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

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

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

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.

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.

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.

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

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.

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

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...

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

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.

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

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.)

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)

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....

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.

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...

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.)

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.)

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)

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

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)

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)

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

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

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.

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.

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

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

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

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

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

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)

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

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

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.

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

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)

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

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...

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

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.

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

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.

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

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

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

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.

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)

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.)

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

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)

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

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.)

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

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

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)

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)

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

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)

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.)

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

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.

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.

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.)

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

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

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.)

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.)

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.

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)

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

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.

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

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.

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)

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.

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)

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)

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

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)

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.

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

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

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

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

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)

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.

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....

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

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.

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.

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

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)

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

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

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)

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

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.

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.

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

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)

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

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

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.

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.)

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

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

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.)

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

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

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

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

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)

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...

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)

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

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

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%

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.

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)

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.

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.

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.

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.

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

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)

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.

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.

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)

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)

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)

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.

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.

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.

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…

Brazing of Sealing for Instrumentation Feed through of high Pressure Vessel

International Nuclear Information System (INIS)

Jeong, H. Y.; Ahn, S. H.; Joung, C. Y.; Lee, J. M.; Lee, C. Y.

2011-01-01

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP's operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo-hydraulic tests. It is composed of an In-Pile test Section(IPS) and an Out- Pile System(OPS). The OPS contains a pressurizer, cooler, pump, heater and purification system which are necessary to maintain the proper fluid conditions. In addition, the OPS contains engineered safety systems that could safely shutdown both HANARO and FTL if an accident occurs. The IPS accommodating fuel pins has loaded IP-1 hole in HANARO has a double pressure vessel for the design conditions of 350 .deg. C, 17.5MPa and is composed of outer assembly and inner assembly. It has instruments such as a thermocouple, LVDT and SPND to measure the fuel performances during the test. FTL coolant is supplied to the IPS at the core of commercial nuclear power plants and the same temperature, pressure and flow conditions. Sensors installed on the inside of IPS to send a signal transmission MI-Cables to the outside for instrumentation is through the pressure boundary. Therefore, pressure boundary should be maintained in the sealing performance. Brazing is typically lower than the melting point of material without melting the material almost would be like welding when it is necessary to use. It is commonly used to use BAg(ASME II SFA-5.8 UNS-P07563) filler metal, but corrosion occurs containing a large quantity of copper in Bag, and when contact with the coolant, the coolant water quality is influenced. Therefore, using BNi-2(ASME II SFA-5.8 UNS-N99620) filler metal is considered. Brazing at the Sealing Plug in the top of IPS was considered for Mi-cable's integrity and to maintain the pressure boundary. After brazing is performed, brazing the Mi-cable integrity and pressure boundary sealing performance was tested

Brazing of Sealing for Instrumentation Feed through of high Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Jeong, H. Y.; Ahn, S. H.; Joung, C. Y.; Lee, J. M.; Lee, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Fuel Test Loop(FTL) is a facility which could conduct a fuel irradiation test at HANARO(High-flux Advanced Neutron Application Reactor). FTL simulates commercial NPP's operating conditions such as the pressure, temperature and neutron flux levels to conduct the irradiation and thermo-hydraulic tests. It is composed of an In-Pile test Section(IPS) and an Out- Pile System(OPS). The OPS contains a pressurizer, cooler, pump, heater and purification system which are necessary to maintain the proper fluid conditions. In addition, the OPS contains engineered safety systems that could safely shutdown both HANARO and FTL if an accident occurs. The IPS accommodating fuel pins has loaded IP-1 hole in HANARO has a double pressure vessel for the design conditions of 350 .deg. C, 17.5MPa and is composed of outer assembly and inner assembly. It has instruments such as a thermocouple, LVDT and SPND to measure the fuel performances during the test. FTL coolant is supplied to the IPS at the core of commercial nuclear power plants and the same temperature, pressure and flow conditions. Sensors installed on the inside of IPS to send a signal transmission MI-Cables to the outside for instrumentation is through the pressure boundary. Therefore, pressure boundary should be maintained in the sealing performance. Brazing is typically lower than the melting point of material without melting the material almost would be like welding when it is necessary to use. It is commonly used to use BAg(ASME II SFA-5.8 UNS-P07563) filler metal, but corrosion occurs containing a large quantity of copper in Bag, and when contact with the coolant, the coolant water quality is influenced. Therefore, using BNi-2(ASME II SFA-5.8 UNS-N99620) filler metal is considered. Brazing at the Sealing Plug in the top of IPS was considered for Mi-cable's integrity and to maintain the pressure boundary. After brazing is performed, brazing the Mi-cable integrity and pressure boundary sealing performance was

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

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

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.

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

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

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)

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

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...

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...

Modelling of film condensation on the reactor containment walls

International Nuclear Information System (INIS)

Leduc, Christian

1995-01-01

A containment code used in nuclear plant safety analysis must be able to predict evolutions of steam, air and hydrogen concentrations and pressure in the containment of a pressurized water reactor in an accidental situation. Steam condensation on cold walls is an essential factor for these evolutions as it allows the release of an important heat flow, and locally reduces steam concentration. In this research thesis, the author proposes a film condensation model in presence of un-condensable gases. The film flow is supposed to be laminar. Three different approaches are used to model transfers in boundary layers: global correlations in which a hybrid Grashof number is used which expresses the mass and thermal nature of convection, a boundary layer calculation using wall rules for a forced convection regime, and a boundary layer calculation using a k-epsilon model with a low Reynolds number for a natural convection regime. Each approach requires very different mesh fineness at the vicinity of the wall. Models are implemented in the 3-D TRIO-VF thermo-hydraulic code. The obtained theoretical heat transfer coefficients are compared with experimental results [fr

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

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.)

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.

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

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.

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

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.

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

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)

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

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.

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.

The pressure exerted by a confined ideal gas

International Nuclear Information System (INIS)

Pang Hai; Dai Wusheng; Xie Mi

2011-01-01

In this paper, we study the pressure exerted by a confined ideal gas on the container boundary and we introduce a surface force in gases. First, the general expression for the local surface pressure tensor is obtained. We find, by examples, that the pressure vanishes at the edges of a box, peaks at the middle of the surface and its magnitude for different statistics satisfies p Fermi > p classical > p Bose on every boundary point. Then, the relation between the surface pressure tensor and generalized forces is studied. Based on the relation, we find that a confined ideal gas can exert forces whose effect is to reduce the total surface area of the boundary of an incompressible object. The force provides mechanisms for several mechanical effects. (1) The force contributes to the adhesion of two thin films in contact with each other. We derive an expression for the adhesion force between two square sheets, estimate its magnitude, and also give a method for distinguishing it from other adhesion forces. (2) The force can lead to the recoiling of a DNA-like column. We study the recoiling process using a simple model and find a deviation from the result given in the thermodynamic limit, which is in accordance with experiments. (3) An open container immersed in a gas can be compressed by this force like the Casimir effect. We discuss the effect for various geometries. (paper)

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.

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.)

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

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.)

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

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

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.)

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.)

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)

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

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

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.)

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.

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

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.

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.

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

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

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

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.

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.

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

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)

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

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

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.

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...

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

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

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

Experiments to evaluate behavior of containment piping bellows under severe accident conditions

International Nuclear Information System (INIS)

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

1993-01-01

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature, and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, New Mexico. Several different bellows geometries, representative of actual containment bellows, are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen tests have been conducted. The tests showed that withstanding relatively large bellows are capable of deformations, up to, or near, the point of full compression before developing leakage. The test data is presented and discussed

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

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)

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)

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.)

Pressure-driven brine migration in a salt repository

International Nuclear Information System (INIS)

Hwang, Y.; Chambre, P.L.; Pigford, T.H.; Lee, W.W.L.

1989-01-01

The traditional view is that salt is the ideal rock for isolation of nuclear waste because it is ''dry'' and probably ''impermeable.'' The existence of salt through geologic time is prima facie evidence of such properties. Experiments and experience at potential salt sites for geologic repositories have indicated that while porosity and permeability of salt are low, the salt may be saturated with brine. If this hypothesis is correct, then it is possible to have brine flow due to pressure differences within the salt. If there is pressure-driven brine migration in salt repositories then it is paramount to know the magnitude of such flow because inward brine flow would affect the corrosion rate of nuclear waste containers and outward brine flow might affect radionuclide transport rates. Brine exists in natural salt as inclusions in salt crystals and in grain boundaries. Brine inclusions in crystals move to nearby grain boundaries when subjected to a temperature gradient, because of temperature-dependent solubility of salt. Brine in grain boundaries moves under the influence of a pressure gradient. When salt is mined to create a waste repository, brine from grain boundaries will migrate into the rooms, tunnels and boreholes because these cavities are at atmospheric pressure. After a heat-emitting waste package is emplaced and backfilled, the heat will impose a temperature gradient in the surrounding salt that will cause inclusions in the nearby salt to migrate to grain boundaries within a few years, adding to the brine that was already present in the grain boundaries. The formulation of brine movement with salt as a thermoelastic porous medium, in the context of the continuum theory of mixtures, has been described. In this report we show the mathematical details and discuss the results predicted by this analysis

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.

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

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)

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.

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.

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

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.

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

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

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.

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)

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

The development of a Type B sample container

International Nuclear Information System (INIS)

Glass, R.E.

1993-01-01

Sandia National Laboratories is developing a package to support chemical agent sampling for the multilateral Chemical Weapons Convention. The package is designed to prevent the release of lethal chemical agents during international transport of chemical agents. The package is being designed to meet the IAEA requirements for Type B container. The configuration of the packaging working from the exterior to the interior is as follows. The outer shell provides a sacrificial boundary which will provide protection against the thermal and structural assaults of the hypothetical accident sequence. This shell provides all of the lifting and tie-down attachments. The closure is provided with a v-clamp. The cylindrical shell is austenitic stainless steel with standard pressure vessel heads. Internal to this shell is approximately 7 cm of ceramic fiber insulation to provide protection for the containment boundary against the all-engulfing fire. The containment vessel consists of a stainless steel cylindrical shell with pressure vessel heads at each end. The closure includes an o-ring test port to sample between an elastomeric double o-ring seal. The interior of the package can hold various teflon inserts which are machined to accept samples. The package has a mass of 35 kg and external dimension of 33 cm in length and 30 cm in diameter. The internal cavity is 10 cm in length and 10 cm in diameter. An insert can be machined to accept multiple samples of any configuration within that envelope. This paper describes the design and testing of the Type B sample container. (author)

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

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)

Feasibility of a tracer gas technique for containment leakage characterization at Bruce NGS

International Nuclear Information System (INIS)

Singh, V.P.

