WorldWideScience

Sample records for containment systems experiment

  1. Practical experience with a local verification system for containment and surveillance sensors

    International Nuclear Information System (INIS)

    Lauppe, W.D.; Richter, B.; Stein, G.

    1984-01-01

    With the growing number of nuclear facilities and a number of large commercial bulk handling facilities steadily coming into operation the International Atomic Energy Agency is faced with increasing requirements as to reducing its inspection efforts. One means of meeting these requirements will be to deploy facility based remote interrogation methods for its containment and surveillance instrumentation. Such a technical concept of remote interrogation was realized through the so-called LOVER system development, a local verification system for electronic safeguards seal systems. In the present investigations the application was extended to radiation monitoring by introducing an electronic interface between the electronic safeguards seal and the neutron detector electronics of a waste monitoring system. The paper discusses the safeguards motivation and background, the experimental setup of the safeguards system and the performance characteristics of this LOVER system. First conclusions can be drawn from the performance results with respect to the applicability in international safeguards. This comprises in particular the definition of design specifications for an integrated remote interrogation system for various types of containment and surveillance instruments and the specifications of safeguards applications employing such a system

  2. Phase equilibria for mixtures containing nonionic surfactant systems: Modeling and experiments

    International Nuclear Information System (INIS)

    Shin, Moon Sam; Kim, Hwayong

    2008-01-01

    Surfactants are important materials with numerous applications in the cosmetic, pharmaceutical, and food industries due to inter-associating and intra-associating bond. We present a lattice fluid equation-of-state that combines the quasi-chemical nonrandom lattice fluid model with Veytsman statistics for (intra + inter) molecular association to calculate phase behavior for mixtures containing nonionic surfactants. We also measured binary (vapor + liquid) equilibrium data for {2-butoxyethanol (C 4 E 1 ) + n-hexane} and {2-butoxyethanol (C 4 E 1 ) + n-heptane} systems at temperatures ranging from (303.15 to 323.15) K. A static apparatus was used in this study. The presented equation-of-state correlated well with the measured and published data for mixtures containing nonionic surfactant systems

  3. Critical experiments on an enriched uranium solution system containing periodically distributed strong thermal neutron absorbers

    International Nuclear Information System (INIS)

    Rothe, R.E.

    1996-01-01

    A series of 62 critical and critical approach experiments were performed to evaluate a possible novel means of storing large volumes of fissile solution in a critically safe configuration. This study is intended to increase safety and economy through use of such a system in commercial plants which handle fissionable materials in liquid form. The fissile solution's concentration may equal or slightly exceed the minimum-critical-volume concentration; and experiments were performed for high-enriched uranium solution. Results should be generally applicable in a wide variety of plant situations. The method is called the 'Poisoned Tube Tank' because strong neutron absorbers (neutron poisons) are placed inside periodically spaced stainless steel tubes which separate absorber material from solution, keeping the former free of contamination. Eight absorbers are investigated. Both square and triangular pitched lattice patterns are studied. Ancillary topics which closely model typical plant situations are also reported. They include the effect of removing small bundles of absorbers as might occur during inspections in a production plant. Not taking the tank out of service for these inspections would be an economic advantage. Another ancillary topic studies the effect of the presence of a significant volume of unpoisoned solution close to the Poisoned Tube Tank on the critical height. A summary of the experimental findings is that boron compounds were excellent absorbers, as expected. This was true for granular materials such as Gerstley Borate and Borax; but it was also true for the flexible solid composed of boron carbide and rubber, even though only thin sheets were used. Experiments with small bundles of absorbers intentionally removed reveal that quite reasonable tanks could be constructed that would allow a few tubes at a time to be removed from the tank for inspection without removing the tank from production service

  4. 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

  5. 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.

  6. Advanced Containment System

    Science.gov (United States)

    Kostelnik, Kevin M.; Kawamura, Hideki; Richardson, John G.; Noda, Masaru

    2004-10-12

    An advanced containment system for containing buried waste and associated leachate. A trench is dug on either side of the zone of interest containing the buried waste so as to accommodate a micro tunnel boring machine. A series of small diameter tunnels are serially excavated underneath the buried waste. The tunnels are excavated by the micro tunnel boring machine at a consistent depth and are substantially parallel to each other. As tunneling progresses, steel casing sections are connected end to end in the excavated portion of the tunnel so that a steel tube is formed. Each casing section has complementary interlocking structure running its length that interlocks with complementary interlocking structure on the adjacent casing section. Thus, once the first tube is emplaced, placement of subsequent tubes is facilitated by the complementary interlocking structure on the adjacent, previously placed, casing sections.

