, and stripped skin, was recorded as the secondary end point. The relative risk (RR and 95% confidence interval (CI were assessed to compare the probability ratios of pressure ulcer development between the groups. Results: There were significantly fewer patients who developed pressure ulcers in the intervention group than in the control group (RR, 0.37; 95% CI, 0.05–0.99; P=0.04. In the post hoc subgroup analysis, the superiority of the intervention group was more marked when patients had a lower body mass index (P=0.02, lower albumin values (P=0.07, and operation time of 3 hours or more and less than 6 hours (P=0.03. There was no evidence of any statistically significant

  13. Nuclear reactor container

    International Nuclear Information System (INIS)

    Ishiyama, Takenori.

    1989-01-01

    This invention concerns a nuclear reactor container in which heat is removed from a container by external water injection. Heat is removed from the container by immersing the lower portion of the container into water and scattering spary water from above. Thus, the container can be cooled by the spray water falling down along the outer wall of the container to condensate and cool vapors filled in the container upon occurrence of accidents. Further, since the inside of the container can be cooled also during usual operation, it can also serve as a dry well cooler. Accordingly, heat is removed from the reactor container upon occurrence of accidents by the automatic operation of a spray device corresponding to the change of the internal temperature and the pressure in the reactor container. Further, since all of these devices are disposed out of container, maintenance is also facilitated. (I.S.)

  14. Final Report Inspection of Aged/Degraded Containments Program.

    Energy Technology Data Exchange (ETDEWEB)

    Naus, Dan J [ORNL; Ellingwood, B R [Georgia Institute of Technology; Oland, C Barry [ORNL

    2005-09-01

    The Inspection of Aged/Degraded Containments Program had primary objectives of (1) understanding the significant factors relating corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and liners of reinforced concrete containments; (2) providing the United States Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments, and concrete containments as limited by liner integrity; (3) providing recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation; and (4) providing technical assistance to the USNRC (as requested) related to concrete technology. Primary program accomplishments have included development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of high-frequency acoustic imaging, magnetostrictive sensor, electromagnetic acoustic transducer, and multimode guided plate wave technologies for inspection of inaccessible regions of containment metallic pressure boundaries; development of a continuum damage mechanics-based approach for structural deterioration; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion. In addition, data and information assembled under this program has been transferred to the technical community through review meetings and briefings, national and international conference participation, technical committee involvement, and publications of reports and journal articles. Appendix A provides a listing of program reports, papers, and publications; and Appendix B contains a listing of

  15. Asymptotic analysis and boundary layers

    CERN Document Server

    Cousteix, Jean

    2007-01-01

    This book presents a new method of asymptotic analysis of boundary-layer problems, the Successive Complementary Expansion Method (SCEM). The first part is devoted to a general comprehensive presentation of the tools of asymptotic analysis. It gives the keys to understand a boundary-layer problem and explains the methods to construct an approximation. The second part is devoted to SCEM and its applications in fluid mechanics, including external and internal flows. The advantages of SCEM are discussed in comparison with the standard Method of Matched Asymptotic Expansions. In particular, for the first time, the theory of Interactive Boundary Layer is fully justified. With its chapter summaries, detailed derivations of results, discussed examples and fully worked out problems and solutions, the book is self-contained. It is written on a mathematical level accessible to graduate and post-graduate students of engineering and physics with a good knowledge in fluid mechanics. Researchers and practitioners will estee...

  16. Hydrogen storage container

    Science.gov (United States)

    Wang, Jy-An John; Feng, Zhili; Zhang, Wei

    2017-02-07

    An apparatus and system is described for storing high-pressure fluids such as hydrogen. An inner tank and pre-stressed concrete pressure vessel share the structural and/or pressure load on the inner tank. The system and apparatus provide a high performance and low cost container while mitigating hydrogen embrittlement of the metal tank. System is useful for distributing hydrogen to a power grid or to a vehicle refueling station.

  17. Waste container and method for containing waste

    International Nuclear Information System (INIS)

    Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

    1990-01-01

    In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

  18. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  19. Methodology of a PWR containment analysis during a thermal-hydraulic accident

    International Nuclear Information System (INIS)

    Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

    2015-01-01

    The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

  20. Containment performance improvement program

    International Nuclear Information System (INIS)

    Beckner, W.; Mitchell, J.; Soffer, L.; Chow, E.; Lane, J.; Ridgely, J.

    1990-01-01

    The Containment Performance Improvement (CPI) program has been one of the main elements in the US Nuclear Regulatory Commission's (NRC's) integrated approach to closure of severe accident issues for US nuclear power plants. During the course of the program, results from various probabilistic risk assessment (PRA) studies and from severe accident research programs for the five US containment types have been examined to identify significant containment challenges and to evaluate potential improvements. The five containment types considered are: the boiling water reactor (BMR) Mark I containment, the BWR Mark II containment, the BWR Mark III containment, the pressurized water reactor (PWR) ice condenser containment, and the PWR dry containments (including both subatmospheric and large subtypes). The focus of the CPI program has been containment performance and accident mitigation, however, insights are also being obtained in the areas of accident prevention and accident management