1985-11-01

Methods for tracer gas test have been conceived and are proposed for use in conjunction with other techniques used during off-power pressurization tests. During pressurization tests is appears possible to quantify leaks through containment boundaries which make up one of the walls in adjacent rooms but quantification of leaks to open areas will require further development. Several gases may be used as tracers during pressurization tests but the preferred tracer gas is sulphur hexafluoride (SF 6 ) at an in-vault concentration of 100 μL/L if open area sampling is to be carried out of 10 μL/L if only closed room sampling is to be performed. Large values of the ratio (tracer gas concentration in containment/lower detection limit) are necessary for identification of leak sites in open areas having significant ventilation flow. It is recommended that in-station trials be carried out to test the validity of this technique. In addition, a tracer gas technique for use during on-power operation is also proposed but leak site identification and quantification during on-power tests is only possible for containment boundaries which make up the wall(s) of adjacent rooms. The use of SF 6 is required for tests conducted during on-power operation. The recommended in-vault concentration is 10 μL/L. Recommendations are made for future work, including leak tests during on-power operation

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.

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)

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

EIT-based fabric pressure sensing.

Science.gov (United States)

Yao, A; Yang, C L; Seo, J K; Soleimani, M

2013-01-01

This paper presents EIT-based fabric sensors that aim to provide a pressure mapping using the current carrying and voltage sensing electrodes attached to the boundary of the fabric patch. Pressure-induced shape change over the sensor area makes a change in the conductivity distribution which can be conveyed to the change of boundary current-voltage data. This boundary data is obtained through electrode measurements in EIT system. The corresponding inverse problem is to reconstruct the pressure and deformation map from the relationship between the applied current and the measured voltage on the fabric boundary. Taking advantage of EIT in providing dynamical images of conductivity changes due to pressure induced shape change, the pressure map can be estimated. In this paper, the EIT-based fabric sensor was presented for circular and rectangular sensor geometry. A stretch sensitive fabric was used in circular sensor with 16 electrodes and a pressure sensitive fabric was used in a rectangular sensor with 32 electrodes. A preliminary human test was carried out with the rectangular sensor for foot pressure mapping showing promising results.

EIT-Based Fabric Pressure Sensing

Directory of Open Access Journals (Sweden)

A. Yao

2013-01-01

Full Text Available This paper presents EIT-based fabric sensors that aim to provide a pressure mapping using the current carrying and voltage sensing electrodes attached to the boundary of the fabric patch. Pressure-induced shape change over the sensor area makes a change in the conductivity distribution which can be conveyed to the change of boundary current-voltage data. This boundary data is obtained through electrode measurements in EIT system. The corresponding inverse problem is to reconstruct the pressure and deformation map from the relationship between the applied current and the measured voltage on the fabric boundary. Taking advantage of EIT in providing dynamical images of conductivity changes due to pressure induced shape change, the pressure map can be estimated. In this paper, the EIT-based fabric sensor was presented for circular and rectangular sensor geometry. A stretch sensitive fabric was used in circular sensor with 16 electrodes and a pressure sensitive fabric was used in a rectangular sensor with 32 electrodes. A preliminary human test was carried out with the rectangular sensor for foot pressure mapping showing promising results.

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...

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

Beta limitation of matter-antimatter boundary layers

International Nuclear Information System (INIS)

Lehnert, B.

1987-08-01

A model has earlier been proposed for a boundary layer which separates a cloud of matter from one of antimatter in a magnetized ambiplasma. In this model steady pressure equilibrium ceases to exist when a certain beta limit is exceeded. The latter is defined as the ratio between the ambiplasma and magnetic field pressures which balance each other in the boundary layer. Thus, at an increasing density, the high-energy particles created by annihilation within the layer are 'pumped up' to a pressure which cannot be balanced by a given magnetic field. The boundary layer then 'disrupts'. The critical beta limit thus obtained falls within the observed parameter ranges of galaxies and other large cosmical objects. Provided that the considered matter-antimatter balance holds true, this limit is thus expected to impose certain existence conditions on matter-antimatter boundary layers. Such a limitation may apply to certain cosmical objects and cosmological models. The maximum time scale for the corresponding disruption development has been estimated to be in the range from about 10 -4 to 10 2 seconds for boundary layers at ambiplasma particle densities in the range from 10 4 to 10 -2 m -3 , respectively. (author)

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.)

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.

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.

The linearized pressure Poisson equation for global instability analysis of incompressible flows

Science.gov (United States)

Theofilis, Vassilis

2017-12-01

The linearized pressure Poisson equation (LPPE) is used in two and three spatial dimensions in the respective matrix-forming solution of the BiGlobal and TriGlobal eigenvalue problem in primitive variables on collocated grids. It provides a disturbance pressure boundary condition which is compatible with the recovery of perturbation velocity components that satisfy exactly the linearized continuity equation. The LPPE is employed to analyze instability in wall-bounded flows and in the prototype open Blasius boundary layer flow. In the closed flows, excellent agreement is shown between results of the LPPE and those of global linear instability analyses based on the time-stepping nektar++, Semtex and nek5000 codes, as well as with those obtained from the FreeFEM++ matrix-forming code. In the flat plate boundary layer, solutions extracted from the two-dimensional LPPE eigenvector at constant streamwise locations are found to be in very good agreement with profiles delivered by the NOLOT/PSE space marching code. Benchmark eigenvalue data are provided in all flows analyzed. The performance of the LPPE is seen to be superior to that of the commonly used pressure compatibility (PC) boundary condition: at any given resolution, the discrete part of the LPPE eigenspectrum contains converged and not converged, but physically correct, eigenvalues. By contrast, the PC boundary closure delivers some of the LPPE eigenvalues and, in addition, physically wrong eigenmodes. It is concluded that the LPPE should be used in place of the PC pressure boundary closure, when BiGlobal or TriGlobal eigenvalue problems are solved in primitive variables by the matrix-forming approach on collocated grids.

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.)

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

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)

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...

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

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

Endurance test report of rubber sealing materials for the containment vessel

International Nuclear Information System (INIS)

Yamamoto, R.; Watanabe, K.; Hanashima, K.

2015-01-01

In the event of a nuclear power plant accident such as a core meltdown and a cooling system failure, the containment contains radioactive materials released from the reactor pressure vessel to reduce the activity of the radioactive materials and the effects of radiation in the vicinity of the plant. Since high sealing performance and high pressure resistance are required of the containment, a silicone or EPDM rubber gasket with high heat and radiation resistance is used for the sealing of the sealing boundary of the containment. In recent years, it has been shown that a large amount of steam is released into the containment in the case of a severe accident. Consequently, radiation resistance at high temperature as well as steam resistance is required of the rubber gasket placed at the sealing boundary. However, the steam resistance of silicone rubber is not necessarily as good as that of EPDM rubber. Therefore, it is necessary to evaluate the sealing characteristics of rubber gaskets in such a degrading environment in a severe accident. O. Kato et al. [1] conducted a study on the degradation status of rubber gaskets and their application limits at high temperature. However, few studies have evaluated rubber gaskets in high-temperature radiation and steam environments. In this study, we degraded silicone rubber and EPDM rubber used for the containment in the high-temperature radiation and steam environments expected to occur in a severe accident and evaluated the useful life of the rubber as a sealing material by estimating the change in its performance as a sealing material from the change in permanent compressive strain in the rubber. (author)

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...

Experimental study on effects of inlet boundary layer thickness and boundary layer fence in a turbine cascade

International Nuclear Information System (INIS)

Jun, Y. M.; Chung, J. T.

2000-01-01

The working fluid from the combustor to the turbine stage of a gas turbine makes various boundary layer thickness. Since the inlet boundary layer thickness is one of the important factors that affect the turbine efficiency, It is necessary to investigate secondary flow and loss with various boundary layer thickness conditions. In the present study, the effect of various inlet boundary layer thickness on secondary flow and loss and the proper height of the boundary layer fences for various boundary layer thickness were investigated. Measurements of secondary flow velocity and total pressure loss within and downstream of the passage were taken under 5 boundary layer thickness conditions, 16, 36, 52, 69, 110mm. It was found that total pressure loss and secondary flow areas were increased with increase of thickness but they were maintained almost at the same position. At the following research about the boundary layer fences, 1/6, 1/3, 1/2 of each inlet boundary layer thickness and 12mm were used as the fence heights. As a result, it was observed that the proper height of the fences was generally constant since the passage vortex remained almost at the same position. Therefore once the geometry of a cascade is decided, the location of the passage vortex and the proper fence height are appeared to be determined at the same time. When the inlet boundary layer thickness is relatively small, the loss caused by the proper fence becomes bigger than end wall loss so that it dominates secondary loss. In these cases the proper fence height is decided not by the cascade geometry but by the inlet boundary layer thickness as previous investigations

Pressurized-water reactors

International Nuclear Information System (INIS)

Bush, S.H.

1983-03-01

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

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.

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

Structural features and in-service inspection of the LTHR-200 pressure vessel

International Nuclear Information System (INIS)

Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

1993-01-01

LTHR-200 is a low temperature district-heating reactor. It adopts double-shell design pressure vessel and metal containment. Because of the safety and structural features of the reactor, the in-service inspection of the pressure vessel can be simplified greatly. LTHR-200 is an integrated arrangement. Both its core components and the main heat exchangers are contained in the reactor pressure vessel. The coolant of the main loop is run by a full-power natural circulation and there need no main pumps and pipes. Thus, the reactor pressure vessel constitutes the pressure boundary of the reactor's main loop coolant. In regard to these features, a small-sized containment is designed for the reactor. The metal safety container with a small volume is placed closely around the reactor pressure vessel. Outside the metal containment, there is a large reinforced concrete construction for the reactor. Their main operation and design parameters are as follows: The pressure vessel: operation pressure = 2.4 MPa; design pressure = 3.0 MPa; design temperature = 250 deg C; 40 year fast neutron (E>1MeV) fluence in the belt-line region = < 10E16n/cm; internal diameter = 5000 mm; material SA516-70; shell thickness 65 mm; The metal containment: maximum operation pressure = 1.8 MPa; design pressure = 1.8 MPa; design temperature = 250 deg. C; upper internal diameter 7000 mm; lower internal diameter = 5600 mm; material = SA516-70; shell thickness, upper part = 80 mm; lower part = 50 mm. All penetrating pipes through the pressure vessel are located at the top penetration section of the shell. All the internal diameters of penetrating pipes are less than 50 mm. Inside and outside the metal containment wall respectively, isolating valves are connected to the reactor coolant pipe which passes through the containment. These two isolating valves use different driving methods. Every penetrating part of the reactor construction uses a proper form of structure according to safety requirements

Reinforced concrete containment structures in high seismic zones

International Nuclear Information System (INIS)

Aziz, T.S.