  7. 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

  8. Operational experience with SLAC's beam containment electronics

    International Nuclear Information System (INIS)

    Constant, T.N.; Crook, K.; Heggie, D.

    1977-03-01

    Considerable operating experience was accumulated at SLAC with an extensive electronic system for the containment of high power accelerated beams. Average beam power at SLAC can approach 900 kilowatts with the potential for burning through beam stoppers, protection collimators, and other power absorbers within a few seconds. Fast, reliable, and redundant electronic monitoring circuits have been employed to provide some of the safeguards necessary for minimizing the risk to personnel. The electronic systems are described, and the design philosophy and operating experience are discussed

  9. The advanced containment experiments (ACE) Project

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Ritzman, R.; Merilo, M.; Rahn, F.; Machiels, A.

    1992-01-01

    The overall structure and content of the ACE Project, which has been obtaining experimental data in four key areas of LWR severe accident technology are described. The key areas consist of filtration systems for vented containment concepts, radioiodine behavior in containment, the interaction of molten core material with structural concrete, and the use of water to terminate the core-concrete interaction process. Experiment procedures used in each phase of the work are summarized and the principal results and conclusions developed to date are discussed

  10. 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

  11. Development and operational experiences of an automated remote inspection system for interior of primary containment vessel of a BWR

    International Nuclear Information System (INIS)

    Ozaki, N.; Chikara, S.; Fumio, T.; Katsuhiro, M.; Katsutoshi, S.; Ken-Ichiro, S.; Masaaki, F.; Masayoshi, S.

    1983-01-01

    A prototype was developed for an automated remote inspection system featuring continuous monitoring of the working status of major components inside the primary containment vessel of a boiling water reactor. This inspection system consists of four units, or vehicles, which are towed by a trolley chain along a monorail; a complex coaxial cable for data transmission and for power supply; and an operator's console. A TV camera, microphone, thermometer, hygrometer, and ionization chamber are mounted on the various units. After several months' testing under high-ambient temperature, the system was installed in the Tokai-2 power station of Japan Atomic Power Company for in situ tests

  12. Prenatal Experiences of Containment in the Light of Bion's Model of Container/Contained

    Science.gov (United States)

    Maiello, Suzanne

    2012-01-01

    This paper explores the idea of possible proto-experiences of the prenatal child in the context of Bion's model of container/contained. The physical configuration of the embryo/foetus contained in the maternal uterus represents the starting point for an enquiry into the unborn child's possible experiences of its state of being contained in a…

  13. 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

  14. Nuclear steam system containment

    International Nuclear Information System (INIS)

    Jabsen, F.

    1980-01-01

    An improved containment used for radiation shielding and pressure suppression comprising a dry well includes a pressure vessel, a plurality of concentric wall means, said plurality of concentric wall means defining at least three annular regions about said dry well. A first annular region provides the containment used for radiation shielding, a second annular region is substantially dry, a third annular region provides a wet well for relieving fluid pressure released from the pressure vessel into the dry well. Pipe connection means extend in the wet well from the dry well, a pool of liquid is disposed to partially fill said third annular region, the upper end portion of the second and third annular regions having an enclosure, and a plurality of baffle plates extending vertically downward from said enclosure in said third annular region into said pool of liquid so as to circumferentially divide the upper portion of said third annular region into a plurality of circumferential upper portions

  15. Seal containment system

    International Nuclear Information System (INIS)

    Kugler, R.W.; Gerkey, K.S.; Kasner, W.H.