1977-01-01

A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxiliary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embedded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. (Auth.)

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

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.

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)

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.

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

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)

A study of mechanical sealing methods using graphite powder for high pressure vessel

Energy Technology Data Exchange (ETDEWEB)

Jeong, H. Y.; Hong, J. T.; Ahn, S. H.; Joung, C. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The Fuel Test Loop (FTL) is a facility that can conduct fuel irradiation tests at the HANARO (High flux Advanced Neutron Application Reactor). The FTL simulates commercial NPP operating conditions such as pressure, temperature and neutron flux levels to conduct irradiation and thermo hydraulic tests. It is composed of an In Pile test Section (IPS) and an Out Pile System (OPS). The OPS contains a pressurizer, cooler, pump, heater and purification system, which are necessary to maintain the proper fluid conditions. In addition, the OPS contains engineered safety systems that can safely shutdown both HANARO and FTL if an accident occurs. The IPS accommodating fuel pins has a loaded IP 1 hole in HANARO, and a double pressure vessel for the design conditions of 350 .deg. C, 17.5MPa and is composed of an outer assembly and inner assembly. It has instruments such as a thermocouple, LVDT and SPND to measure the fuel performances during the test. FTL coolant is supplied to the IPS at the core of commercial nuclear power plants at the same temperature, pressure and flow conditions. Sensors are installed on the inside of the IPS to send signal transmission MI Cables to the outside for instrumentation through the pressure boundary. Therefore, the pressure boundary should be maintained in the sealing performance. Currently, the sealing of the IPS of the the FTL is maintained through a brazing method. However, A brazing method has disadvantages that can occur owing to thermal deformation or breakage in the instrumentation Mi cable. IPS inner assembly is a very long design length (approximately 5.29m), so it is difficult to perform in a vacuum chamber. Therefore, an easy and reliable way to assemble the instrumentation Mi cable mechanical sealing method has been studied. In this study, criteria tests at the pressure boundary were performed using universally applicable graphite powder for the instrumentation MI cable of various sizes.

A study of mechanical sealing methods using graphite powder for high pressure vessel

International Nuclear Information System (INIS)

Jeong, H. Y.; Hong, J. T.; Ahn, S. H.; Joung, C. Y.

2012-01-01

The Fuel Test Loop (FTL) is a facility that can conduct fuel irradiation tests at the HANARO (High flux Advanced Neutron Application Reactor). The FTL simulates commercial NPP operating conditions such as pressure, temperature and neutron flux levels to conduct irradiation and thermo hydraulic tests. It is composed of an In Pile test Section (IPS) and an Out Pile System (OPS). The OPS contains a pressurizer, cooler, pump, heater and purification system, which are necessary to maintain the proper fluid conditions. In addition, the OPS contains engineered safety systems that can safely shutdown both HANARO and FTL if an accident occurs. The IPS accommodating fuel pins has a loaded IP 1 hole in HANARO, and a double pressure vessel for the design conditions of 350 .deg. C, 17.5MPa and is composed of an outer assembly and inner assembly. It has instruments such as a thermocouple, LVDT and SPND to measure the fuel performances during the test. FTL coolant is supplied to the IPS at the core of commercial nuclear power plants at the same temperature, pressure and flow conditions. Sensors are installed on the inside of the IPS to send signal transmission MI Cables to the outside for instrumentation through the pressure boundary. Therefore, the pressure boundary should be maintained in the sealing performance. Currently, the sealing of the IPS of the the FTL is maintained through a brazing method. However, A brazing method has disadvantages that can occur owing to thermal deformation or breakage in the instrumentation Mi cable. IPS inner assembly is a very long design length (approximately 5.29m), so it is difficult to perform in a vacuum chamber. Therefore, an easy and reliable way to assemble the instrumentation Mi cable mechanical sealing method has been studied. In this study, criteria tests at the pressure boundary were performed using universally applicable graphite powder for the instrumentation MI cable of various sizes

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.

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)

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.)

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

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)

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.

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.

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

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.)

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....

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.

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...

Leak before break behaviour of austenitic and ferritic pipes containing circumferential defects

Energy Technology Data Exchange (ETDEWEB)

Stadtmueller, W.; Sturm, D.

1997-04-01

Several research projects carried out at MPA Stuttgart to investigate the Leak-before-Break (LBB) behavior of safety relevant pressure bearing components are summarized. Results presented relate to pipes containing circumferential defects subjected to internal pressure and external bending loading. An overview of the experimentally determined results for ferritic components is presented. For components containing postulated or actual defects, the dependence of the critical loading limit on the defect size is shown in the form of LBB curves. These are determined experimentally and/or by calculation for through-wall slits, and represent the boundary curve between leakage and massive fracture. For surface defects and a given bending moment and internal pressure, no fracture will occur if the length at leakage remains smaller than the critical defect length given by the LBB curve for through-wall defects. The predictive capability of engineering calculational methods are presented by way of example. The investigation programs currently underway, testing techniques, and initial results are outlined.

Energy principle with included boundary conditions

International Nuclear Information System (INIS)

Lehnert, B.

1994-01-01

Earlier comments by the author on the limitations of the classical form of the extended energy principle are supported by a complementary analysis on the potential energy change arising from free-boundary displacements of a magnetically confined plasma. In the final formulation of the extended principle, restricted displacements, satisfying pressure continuity by means of plasma volume currents in a thin boundary layer, are replaced by unrestricted (arbitrary) displacements which can give rise to induced surface currents. It is found that these currents contribute to the change in potential energy, and that their contribution is not taken into account by such a formulation. A general expression is further given for surface currents induced by arbitrary displacements. The expression is used to reformulate the energy principle for the class of displacements which satisfy all necessary boundary conditions, including that of the pressure balance. This makes a minimization procedure of the potential energy possible, for the class of all physically relevant test functions which include the constraints imposed by the boundary conditions. Such a procedure is also consistent with a corresponding variational calculus. (Author)

Direct torus venting analysis for Chinshan BWR-4 plant with MARK-I containment

Energy Technology Data Exchange (ETDEWEB)

Yuann, Yng-Ruey, E-mail: ryyuann@iner.gov.tw

2017-03-15

Highlights: • Study the effectiveness of Direct Torus Venting System (DTVS) during extended SBO of 24 h for Chinshan MARK-I plant. • Containment response is analyzed by GOTHIC based on boundary conditions from RETRAN calculation. • Analyses are performed with and without DTVS, respectively. • Suppression pool is sub-divided and thermal stratification is observed. - Abstract: The Chinshan plant, owned by Taiwan Power Company, has twin units of BWR-4 reactor and MARK-I containment. Both units have been operating at rated core thermal power of 1840 MWt. The existing Direct Torus Venting System (DTVS) is the main system used for venting the containment during the extended station blackout event. The purpose of this paper is to study the effects of the DTVS venting on the response of the containment pressure and temperature. The reactor is depressurized by manually opening the safety relief valves (SRVs) during the SBO, which causes the mass and energy to be discharged into and heat up the suppression pool. The RETRAN model is used to calculate the Nuclear Steam Supply System (NSSS) response and generate the SRV blowdown conditions, including SRV pressure, enthalpy, and mass flow rate. These conditions are then used as the time-dependent boundary conditions for the GOTHIC code to calculate the containment pressure and temperature response. The DTVS model is established in the GOTHIC model based on the venting size, venting piping loss, venting initiation time, and venting source. The lumped volume model, 1-D coarse-mesh model, and 3-D coarse-mesh model are considered in the torus volume. The calculation is first done without DTVS venting to establish a reference basis. Then a case with DTVS available is performed. Comparison of the two cases shows that the existing DTVS design is effective in mitigating the severity of the containment pressure and temperature transients. The results also show that the 1-D coarse-mesh model may not be appropriate since a

A new boundary control scheme for simultaneous achievement of H-mode and radiative cooling (SHC boundary)

International Nuclear Information System (INIS)

Ohyabu, N.

1995-05-01

We have proposed a new boundary control scheme (SHC boundary), which could allow simultaneous achievement of the H-mode type confinement improvement and radiative cooling with wide heat flux distribution. In our proposed configuration, a low m island layer sharply separates a plasma confining region from an open 'ergodic' boundary. The degree of openness in the ergodic boundary must be high enough to make the plasma pressure constant along the field line, which in turn separates low density plasma just outside the plasma confining region (the key external condition for achieving a good H-mode discharge) from very high density, cold radiative plasma near the wall (required for effective edge radiative cooling). Examples of such proposed SHC boundaries for Heliotron typed devices and tokamaks are presented. (author)

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.)

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)

Investigation of turbulence models with compressibility corrections for hypersonic boundary flows

Directory of Open Access Journals (Sweden)

Han Tang

2015-12-01

Full Text Available The applications of pressure work, pressure-dilatation, and dilatation-dissipation (Sarkar, Zeman, and Wilcox models to hypersonic boundary flows are investigated. The flat plate boundary layer flows of Mach number 5–11 and shock wave/boundary layer interactions of compression corners are simulated numerically. For the flat plate boundary layer flows, original turbulence models overestimate the heat flux with Mach number high up to 10, and compressibility corrections applied to turbulence models lead to a decrease in friction coefficients and heating rates. The pressure work and pressure-dilatation models yield the better results. Among the three dilatation-dissipation models, Sarkar and Wilcox corrections present larger deviations from the experiment measurement, while Zeman correction can achieve acceptable results. For hypersonic compression corner flows, due to the evident increase of turbulence Mach number in separation zone, compressibility corrections make the separation areas larger, thus cannot improve the accuracy of calculated results. It is unreasonable that compressibility corrections take effect in separation zone. Density-corrected model by Catris and Aupoix is suitable for shock wave/boundary layer interaction flows which can improve the simulation accuracy of the peak heating and have a little influence on separation zone.

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

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

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...