    1978-01-01

    An automated system for transporting nuclear fuel elements between fuel element assembly stations without contaminating the area outside the sealed assembly stations is described. The system comprises a plurality of assembly stations connected together by an elongated horizontal sealing mechanism and an automatic transport mechanism for transporting a nuclear fuel element in a horizontal attitude between the assembly stations while the open end of the fuel element extends through the sealing mechanism into the assembly station enclosure. The sealing mechanism allows the fuel element to be advanced by the transport mechanism while limiting the escape of radioactive particles from within the assembly station enclosure. 4 claims, 6 figures

  16. Authenticated Secure Container System (ASCS)

    International Nuclear Information System (INIS)

    1991-01-01

    Sandia National Laboratories developed an Authenticated Secure Container System (ASCS) for the International Atomic Energy Agency (IAEA). Agency standard weights and safeguards samples can be stored in the ASCS to provide continuity of knowledge. The ASCS consists of an optically clear cover, a base containing the Authenticated Item Monitoring System (AIMS) transmitter, and the AIMS receiver unit for data collection. The ASCS will provide the Inspector with information concerning the status of the system, during a surveillance period, such as state of health, tampering attempts, and movement of the container system. The secure container is located inside a Glove Box with the receiver located remotely from the Glove Box. AIMS technology uses rf transmission from the secure container to the receiver to provide a record of state of health and tampering. The data is stored in the receiver for analysis by the Inspector during a future inspection visit. 2 refs

  17. Disposal/storage container development experience

    International Nuclear Information System (INIS)

    Morrow, R.W. Jr.; Van Hoesen, S.D.; Fowler, E.; Barreira, D.G.; Emmett, R.W.

    1988-01-01

    Developmental work is currently underway at the Oak Ridge National Laboratory to design and manufacture a radioactive waste container suitable for both storage and disposal of radioactive wastes. The container is designed to fulfill the Department of Energy and Nuclear Regulatory Commission requirements for on-site storage, as well as the Nuclear Regulatory Commission's requirements for high integrity containers. The project also involves meeting the strict design and manufacturing ANSI/ASME NQA-1 guidelines. Special provisions of the container include a double containment system, with the inner barrier being corrosion resistant, the capability to monitor the internal cavity of the container, and off-gas venting capability. Further, yet related developmental work includes evaluating the cask for other varied uses, such as a processing cask, an ALARA shield, and even the possibility of Department of Transportation approval for an over-the-road transport cask

  18. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Counts, C.A.

    1984-06-01

    The Pacific Northwest Laboratory (PNL) is reviewing available information on containment systems design, operating experience, and related research as part of a project being conducted by the Division of Systems Integration, US Nuclear Regulatory Commission. The basic objective of this work is to collect and consolidate data relevant to assessing the functional performance of containment isolation systems and to use this data to the extent possible to characterize containment isolation system reliability for selected reference designs. This paper summarizes the results from initial efforts which focused on collection of data from available documents and briefly describes detailed review and analysis efforts which commenced recently. 5 references

  19. Accounting for variation in designing greenhouse experiments with special reference to greenhouses containing plants on conveyor systems

    Science.gov (United States)

    2013-01-01

    Background There are a number of unresolved issues in the design of experiments in greenhouses. They include whether statistical designs should be used and, if so, which designs should be used. Also, are there thigmomorphogenic or other effects arising from the movement of plants on conveyor belts within a greenhouse? A two-phase, single-line wheat experiment involving four tactics was conducted in a conventional greenhouse and a fully-automated phenotyping greenhouse (Smarthouse) to investigate these issues. Results and discussion Analyses of our experiment show that there was a small east–west trend in total area of the plants in the Smarthouse. Analyses of the data from three multiline experiments reveal a large north–south trend. In the single-line experiment, there was no evidence of differences between trios of lanes, nor of movement effects. Swapping plant positions during the trial was found to decrease the east–west trend, but at the cost of increased error variance. The movement of plants in a north–south direction, through a shaded area for an equal amount of time, nullified the north–south trend. An investigation of alternative experimental designs for equally-replicated experiments revealed that generally designs with smaller blocks performed best, but that (nearly) trend-free designs can be effective when blocks are larger. Conclusions To account for variation in microclimate in a greenhouse, using statistical design and analysis is better than rearranging the position of plants during the experiment. For the relocation of plants to be successful requires that plants spend an equal amount of time in each microclimate, preferably during comparable growth stages. Even then, there is no evidence that this will be any more precise than statistical design and analysis of the experiment, and the risk is that it will not be successful at all. As for statistical design and analysis, it is best to use either (i) smaller blocks, (ii) (nearly) trend

  20. Flow column experiments on the 152Eu migration in systems of loose sediments and water containing humic acids

    International Nuclear Information System (INIS)

    Klotz, D.; Wolf, M.