Container materials in environments of corroded spent nuclear fuel

Science.gov (United States)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

Homogenization of the stochastic Navier–Stokes equation with a stochastic slip boundary condition

KAUST Repository

Bessaih, Hakima

2015-11-02

The two-dimensional Navier–Stokes equation in a perforated domain with a dynamical slip boundary condition is considered. We assume that the dynamic is driven by a stochastic perturbation on the interior of the domain and another stochastic perturbation on the boundaries of the holes. We consider a scaling (ᵋ for the viscosity and 1 for the density) that will lead to a time-dependent limit problem. However, the noncritical scaling (ᵋ, β > 1) is considered in front of the nonlinear term. The homogenized system in the limit is obtained as a Darcy’s law with memory with two permeabilities and an extra term that is due to the stochastic perturbation on the boundary of the holes. The nonhomogeneity on the boundary contains a stochastic part that yields in the limit an additional term in the Darcy’s law. We use the two-scale convergence method after extending the solution with 0 inside the holes to pass to the limit. By Itô stochastic calculus, we get uniform estimates on the solution in appropriate spaces. Due to the stochastic integral, the pressure that appears in the variational formulation does not have enough regularity in time. This fact made us rely only on the variational formulation for the passage to the limit on the solution. We obtain a variational formulation for the limit that is solution of a Stokes system with two pressures. This two-scale limit gives rise to three cell problems, two of them give the permeabilities while the third one gives an extra term in the Darcy’s law due to the stochastic perturbation on the boundary of the holes.

Ultrasound enhanced plasma surface modification at atmospheric pressure

DEFF Research Database (Denmark)

Kusano, Yukihiro; Singh, Shailendra Vikram; Norrman, Kion

and the material surface, and thus many reactive species generated in the plasma can reach the surface before inactivated, and be efficiently utilized for surface modification. In the present work polyester plates are treated using a dielectric barrier discharge (DBD) and a gliding arc at atmospheric pressure......Atmospheric pressure plasma treatment can be highly enhanced by simultaneous high-power ultrasonic irradiation onto the treating surface. It is because ultrasonic waves with a sound pressure level (SPL) above approximately 140 dB can reduce the thickness of a boundary gas layer between the plasma...... irradiation, the water contact angle dropped markedly, and tended to decrease furthermore at higher power. The ultrasonic irradiation during the plasma treatment consistently improved the wettability. Oxygen containing polar functional groups were introduced at the surface by the plasma treatment...

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)

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

Robust boundary treatment for open-channel flows in divergence-free incompressible SPH

Science.gov (United States)

Pahar, Gourabananda; Dhar, Anirban

2017-03-01

A robust Incompressible Smoothed Particle Hydrodynamics (ISPH) framework is developed to simulate specified inflow and outflow boundary conditions for open-channel flow. Being purely divergence-free, the framework offers smoothed and structured pressure distribution. An implicit treatment of Pressure Poison Equation and Dirichlet boundary condition is applied on free-surface to minimize error in velocity-divergence. Beyond inflow and outflow threshold, multiple layers of dummy particles are created according to specified boundary condition. Inflow boundary acts as a soluble wave-maker. Fluid particles beyond outflow threshold are removed and replaced with dummy particles with specified boundary velocity. The framework is validated against different cases of open channel flow with different boundary conditions. The model can efficiently capture flow evolution and vortex generation for random geometry and variable boundary conditions.

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.

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

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

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

Containment atmosphere response to external sprays

International Nuclear Information System (INIS)

Green, J.; Almenas, K.

1995-01-01

The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J 2 /He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated

Containment atmosphere response to external sprays

Energy Technology Data Exchange (ETDEWEB)

Green, J.; Almenas, K. [Univ. of Maryland, College Park, MD (United States)

1995-09-01

The application of external sprays to a containment steel shell can be an effective energy removal method and has been proposed in the passive AP-600 design. Reduction of the steel shell temperature in contact with the containment atmosphere enhances both heat and mass transfer driving forces. Large scale experimental data in this area is scarce, therefore the measurements obtained from the E series tests conducted at the German HDR facility deserve special attention. These long term tests simulated various severe accident conditions, including external spraying of the hemispherical steel shell. This investigation focuses upon the integral response of the HDR containment atmosphere during spray periods and upon methods by which lumped parameter system codes, like CONTAIN, model the underlying condensation phenomena. Increases in spray water flowrates above a minimum value were ineffective at improving containment pressure reduction since the limiting resistance for energy transfer lies in the noncondensable-vapor boundary layer at the inner condensing surface. The spray created an unstable condition by cooling the upper layers of a heated atmosphere and thus inducing global natural circulation flows in the facility and subsequently, abrupt changes in lighter-than-air noncondensable (J{sub 2}/He) concentrations. Modeling results using the CONTAIN code are outlined and code limitations are delineated.

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

Westinghouse-GOTHIC comparisons to AP600 passive containment cooling tests

International Nuclear Information System (INIS)

Kennedy, M.D.; Woodcock, J.; Gresham, J.A.

1994-01-01

Westinghouse-GOTHIC is a thermal-hydraulics code well suited to analyzing passively cooled containments which depend on heat removal primarily through the containment shell. The code includes boundary layer heat and mass transfer correlations. A liquid film convective energy transport model has been added to the Westinghouse-GOTHIC code to account for the sensible heat change of the applied exterior water. The objective of this paper is to compare the code's predictions of the AP600 large scale test facility with and without the liquid film convective energy transport model. The predicted vessel pressure and integrated heat rate with and without the film convective energy transport model will be compared to the measured data. (author)

On the interpretation of different flow vectors of different ion species in the magnetospheric boundary layer

International Nuclear Information System (INIS)

Lundin, R.; Stasiewicz, K.; Hultqvist, B.

1986-05-01

Recent measurements of the ion composition in the magnetospheric boundary layer indicate that the boundary layer may contain clouds of magnetosheath plasma which are gradually becoming mixed with the magnetospheric plasma. A significant difference between flow vectors of different ion species (ca50-100 km/s) implies that an ideal MHD equation E+VxB=0, does not describe the macroscopic plasma flow inside such inhomogeneities. An analysis based on the first order drift theory indicates that gradients of the partial ion pressure and of the magnetic field could induce differential ion drifts comparable in magnitude to the electric drift velocity. We discuss some implications of these results on the physics of solar wind-magnetosphere interactions. (authors)

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

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

The probability of containment failure by direct containment heating in surry

International Nuclear Information System (INIS)

Pilch, M.M.; Allen, M.D.; Bergeron, K.D.; Tadios, E.L.; Stamps, D.W.; Spencer, B.W.; Quick, K.S.; Knudson, D.L.

1995-05-01

In a light-water reactor core melt accident, if the reactor pressure vessel (RPV) fails while the reactor coolant system (RCS) at high pressure, the expulsion of molten core debris may pressurize the reactor containment building (RCB) beyond its failure pressure. A failure in the bottom head of the RPV, followed by melt expulsion and blowdown of the RCS, will entrain molten core debris in the high-velocity steam blowdown gas. This chain of events is called a high-pressure melt ejection (HPME). Four mechanisms may cause a rapid increase in pressure and temperature in the reactor containment: (1) blowdown of the RCS, (2) efficient debris-to-gas heat transfer, (3) exothermic metal-steam and metal-oxygen reactions, and (4) hydrogen combustion. These processes, which lead to increased loads on the containment building, are collectively referred to as direct containment heating (DCH). It is necessary to understand factors that enhance or mitigate DCH because the pressure load imposed on the RCB may lead to early failure of the containment

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.)

Impact of Compound Hydrate Dynamics on Phase Boundary Changes

Science.gov (United States)

Osegovic, J. P.; Max, M. D.

2006-12-01

Compound hydrate reactions are affected by the local concentration of hydrate forming materials (HFM). The relationship between HFM composition and the phase boundary is as significant as temperature and pressure. Selective uptake and sequestration of preferred hydrate formers (PF) has wide ranging implications for the state and potential use of natural hydrate formation, including impact on climate. Rising mineralizing fluids of hydrate formers (such as those that occur on Earth and are postulated to exist elsewhere in the solar system) will sequester PF before methane, resulting in a positive relationship between depth and BTU content as ethane and propane are removed before methane. In industrial settings the role of preferred formers can separate gases. When depressurizing gas hydrate to release the stored gas, the hydrate initial composition will set the decomposition phase boundary because the supporting solution takes on the composition of the hydrate phase. In other settings where hydrate is formed, transported, and then dissociated, similar effects can control the process. The behavior of compound hydrate systems can primarily fit into three categories: 1) In classically closed systems, all the material that can form hydrate is isolated, such as in a sealed laboratory vessel. In such systems, formation and decomposition are reversible processes with observed hysteresis related to mass or heat transfer limitations, or the order and magnitude in which individual hydrate forming gases are taken up from the mixture and subsequently released. 2) Kinetically closed systems are exposed to a solution mass flow across a hydrate mass. These systems can have multiple P-T phase boundaries based on the local conditions at each face of the hydrate mass. A portion of hydrate that is exposed to fresh mineralizing solution will contain more preferred hydrate formers than another portion that is exposed to a partially depleted solution. Examples of kinetically closed

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.)

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)

Critical deflagration waves leading to detonation onset under different boundary conditions

International Nuclear Information System (INIS)

Lin Wei; Zhou Jin; Lin Zhi-Yong; Fan Xiao-Hua

2015-01-01

High-speed turbulent critical deflagration waves before detonation onset in H 2 –air mixture propagated into a square cross section channel, which was assembled of optional rigid rough, rigid smooth, or flexible walls. The corresponding propagation characteristic and the influence of the wall boundaries on the propagation were investigated via high-speed shadowgraph and a high-frequency pressure sampling system. As a comprehensive supplement to the different walls effect investigation, the effect of porous absorbing walls on the detonation propagation was also investigated via smoke foils and the high-frequency pressure sampling system. Results are as follows. In the critical deflagration stage, the leading shock and the closely following turbulent flame front travel at a speed of nearly half the CJ detonation velocity. In the preheated zone, a zonary flame arises from the overlapping part of the boundary layer and the pressure waves, and then merges into the mainstream flame. Among these wall boundary conditions, the rigid rough wall plays a most positive role in the formation of the zonary flame and thus accelerates the transition of the deflagration to detonation (DDT), which is due to the boost of the boundary layer growth and the pressure wave reflection. Even though the flexible wall is not conducive to the pressure wave reflection, it brings out a faster boundary layer growth, which plays a more significant role in the zonary flame formation. Additionally, the porous absorbing wall absorbs the transverse wave and yields detonation decay and velocity deficit. After the absorbing wall, below some low initial pressure conditions, no re-initiation occurs and the deflagration propagates in critical deflagration for a relatively long distance. (paper)

Flat Plate Boundary Layer Stimulation Using Trip Wires and Hama Strips

Science.gov (United States)

Peguero, Charles; Henoch, Charles; Hrubes, James; Fredette, Albert; Roberts, Raymond; Huyer, Stephen

2017-11-01

Water tunnel experiments on a flat plate at zero angle of attack were performed to investigate the effect of single roughness elements, i.e., trip wires and Hama strips, on the transition to turbulence. Boundary layer trips are traditionally used in scale model testing to force a boundary layer to transition from laminar to turbulent flow at a single location to aid in scaling of flow characteristics. Several investigations of trip wire effects exist in the literature, but there is a dearth of information regarding the influence of Hama strips on the flat plate boundary layer. The intent of this investigation is to better understand the effects of boundary layer trips, particularly Hama strips, and to investigate the pressure-induced drag of both styles of boundary layer trips. Untripped and tripped boundary layers along a flat plate at a range of flow speeds were characterized with multiple diagnostic measurements in the NUWC/Newport 12-inch water tunnel. A wide range of Hama strip and wire trip thicknesses were used. Measurements included dye flow visualization, direct skin friction and parasitic drag force, boundary layer profiles using LDV, wall shear stress fluctuations using hot film anemometry, and streamwise pressure gradients. Test results will be compared to the CFD and boundary layer model results as well as the existing body of work. Conclusions, resulting in guidance for application of Hama strips in model scale experiments and non-dimensional predictions of pressure drag will be presented.