    2001-01-01

    Humic acid transport of 152 in non-binding loose sediments of different grain sizes was investigated using a groundwater of the tertiary lignite of Northern Germany with a high humic acid concentration. The migration experiments were carried out in flow columns at natural filter flow rates and natural flow lengths [de

  1. Docker Containers for Deep Learning Experiments

    OpenAIRE

    Gerke, Paul K.

    2017-01-01

    Deep learning is a powerful tool to solve problems in the area of image analysis. The dominant compute platform for deep learning is Nvidia’s proprietary CUDA, which can only be used together with Nvidia graphics cards. The nivida-docker project allows exposing Nvidia graphics cards to docker containers and thus makes it possible to run deep learning experiments in docker containers.In our department, we use deep learning to solve problems in the area of medical image analysis and use docker ...

  2. CONTAIN assessment of the NUPEC mixing experiments

    International Nuclear Information System (INIS)

    Stamps, D.W.

    1995-08-01

    The ability of the CONTAIN code to predict the thermal hydraulics of five experiments performed in the NUPEC 1/4-scale model containment was assessed. These experiments simulated severe accident conditions in a nuclear power plant in which helium (as a nonflammable substitute for hydrogen) and steam were coinjected at different locations in the facility with and without the concurrent injection of water sprays in the dome. Helium concentrations, gas temperatures and pressures, and wall temperatures were predicted and compared with the data. The use of different flow solvers, nodalization schemes, and analysis methods for the treatment of water sprays was emphasized. As a result, a general procedure was suggested for lumped-parameter code analyses of problems in which the thermal hydraulics are dominated by water sprays

  3. Building a secondary containment system

    Energy Technology Data Exchange (ETDEWEB)

    Broder, M.F.

    1994-10-01

    Retail fertilizer and pesticide dealers across the United States are installing secondary containment at their facilities or are seriously considering it. Much of this work is in response to new state regulations; however, many dealers not facing new regulations are upgrading their facilities to reduce their liability, lower their insurance costs, or comply with anticipated regulations. The Tennessee Valley Authority`s (TVA) National Fertilizer and Environmental Research Center (NFERC) has assisted dealers in 22 states in retrofitting containment to their facilities. Simultaneous improvements in the operational efficiency of the facilities have been achieved at many of the sites. This paper is based on experience gained in that work and details the rationale used in planning secondary containment and facility modifications.

  4. Design experiments for a vented containment

    International Nuclear Information System (INIS)

    Hesboel, R.

    1985-01-01

    A filtered containment venting system, operable late in 1985, is currently under installation at the Barsebaeck twin nuclear power station in Sweden. The filter unit, which communicates with the containments of both reactor units, but is separated from them by rupture discs, consists of a concrete bed, 40 m high and 20 m in diameter, filled with gravel of grain size 25-35 mm. The performance of the gravel bed under such accident conditions which might lead to an activation of this safeguard system has been the subject for investigation within the FILTRA project. These investigations have shown that the gravel bed acts as: an expansion volume for decreasing gas pressure and increasing gas residence time, a heat sink for condensing steam, an excellent filter medium for removing aerosols and elemental iodine, and a sump volume for collecting radioactive condensate. The results from iodine retention studies in gravel beds are mainly considered

  5. Reliability analysis of containment isolation systems

    International Nuclear Information System (INIS)

    Pelto, P.J.; Ames, K.R.; Gallucci, R.H.