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

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.

Receptivity of Hypersonic Boundary Layers over Straight and Flared Cones

Science.gov (United States)

Balakumar, Ponnampalam; Kegerise, Michael A.

2010-01-01

The effects of adverse pressure gradients on the receptivity and stability of hypersonic boundary layers were numerically investigated. Simulations were performed for boundary layer flows over a straight cone and two flared cones. The steady and the unsteady flow fields were obtained by solving the two-dimensional Navier-Stokes equations in axi-symmetric coordinates using the 5th order accurate weighted essentially non-oscillatory (WENO) scheme for space discretization and using third-order total-variation-diminishing (TVD) Runge-Kutta scheme for time integration. The mean boundary layer profiles were analyzed using local stability and non-local parabolized stability equations (PSE) methods. After the most amplified disturbances were identified, two-dimensional plane acoustic waves were introduced at the outer boundary of the computational domain and time accurate simulations were performed. The adverse pressure gradient was found to affect the boundary layer stability in two important ways. Firstly, the frequency of the most amplified second-mode disturbance was increased relative to the zero pressure gradient case. Secondly, the amplification of first- and second-mode disturbances was increased. Although an adverse pressure gradient enhances instability wave growth rates, small nose-tip bluntness was found to delay transition due to the low receptivity coefficient and the resulting weak initial amplitude of the instability waves. The computed and measured amplitude-frequency spectrums in all three cases agree very well in terms of frequency and the shape except for the amplitude.

Direct measurement of methane hydrate composition along the hydrate equilibrium boundary

Science.gov (United States)

Circone, S.; Kirby, S.H.; Stern, L.A.

2005-01-01

The composition of methane hydrate, namely nW for CH 4??nWH2O, was directly measured along the hydrate equilibrium boundary under conditions of excess methane gas. Pressure and temperature conditions ranged from 1.9 to 9.7 MPa and 263 to 285 K. Within experimental error, there is no change in hydrate composition with increasing pressure along the equilibrium boundary, but nW may show a slight systematic decrease away from this boundary. A hydrate stoichiometry of n W = 5.81-6.10 H2O describes the entire range of measured values, with an average composition of CH4??5.99(??0.07) H2O along the equilibrium boundary. These results, consistent with previously measured values, are discussed with respect to the widely ranging values obtained by thermodynamic analysis. The relatively constant composition of methane hydrate over the geologically relevant pressure and temperature range investigated suggests that in situ methane hydrate compositions may be estimated with some confidence. ?? 2005 American Chemical Society.

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

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)

Boundary-Layer Characteristics of Persistent Regional Haze Events and Heavy Haze Days in Eastern China

Directory of Open Access Journals (Sweden)

Peng Huaqing

2016-01-01

Full Text Available This paper analyzed the surface conditions and boundary-layer climate of regional haze events and heavy haze in southern Jiangsu Province in China. There are 5 types with the surface conditions which are equalized pressure (EQP, the advancing edge of a cold front (ACF, the base of high pressure (BOH, the backside of high pressure (BAH, the inverted trough of low pressure (INT, and saddle pressure (SAP with the haze days. At that time, 4 types are divided with the regional haze events and each of which has a different boundary-layer structure. During heavy haze, the surface mainly experiences EQP, ACF, BOH, BAH, and INT which also have different boundary-layer structures.

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...

The probability of containment failure by direct containment heating in Zion. Supplement 1

International Nuclear Information System (INIS)

Pilch, M.M.; Allen, M.D.; Stamps, D.W.; Tadios, E.L.; Knudson, D.L.

1994-12-01

Supplement 1 of NUREG/CR-6075 brings to closure the DCH issue for the Zion plant. It includes the documentation of the peer review process for NUREG/CR-6075, the assessments of four new splinter scenarios defined in working group meetings, and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes. SCDAP/RELAP5 was used to analyze three short-term station blackout cases with Different lead rates. In all three case, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a threat. However, these calculations were continued to lower head failure in order to gain insights that were useful in establishing the initial and boundary conditions. The most useful insights are that the RCS pressure is-low at vessel breach metallic blockages in the core region do not melt and relocate into the lower plenum, and melting of upper plenum steel is correlated with hot leg failure. THE SCDAP/RELAP output was used as input to CONTAIN to assess the containment conditions at vessel breach. The containment-side conditions predicted by CONTAIN are similar to those originally specified in NUREG/CR-6075

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)

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)

Boundary vapor contentsin an annular channel

International Nuclear Information System (INIS)

Remizov, O.V.; Shurkin, N.G.; Podgornyj, K.K.; Gal'chenko, Eh.F.; Bukhteev, I.S.

1978-01-01

The work is aimed at the experimental investigation of the worsening of the heat transfer in an annular channel. The experiments have been carried out on the annular channel 32x28x3000 mm with the even distribution of the heat flux along the length at pressures of 6.9-19.6 MPa, flow rate of 350-1000 kg/m 2 s, and specific heat fluxes from 0.18 up to 0.6 MW/m 2 . Heating is external, oneside. Water monodistillate of the following composition has been used as a coolant: pH 9; dry residue - 0.8-1.2 mg/kg, oxygen -10-15 mg/kg. It is found out that the change character of the temperature field of the heating surface of the annular channel at the regime with the worsen of heat emission depends on the ratio of regime parameters. At pressures of 6.9-13.7 MPa and flow rate of 350-500 kg/m 2 s the channel wall temperature rises monotoneously, never reaching its maximum. With pressure rise > 13.7 MPa and mass velocity > 500 kg/m 2 s the temperature of the heat emitting surface reaches its maximum, and then slowly falls. At pressures of 6.9-11.8 MPa the boundary vapor content value within the whole range of mass velocities does not depend on the specific heat flux q. At pressures higher than 13.7 MPa and mass velocities of 350-1000 kg/m 2 s the boundary vapor content depends on q. The heating of the external or internal surface of the annular channel affects the value of the boundary vapor content within the whole range of regime parameters' change under investigation

Influence of model boundary conditions on blood flow patterns in a patient specific stenotic right coronary artery.

Science.gov (United States)

Liu, Biyue; Zheng, Jie; Bach, Richard; Tang, Dalin

2015-01-01

In literature, the effect of the inflow boundary condition was investigated by examining the impact of the waveform and the shape of the spatial profile of the inlet velocity on the cardiac hemodynamics. However, not much work has been reported on comparing the effect of the different combinations of the inlet/outlet boundary conditions on the quantification of the pressure field and flow distribution patterns in stenotic right coronary arteries. Non-Newtonian models were used to simulate blood flow in a patient-specific stenotic right coronary artery and investigate the influence of different boundary conditions on the phasic variation and the spatial distribution patterns of blood flow. The 3D geometry of a diseased artery segment was reconstructed from a series of IVUS slices. Five different combinations of the inlet and the outlet boundary conditions were tested and compared. The temporal distribution patterns and the magnitudes of the velocity, the wall shear stress (WSS), the pressure, the pressure drop (PD), and the spatial gradient of wall pressure (WPG) were different when boundary conditions were imposed using different pressure/velocity combinations at inlet/outlet. The maximum velocity magnitude in a cardiac cycle at the center of the inlet from models with imposed inlet pressure conditions was about 29% lower than that from models using fully developed inlet velocity data. Due to the fact that models with imposed pressure conditions led to blunt velocity profile, the maximum wall shear stress at inlet in a cardiac cycle from models with imposed inlet pressure conditions was about 29% higher than that from models with imposed inlet velocity boundary conditions. When the inlet boundary was imposed by a velocity waveform, the models with different outlet boundary conditions resulted in different temporal distribution patterns and magnitudes of the phasic variation of pressure. On the other hand, the type of different boundary conditions imposed at the

Axisymmetric MHD stability of sharp-boundary Tokamaks

International Nuclear Information System (INIS)

Rebhan, E.; Salat, A.

1976-09-01

For a sharp-boundary, constant pressure plasma model of axisymmetric equilibria the MHD stability problem of axisymmetric perturbations is solved by analytic reduction to a one-dimensional problem on the boundary and subsequent numerical treatment, using the energy principle. The stability boundaries are determined for arbitrary aspect ratio, arbitrary βsub(p) and elliptical, triangular and rectangular plasma cross-sections, wall stabilization not being taken into account. It is found that the axisymmetric stability strongly depends on the plasma shape and is almost independent of the safety factor q. (orig.) [de

Jets from pulsed-ultrasound-induced cavitation bubbles near a rigid boundary

Science.gov (United States)

Brujan, Emil-Alexandru

2017-06-01

The dynamics of cavitation bubbles, generated from short (microsecond) pulses of ultrasound and situated near a rigid boundary, are investigated numerically. The temporal development of the bubble shape, bubble migration, formation of the liquid jet during bubble collapse, and the kinetic energy of the jet are investigated as a function of the distance between bubble and boundary. During collapse, the bubble migrates towards the boundary and the liquid jet reaches a maximum velocity between 80 m s-1 and 120 m s-1, depending on the distance between bubble and boundary. The conversion of bubble energy to kinetic energy of the jet ranges from 16% to 23%. When the bubble is situated in close proximity to the boundary, the liquid jet impacts the boundary with its maximum velocity, resulting in an impact pressure of the order of tens of MPa. The rapid expansion of the bubble, the impact of the liquid jet onto the nearby boundary material, and the high pressure developed inside the bubble at its minimum volume can all contribute to the boundary material damage. The high pressure developed during the impact of the liquid jet onto the biological material and the shearing forces acting on the material surface as a consequence of the radial flow of the jet outward from the impact site are the main damage mechanisms of rigid biological materials. The results are discussed with respect to cavitation damage of rigid biological materials, such as disintegration of renal stones and calcified tissue and collateral effects in pulsed ultrasound surgery.

Jets from pulsed-ultrasound-induced cavitation bubbles near a rigid boundary

International Nuclear Information System (INIS)

Brujan, Emil-Alexandru

2017-01-01

The dynamics of cavitation bubbles, generated from short (microsecond) pulses of ultrasound and situated near a rigid boundary, are investigated numerically. The temporal development of the bubble shape, bubble migration, formation of the liquid jet during bubble collapse, and the kinetic energy of the jet are investigated as a function of the distance between bubble and boundary. During collapse, the bubble migrates towards the boundary and the liquid jet reaches a maximum velocity between 80 m s −1 and 120 m s −1 , depending on the distance between bubble and boundary. The conversion of bubble energy to kinetic energy of the jet ranges from 16% to 23%. When the bubble is situated in close proximity to the boundary, the liquid jet impacts the boundary with its maximum velocity, resulting in an impact pressure of the order of tens of MPa. The rapid expansion of the bubble, the impact of the liquid jet onto the nearby boundary material, and the high pressure developed inside the bubble at its minimum volume can all contribute to the boundary material damage. The high pressure developed during the impact of the liquid jet onto the biological material and the shearing forces acting on the material surface as a consequence of the radial flow of the jet outward from the impact site are the main damage mechanisms of rigid biological materials. The results are discussed with respect to cavitation damage of rigid biological materials, such as disintegration of renal stones and calcified tissue and collateral effects in pulsed ultrasound surgery. (paper)

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

Probabilistic finite element investigation of prestressing loss in nuclear containment wall segments

International Nuclear Information System (INIS)

Balomenos, Georgios P.; Pandey, Mahesh D.

2017-01-01

Highlights: • Probabilistic finite element framework for assessing concrete strain distribution. • Investigation of prestressing loss based on concrete strain distribution. • Application to 3D nuclear containment wall segments. • Use of ABAQUS with python programing for Monte Carlo simulation. - Abstract: The main function of the concrete containment structures is to prevent radioactive leakage to the environment in case of a loss of coolant accident (LOCA). The Canadian Standard CSA N287.6 (2011) proposes periodic inspections, i.e., pressure testing, in order to assess the strength and design criteria of the containment (proof test) and the leak tightness of the containment boundary (leakage rate test). During these tests, the concrete strains are measured and are expected to have a distribution due to several uncertainties. Therefore, this study aims to propose a probabilistic finite element analysis framework. Then, investigates the relationship between the concrete strains and the prestressing loss, in order to examine the possibility of estimating the average prestressing loss during pressure testing inspections. The results indicate that the concrete strain measurements during the leakage rate test may provide information with respect to the prestressing loss of the bonded system. In addition, the demonstrated framework can be further used for the probabilistic finite element analysis of real scale containments.

Probabilistic finite element investigation of prestressing loss in nuclear containment wall segments

Energy Technology Data Exchange (ETDEWEB)

Balomenos, Georgios P., E-mail: gbalomen@uwaterloo.ca; Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca

2017-01-15

Highlights: • Probabilistic finite element framework for assessing concrete strain distribution. • Investigation of prestressing loss based on concrete strain distribution. • Application to 3D nuclear containment wall segments. • Use of ABAQUS with python programing for Monte Carlo simulation. - Abstract: The main function of the concrete containment structures is to prevent radioactive leakage to the environment in case of a loss of coolant accident (LOCA). The Canadian Standard CSA N287.6 (2011) proposes periodic inspections, i.e., pressure testing, in order to assess the strength and design criteria of the containment (proof test) and the leak tightness of the containment boundary (leakage rate test). During these tests, the concrete strains are measured and are expected to have a distribution due to several uncertainties. Therefore, this study aims to propose a probabilistic finite element analysis framework. Then, investigates the relationship between the concrete strains and the prestressing loss, in order to examine the possibility of estimating the average prestressing loss during pressure testing inspections. The results indicate that the concrete strain measurements during the leakage rate test may provide information with respect to the prestressing loss of the bonded system. In addition, the demonstrated framework can be further used for the probabilistic finite element analysis of real scale containments.

Acoustic Radiation From a Mach 14 Turbulent Boundary Layer

Science.gov (United States)

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

2016-01-01

Direct numerical simulations (DNS) are used to examine the turbulence statistics and the radiation field generated by a high-speed turbulent boundary layer with a nominal freestream Mach number of 14 and wall temperature of 0:18 times the recovery temperature. The flow conditions fall within the range of nozzle exit conditions of the Arnold Engineering Development Center (AEDC) Hypervelocity Tunnel No. 9 facility. The streamwise domain size is approximately 200 times the boundary-layer thickness at the inlet, with a useful range of Reynolds number corresponding to Re 450 ?? 650. Consistent with previous studies of turbulent boundary layer at high Mach numbers, the weak compressibility hypothesis for turbulent boundary layers remains applicable under this flow condition and the computational results confirm the validity of both the van Driest transformation and Morkovin's scaling. The Reynolds analogy is valid at the surface; the RMS of fluctuations in the surface pressure, wall shear stress, and heat flux is 24%, 53%, and 67% of the surface mean, respectively. The magnitude and dominant frequency of pressure fluctuations are found to vary dramatically within the inner layer (z/delta 0.< or approx. 0.08 or z+ < or approx. 50). The peak of the pre-multiplied frequency spectrum of the pressure fluctuation is f(delta)/U(sub infinity) approx. 2.1 at the surface and shifts to a lower frequency of f(delta)/U(sub infinity) approx. 0.7 in the free stream where the pressure signal is predominantly acoustic. The dominant frequency of the pressure spectrum shows a significant dependence on the freestream Mach number both at the wall and in the free stream.

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

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)

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

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...

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.

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)

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.)

Liquid radioactive waste processing system for pressurized water reactor plants

International Nuclear Information System (INIS)

Anon.

1976-01-01

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

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

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

International Nuclear Information System (INIS)

Baratta, A.J.

1997-01-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

Energy Technology Data Exchange (ETDEWEB)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.

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.)

On hairpin vortices in a transitional boundary layer

Directory of Open Access Journals (Sweden)

Uruba Václav

2012-04-01

Full Text Available In the presented paper the results of experiments on transitional boundary layer are presented. The boundary layer was generated on smooth flat wall with zero pressure gradient forming one side of the channel of rectangular cross section. The hairpin vortices, packets of hairpin vortices, turbulent spots and calmed regions were experimentally investigated using time-resolved PIV technique.

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

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.

A Comment Upon Previous Studies on 3-D Boundary Layer Transition

OpenAIRE

ÇARPINLIOĞLU, Melda Özdinç

2014-01-01

The common feature of the experimental studies upon 3-D boundary layer development on swept flat plates cited in the available literature is the application of streamwise and/or spanwise pressure gradients. In fact; presence of the pressure gradients was suggested to be vital for having crossflow effective in 3-D boundary layer transition. In the presented paper here, this idea is questioned evaluating the results of an experimental investigation conducted on swept flat plates under the ab...

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...

HIFiRE-1 Turbulent Shock Boundary Layer Interaction - Flight Data and Computations

Science.gov (United States)

Kimmel, Roger L.; Prabhu, Dinesh

2015-01-01

The Hypersonic International Flight Research Experimentation (HIFiRE) program is a hypersonic flight test program executed by the Air Force Research Laboratory (AFRL) and Australian Defence Science and Technology Organisation (DSTO). This flight contained a cylinder-flare induced shock boundary layer interaction (SBLI). Computations of the interaction were conducted for a number of times during the ascent. The DPLR code used for predictions was calibrated against ground test data prior to exercising the code at flight conditions. Generally, the computations predicted the upstream influence and interaction pressures very well. Plateau pressures on the cylinder were predicted well at all conditions. Although the experimental heat transfer showed a large amount of scatter, especially at low heating levels, the measured heat transfer agreed well with computations. The primary discrepancy between the experiment and computation occurred in the pressures measured on the flare during second stage burn. Measured pressures exhibited large overshoots late in the second stage burn, the mechanism of which is unknown. The good agreement between flight measurements and CFD helps validate the philosophy of calibrating CFD against ground test, prior to exercising it at flight conditions.

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

CONTAIN code analyses of direct containment heating experiments

International Nuclear Information System (INIS)

Williams, D.C.; Griffith, R.O.; Tadios, E.L.; Washington, K.E.

1995-01-01

In some nuclear reactor core-melt accidents, a potential exists for molten core-debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the U.S. Nuclear Regulatory Commission (NRC). The DCH modeling has been an important focus for the development of the CONTAIN code. Results of a detailed independent peer review of the CONTAIN code were published recently. This paper summarizes work performed in support of the peer review in which the CONTAIN code was applied to analyze DCH experiments. Goals of this work were comparison of calculated and experimental results, CONTAIN DCH model assessment, and development of guidance for code users, including development of a standardized input prescription for DCH analysis

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

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.

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.)

Planetary boundaries: exploring the safe operating space for humanity

Science.gov (United States)

Johan Rockström; Will Steffen; Kevin Noone; Asa Persson; F. Stuart Chapin; Eric Lambin; Timothy M. Lenton; Marten Scheffer; Carl Folke; Hans Joachim Schellnhuber; Björn Nykvist; Cynthia A. de Wit; Terry Hughes; Sander van der Leeuw; Henning Rodhe; Sverker Sörlin; Peter K. Snyder; Robert Costanza; Uno Svedin; Malin Falkenmark; Louise Karlberg; Robert W. Corell; Victoria J. Fabry; James Hansen; Brian Walker; Diana Liverman; Katherine Richardson; Paul Crutzen; Jonathan Foley

2009-01-01

Anthropogenic pressures on the Earth System have reached a scale where abrupt global environmental change can no longer be excluded. We propose a new approach to global sustainability in which we define planetary boundaries within which we expect that humanity can operate safely. Transgressing one or more planetary boundaries may be deleterious or even catastrophic due...

Reinforced concrete containment structures in high seismic zones

International Nuclear Information System (INIS)

Aziz, T.S.

1977-01-01

A new structural concept for reinforced concrete containment structures at sites where earthquake ground motions in terms of the Safe Shutdown Earthquake (SSE) exceeds 0.3 g is presented. The structural concept is based on: (1) an inner steel-lined concrete shell which houses the reactor and provides shielding and containment in the event of loss of coolant accident; (2) an outer annular concrete shell structure which houses auxilary reactor equipment and safeguards systems. These shell structures are supported on a common foundation mat which is embeded in the subgrade. Under stipulated earthquake conditions the two shell structures interact to resist lateral inertia forces. Thus the annular structure which is not a pressure boundary acts as a lateral support for the inner containment shell. The concept is practical, economically feasible and new to practice. An integrated configuration which includes the interior shell, the annular structure and the subgrade is analyzed for several static and dynamic loading conditions. The analysis is done using a finite difference solution scheme for the static loading conditions. A semi-analytical three-dimensional finite element scheme combined with a Fast Fourier Transform (FFT) algorithm is used for the dynamic loading conditions. The effects of cracking of the containment structure due to pressurization in conjunction with earthquake loading are discussed. Analytical results for both the finite difference and the finite element schemes are presented and the sensitivity of the results to changes in the input parameters is studied. General recommendations are given for plant configurations where high seismic loading is a major design consideration

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.)

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

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

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.

Characteristics of wall pressure over wall with permeable coating

Energy Technology Data Exchange (ETDEWEB)

Song, Woo Seog; Shin, Seungyeol; Lee, Seungbae [Inha Univ., Incheon (Korea, Republic of)

2012-11-15

Fluctuating wall pressures were measured using an array of 16 piezoelectric transducers beneath a turbulent boundary layer. The coating used in this experiment was an open cell, urethane type foam with a porosity of approximately 50 ppi. The ultimate objective of the coating is to provide a mechanical filter to reduce the wall pressure fluctuations. The ultimate objective of the coating is to provide a mechanical filter to reduce the wall pressure fluctuations. The boundary layer on the flat plate was measured by using a hot wire probe, and the CPM method was used to determine the skin friction coefficient. The wall pressure autospectra and streamwise wavenumber frequency spectra were compared to assess the attenuation of the wall pressure field by the coating. The coating is shown to attenuate the convective wall pressure energy. However, the relatively rough surface of the coating in this investigation resulted in a higher mean wall shear stress, thicker boundary layer, and higher low frequency wall pressure spectral levels compared to a smooth wall.

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.)

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)

Fuselage boundary-layer refraction of fan tones radiated from an installed turbofan aero-engine.

Science.gov (United States)

Gaffney, James; McAlpine, Alan; Kingan, Michael J

2017-03-01

A distributed source model to predict fan tone noise levels of an installed turbofan aero-engine is extended to include the refraction effects caused by the fuselage boundary layer. The model is a simple representation of an installed turbofan, where fan tones are represented in terms of spinning modes radiated from a semi-infinite circular duct, and the aircraft's fuselage is represented by an infinitely long, rigid cylinder. The distributed source is a disk, formed by integrating infinitesimal volume sources located on the intake duct termination. The cylinder is located adjacent to the disk. There is uniform axial flow, aligned with the axis of the cylinder, everywhere except close to the cylinder where there is a constant thickness boundary layer. The aim is to predict the near-field acoustic pressure, and in particular, to predict the pressure on the cylindrical fuselage which is relevant to assess cabin noise. Thus no far-field approximations are included in the modelling. The effect of the boundary layer is quantified by calculating the area-averaged mean square pressure over the cylinder's surface with and without the boundary layer included in the prediction model. The sound propagation through the boundary layer is calculated by solving the Pridmore-Brown equation. Results from the theoretical method show that the boundary layer has a significant effect on the predicted sound pressure levels on the cylindrical fuselage, owing to sound radiation of fan tones from an installed turbofan aero-engine.

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)

Transonic shock wave. Boundary layer interaction at a convex wall

NARCIS (Netherlands)

Koren, B.; Bannink, W.J.

1984-01-01

A standard finite element procedure has been applied to the problem of transonic shock wave – boundary layer interaction at a convex wall. The method is based on the analytical Bohning-Zierep model, where the boundary layer is perturbed by a weak normal shock wave which shows a singular pressure

Boundary Slip and Surface Interaction: A Lattice Boltzmann Simulation

International Nuclear Information System (INIS)

Yan-Yan, Chen; Hua-Bing, Li; Hou-Hui, Yi

2008-01-01

The factors affecting slip length in Couette geometry flows are analysed by means of a two-phase mesoscopic lattice Boltzmann model including non-ideal fluid-fluid and fluid-wall interactions. The main factors influencing the boundary slip are the strength of interactions between fluid-fluid and fluid-wall particles. Other factors, such as fluid viscosity, bulk pressure may also change the slip length. We find that boundary slip only occurs under a certain density (bulk pressure). If the density is large enough, the slip length will tend to zero. In our simulations, a low density layer near the wall does not need to be postulated a priori but emerges naturally from the underlying non-ideal mesoscopic dynamics. It is the low density layer that induces the boundary slip. The results may be helpful to understand recent experimental observations on the slippage of micro flows

BWR containments license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Smith, S.; Gregor, F.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures, and components, in the license renewal technical Industry Reports (IR's). License renewal applicants may choose to reference these IR's in support of their plant-specific license renewal applications as an equivalent to the integrated plant assessment provisions of the license renewal rule (IOCFR54). The scope of the IR provides the technical basis for license renewal for U.S. Boiling Water Reactor (BWR) containments. The scope of the report includes containments constructed of reinforced or prestressed concrete with steel liners and freestanding stell containments. Those domestic BWR containments designated as Mark I, Mark II or Mark III are covered, but no containments are addressed before these designs. The report includes those items within the jurisdictional boundaries for metal and concrete containments defined by Section III of the ASME Boiler and Pressure Vessel Code, Division 1, Subsection NE (Class MC) and Division 2 (Class CC) and their supports, but excluding snubbers

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

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

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.

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.

Airfoil boundary layer separation and control at low Reynolds numbers

Energy Technology Data Exchange (ETDEWEB)

Yarusevych, S.; Sullivan, P.E. [University of Toronto, Department of Mechanical and Industrial Engineering, Toronto, ON (Canada); Kawall, J.G. [Ryerson University, Department of Mechanical and Industrial Engineering, Toronto, ON (Canada)

2005-04-01

The boundary layer separation on a NACA 0025 airfoil was studied experimentally via hot-wire anemometry and surface pressure measurements. The results provide added insight into periodic boundary layer control, suggesting that matching the excitation frequency with the most amplified disturbance in the separated shear layer is optimal for improving airfoil performance. (orig.)

Numerical modelling of the processes in the WWER-1000 containment building during cold leg LOCA using the CONTEMPT-LT/026 code

International Nuclear Information System (INIS)

Kolev, N.I.; Sybotinov, L.S.

1984-01-01

The CONTEMPT-LT/026 code has been used to produce numerical results for the processes in a WWER-1000 containment building during cold leg LOCA with break at the reactor vessel. The objective of the analysis is to estimate the maximal loads on the containment in case of LOCA. Available design data for the geometry and for the operational characteristics of the low-pressure ECC system and the sprinkler system have been used. Boundary conditions such as mass flow and enthalpies at the breach are given by a RELAP4/MOD6 computation. Hydrogen explosions in the containment are not considered. It is found that in case of normal functioning of the low-pressure ECC system the maximal pressure is 3,26±0,44 bar. In the case of malfunctioning of the low-pressure ECC system, the predicted maximal pressure is 4±0,44 bar, when: a) only 50% of the heat transfer surface of the heat exchanger is effectively used due to pollution; b) the main pipeline of the sprinkler is broken; c) the pipeline to the heat exchanger is partially broken so that the mass flow through the exchanger is only 50% of the nominal; and d) ECC low-pressure ECC system attains its maximal efficiency within 3 min, the predicted maximal pressure is 4±0,44 bar

Structural integrity evaluation of PWR nuclear reactor pressure vessels

International Nuclear Information System (INIS)

Cruz, Julio R.B.; Mattar Neto, Miguel

1999-01-01

The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

Problems of matter-antimatter boundary layers

International Nuclear Information System (INIS)

Lehnert, B.

1975-01-01

This paper outlines the problems of the quasi-steady matter-antimatter boundary layers discussed in Klein-Alfven's cosmological theory, and a crude model of the corresponding ambiplasma balance is presented: (i) at interstellar particle densities, no well-defined boundary layer can exist in presence of neutral gas, nor can such a layer be sustained in an unmagnetized fully ionized ambiplasma. (ii) Within the limits of applicability of the present model, sharply defined boundary layers are under certain conditions found to exist in a magnetized ambiplasma. Thus, at beta values less than unity, a steep pressure drop of the low-energy components of matter and antimatter can be balanced by a magnetic field and the electric currents in the ambiplasma. (iii) The boundary layer thickness is of the order of 2x 0 approximately 10/BT 0 sup(1/4) meters, where B is the magnetic field strength in MKS units and T 0 the characteristic temperature of the low-energy components in the layer. (Auth.)

Mark III Containment vessel/annulus concrete design

International Nuclear Information System (INIS)

Chang, P.S.; Moussa, M.M.

1981-01-01

Recently, engineers have been considering the significant dynamic impact of safety/relief valve (S/RV) discharge loads on the containment structures, safety equipment, and piping systems in BWR type reactors. For a plant in the construction stage, extensive modifications will be made to qualify these new loads. The lower portion of the containment vessel serves as a suppression pool pressure boundary and is designed to sustain the effects of postulated loss of coolant accidents, seismic occurrences, S/RV discharge loads, and other effects. Extremely high spectral peak accelerations of the free-standing steel containment vessel can be obtained during the air dearing process of the S/RV discharge. Parametric studies indicated that a substantial reduction in response can be obtained by increasing the stiffness of the steel containment vessel in the lover area. A concrete backing configuration in the suppression pool area of Mark III Containment is proposed in this paper. A composite action is assumed between the steel containment vessel shell and the concrete section. The system is physically separated from the shield building. This approach warrants an early erection of the shield building and a late installation of piping systems in the containment vessel suppression pool area. Finite element analyses are performed by using ASHSD2 and EASE2 computer codes. The results of the analyses have shown the proposed stress criteria are satisfied. The approach pressented is justified to be a workable system for a new plant design. (orig./HP)

Integration of piezo-capacitive and piezo-electric nanoweb based pressure sensors for imaging of static and dynamic pressure distribution.

Science.gov (United States)

Jeong, Y J; Oh, T I; Woo, E J; Kim, K J

2017-07-01

Recently, highly flexible and soft pressure distribution imaging sensor is in great demand for tactile sensing, gait analysis, ubiquitous life-care based on activity recognition, and therapeutics. In this study, we integrate the piezo-capacitive and piezo-electric nanowebs with the conductive fabric sheets for detecting static and dynamic pressure distributions on a large sensing area. Electrical impedance tomography (EIT) and electric source imaging are applied for reconstructing pressure distribution images from measured current-voltage data on the boundary of the hybrid fabric sensor. We evaluated the piezo-capacitive nanoweb sensor, piezo-electric nanoweb sensor, and hybrid fabric sensor. The results show the feasibility of static and dynamic pressure distribution imaging from the boundary measurements of the fabric sensors.

Experimental results from containment piping bellows subjected to severe accident conditions. Volume 1, Results from bellows tested in 'like-new' conditions

International Nuclear Information System (INIS)

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

1994-09-01

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall, while minimizing the load imposed on the piping and wall. Piping bellows are primarily used in steel containments; however, they have received limited use in some concrete (reinforced and prestressed) containments. In a severe accident they may be subjected to pressure and temperature conditions that exceed the design values, along with a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted under the sponsorship of the US Nuclear Regulatory Commission at Sandia National Laboratories. Several different bellows geometries, representative of actual containment bellows, have been subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of thirteen bellows have been tested, all in the 'like-new' condition. (Additional tests are planned of bellows that have been subjected to corrosion.) The tests showed that bellows are capable of withstanding relatively large deformations, up to, or near, the point of full compression or elongation, before developing leakage. The test data is presented and discussed

Emergency operating procedures guidelines for pressurized water reactors - a progress report

International Nuclear Information System (INIS)

Lyon, W.C.

1984-01-01

Emergency Operating Procedures (EOPs) contain the instructions the operator will follow to control a nuclear plant whenever a condition exists that potentially jeopardizes the fuel cladding, the reactor coolant system (RCS) pressure boundary, or the containment. The EOPs are prepared from guidelines which contain the major operator instructions that will be in the EOPs. Guidelines have been prepared by owners' groups having Babcock and Wilcox (BandW), Combustion Engineering (CE), General Electric (GE), and Westinghouse (W) plants. These guidelines cover many aspects of full power operation. Future effort is anticipated to complete coverage of transient events, including severe accidents, all power conditions, and shutdown. This paper describes the philosophy which has guided NRC technical review of guidelines, progress achieved in providing comprehensive coverage of emergency conditions for PWRs, and anticipated future technical activities

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

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

The status of improved pressurized heavy water reactor development - A new option for PHWR -

International Nuclear Information System (INIS)

Park, Tae Keun; Yeo, Ji Won

1996-03-01

Currently, the 900 MWe class Improved Pressurized Heavy Water Reactor (PHWR), which is a type of CANDU reactor based on the systems and components of operating CANDU plants, is under development. The improved PHWR has a 480 fuel channel calandria, uses 37 element natural uranium fuel bundles and has a single unit containment. Adaptation of a steel-lined containment structure and improved containment isolation systems permit a reduced exclusion area boundary (EAB) compared to the existing larger capacity CANDU reactors (Darlington, Bruce B). The improved PHWR buildings are arranged to achieve minimum spacing between reactor units. Plant safety and economy are increased through various design changes based on the operating experience of existing CANDU plants. 4 refs. (Author)

Ventilation air conditioner for a reactor container

International Nuclear Information System (INIS)

Ikegame, Noboru; Nakagawa, Takeshi.

1980-01-01

Purpose: To suppress the variations in the internal pressure of a reactor container and smoothly ventilate the reactor container. Constitution: The air conditioner provides an air-flow-rate-control damper, a purge-air supply fan, and a filter device in the air-supply pipe of a reactor container. Furthermore, it provides a pressure difference detector at a part of the container. The air-flow-rate-control damper is connected electrically through a position-modulator-comparison amplifier to the pressure difference detector. When the filtration becomes insufficient by clogging of the filter device and the internal pressure increased abruptly in the container, the pressure-difference detector can detect it, and the damper is operated by a pressure regulator and the comparator so as to control the air flow to the container. Thus, the internal pressure variation is controlled so as to easily ventilate the container. (J.P.N.)

Grain boundary corrosion of copper canister material

International Nuclear Information System (INIS)

Fennell, P.A.H.; Graham, A.J.; Smart, N.R.; Sofield, C.J.

2001-03-01

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. The potential for grain boundary corrosion was investigated by exposing copper specimens, which had undergone different heat treatments and hence had different grain sizes, to aerated artificial bentonite-equilibrated groundwater with two concentrations of chloride, for increasing periods of time. The degree of grain boundary corrosion was determined by atomic force microscopy (AFM) and optical microscopy. AFM showed no increase in grain boundary 'ditching' for low chloride groundwater. In high chloride groundwater the surface was covered uniformly with a fine-grained oxide. No increases in oxide thickness were observed. No significant grain boundary attack was observed using optical microscopy either. The work suggests that in aerated artificial groundwaters containing chloride ions, grain boundary corrosion of copper is unlikely to adversely affect SKB's copper canisters

Boundary-Layer Bypass Transition Over Large-Scale Bodies

Science.gov (United States)

2016-12-16

behaviour of the velocity and pressure changes with the curvature. This work aims to extend the results of the flat-plate boundary layer to a Rankine...example, consume an enormous amount of energy due to friction, many works have been directed to the suppression of transitional boundary layer disturbances...decrease of the enormous amount of energy consumed by airplanes during flight, moreover flight costs and aerodynamic noise could be reduced and number

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.

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

Boundary-layer effects in droplet splashing

Science.gov (United States)

Riboux, Guillaume; Gordillo, Jose Manuel

2017-11-01

A drop falling onto a solid substrate will disintegrate into smaller parts when its impact velocity exceeds the so called critical velocity for splashing. Under these circumstances, the very thin liquid sheet ejected tangentially to the solid after the drop touches the substrate, lifts off as a consequence of the aerodynamic forces exerted on it and finally breaks into smaller droplets, violently ejected radially outwards, provoking the splash. Here, the tangential deceleration experienced by the fluid entering the thin liquid sheet is investigated making use of boundary layer theory. The velocity component tangent to the solid, computed using potential flow theory provides the far field boundary condition as well as the pressure gradient for the boundary layer equations. The structure of the flow permits to find a self similar solution of the boundary layer equations. This solution is then used to calculate the boundary layer thickness at the root of the lamella as well as the shear stress at the wall. The splash model presented in, which is slightly modified to account for the results obtained from the boundary layer analysis, provides a very good agreement between the measurements and the predicted values of the critical velocity for the splash.

A simple and efficient outflow boundary condition for the incompressible Navier–Stokes equations

Directory of Open Access Journals (Sweden)

Yibao Li

2017-01-01

Full Text Available Many researchers have proposed special treatments for outlet boundary conditions owing to lack of information at the outlet. Among them, the simplest method requires a large enough computational domain to prevent or reduce numerical errors at the boundaries. However, an efficient method generally requires special treatment to overcome the problems raised by the outlet boundary condition used. For example, mass flux is not conserved and the fluid field is not divergence-free at the outlet boundary. Overcoming these problems requires additional computational cost. In this paper, we present a simple and efficient outflow boundary condition for the incompressible Navier–Stokes equations, aiming to reduce the computational domain for simulating flow inside a long channel in the streamwise direction. The proposed outflow boundary condition is based on the transparent equation, where a weak formulation is used. The pressure boundary condition is derived by using the Navier–Stokes equations and the outlet flow boundary condition. In the numerical algorithm, a staggered marker-and-cell grid is used and temporal discretization is based on a projection method. The intermediate velocity boundary condition is consistently adopted to handle the velocity–pressure coupling. Characteristic numerical experiments are presented to demonstrate the robustness and accuracy of the proposed numerical scheme. Furthermore, the agreement of computational results from small and large domains suggests that our proposed outflow boundary condition can significantly reduce computational domain sizes.

Component nuclear containment structure

International Nuclear Information System (INIS)

Harstead, G.A.

1979-01-01

The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

Study of some properties of point defects in grain boundaries

International Nuclear Information System (INIS)

Martin, Georges

1973-01-01

With the aim of deducing simple informations on the grain boundary core structure, we investigated self diffusion under hydrostatic pressure, impurity diffusion (In and Au), electromigration (Sb) along certain types of grain boundaries in Ag bicrystals, and the Moessbauer effect of 57 Co located in the grain boundaries of polycrystalline Be. Our results lead to the following conclusions: the formation of a vacancy like defects is necessary to grain boundary diffusion; solute atoms may release most of their elastic energy of dissolution as they segregate at the boundary; in an electrical field, the drift of Sb ions parallel to the boundary takes place toward the anode as in the bulk. The force on the grain boundary ions is larger than in the bulk; Moessbauer spectroscopy revealed the formation of Co-rich aggregates, which may proves important in the study of early stages of grain boundary precipitation. (author) [fr

Inspection of Nuclear Power Plant Containment Structures

Energy Technology Data Exchange (ETDEWEB)

Graves, H.L.; Naus, D.J.; Norris, W.E.

1998-12-01

Safety-related nuclear power plant (NPP) structures are designed to withstand loadings from a number of low-probability external and interval events, such as earthquakes, tornadoes, and loss-of-coolant accidents. Loadings incurred during normal plant operation therefore generally are not significant enough to cause appreciable degradation. However, these structures are susceptible to aging by various processes depending on the operating environment and service conditions. The effects of these processes may accumulate within these structures over time to cause failure under design conditions, or lead to costly repair. In the late 1980s and early 1990s several occurrences of degradation of NPP structures were discovered at various facilities (e.g., corrosion of pressure boundary components, freeze- thaw damage of concrete, and larger than anticipated loss of prestressing force). Despite these degradation occurrences and a trend for an increasing rate of occurrence, in-service inspection of the safety-related structures continued to be performed in a somewhat cursory manner. Starting in 1991, the U.S. Nuclear Regulatory Commission (USNRC) published the first of several new requirements to help ensure that adequate in-service inspection of these structures is performed. Current regulatory in-service inspection requirements are reviewed and a summary of degradation experience presented. Nondestructive examination techniques commonly used to inspect the NPP steel and concrete structures to identify and quantify the amount of damage present are reviewed. Finally, areas where nondestructive evaluation techniques require development (i.e., inaccessible portions of the containment pressure boundary, and thick heavily reinforced concrete sections are discussed.

LBA-ECO LC-01 National, Provincial, and Park Boundaries, Ecuador

Data.gov (United States)

National Aeronautics and Space Administration — This data set contains the national and provincial boundaries of Ecuador as well as the boundaries of two national parks: the Cuyabeno Wildlife Reserve and the...

LBA-ECO LC-01 National, Provincial, and Park Boundaries, Ecuador

Data.gov (United States)

National Aeronautics and Space Administration — ABSTRACT: This data set contains the national and provincial boundaries of Ecuador as well as the boundaries of two national parks: the Cuyabeno Wildlife Reserve and...

On some boundary value problems in quantum statistical mechanics