    1985-06-01

    This report summarizes the results of the Reliability Analysis of Containment Isolation System Project. Work was performed in five basic areas: design review, operating experience review, related research review, generic analysis and plant specific analysis. Licensee Event Reports (LERs) and Integrated Leak Rate Test (ILRT) reports provided the major sources of containment performance information used in this study. Data extracted from LERs were assembled into a computer data base. Qualitative and quantitative information developed for containment performance under normal operating conditions and design basis accidents indicate that there is room for improvement. A rough estimate of overall containment unavailability for relatively small leaks which violate plant technical specifications is 0.3. An estimate of containment unavailability due to large leakage events is in the range of 0.001 to 0.01. These estimates are dependent on several assumptions (particularly on event duration times) which are documented in the report

  6. Remote operation system for container

    International Nuclear Information System (INIS)

    Nakahara, Hirotaka; Hayata, Takashi; Kajiyama, Shigeru; Takahashi, Fuminobu

    1998-01-01

    The present invention provides a remote operation system for conducting operation with operation reaction for the inside of a container filled with water (liquid), such as of inner walls and inner structural materials of a BWR type reactor. Namely, a swimming robot comprises a swimming device swimming in the liquid and an attaching/detaching device for holding/releasing the handling robot. A control device remotely operate the swimming robot and the handling robot by way of a cable. A cable processing device takes up or dispenses the cable. In addition, when the swimming robot grasps the handling robot by the attaching/detaching device, the swimming robot transmits an operation instruction sent from the control device by way of the cable to the handling robot. After the attaching/detaching device of the swimming robot releases the handling robot, the handling robot operates based on the transmitted operation instruction. It is preferable that the handling robot has an adsorptive moving device for moving itself while being adsorbed on the wall surface of the container. (I.S.)

  7. Safety system for reactor container

    International Nuclear Information System (INIS)

    Shimizu, Miwako; Seki, Osamu; Mano, Takio.

    1995-01-01

    A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

  8. Halon 1301 protection system for nuclear containments

    International Nuclear Information System (INIS)

    McHale, E.T.

    1981-01-01

    Halon 1301 can provide protection against any combustion hazard that hydrogen gas might present in an LWR containment following a loss-of-coolant accident. A development program was conducted, comprising analytical study, laboratory experiments and large-scale testing, to define the requirements for a Halon 1301 system and to examine certain operational problems that were hypothesized. Some results of the study are presented in this paper

  9. Batch experiments for assessing the sorption/desorption characteristics of 152Eu in systems of loose sediments and water containing humic acids

    International Nuclear Information System (INIS)

    Klotz, D.

    2001-01-01

    The 152 Eu distribution coefficients of the sorption and desorption of non-binding loose sediments of different grain sizes are investigated using a groundwater of tertiary lignite from Northern Germany which contains high concentrations of humic acids. The batch experiments were carried out with a ratio of 2.5cm 3 /g of solution volume to sediment mass, without mixing [de

  10. Wellbore Completion Systems Containment Breach Solution Experiments at a Large Scale Underground Research Laboratory : Sealant placement & scale-up from Lab to Field

    Science.gov (United States)

    Goodman, H.

    2017-12-01

    This investigation seeks to develop sealant technology that can restore containment to completed wells that suffer CO2 gas leakages currently untreatable using conventional technologies. Experimentation is performed at the Mont Terri Underground Research Laboratory (MT-URL) located in NW Switzerland. The laboratory affords investigators an intermediate-scale test site that bridges the gap between the laboratory bench and full field-scale conditions. Project focus is the development of CO2 leakage remediation capability using sealant technology. The experimental concept includes design and installation of a field scale completion package designed to mimic well systems heating-cooling conditions that may result in the development of micro-annuli detachments between the casing-cement-formation boundaries (Figure 1). Of particular interest is to test novel sealants that can be injected in to relatively narrow micro-annuli flow-paths of less than 120 microns aperture. Per a special report on CO2 storage submitted to the IPCC[1], active injection wells, along with inactive wells that have been abandoned, are identified as one of the most probable sources of leakage pathways for CO2 escape to the surface. Origins of pressure leakage common to injection well and completions architecture often occur due to tensile cracking from temperature cycles, micro-annulus by casing contraction (differential casing to cement sheath movement) and cement sheath channel development. This discussion summarizes the experiment capability and sealant testing results. The experiment concludes with overcoring of the entire mock-completion test site to assess sealant performance in 2018. [1] IPCC Special Report on Carbon Dioxide Capture and Storage (September 2005), section 5.7.2 Processes and pathways for release of CO2 from geological storage sites, page 244

  11. Demonstration of an Emergency Containment System

    International Nuclear Information System (INIS)

    Flanagan, T.M.; Rogers, M.L.; Wilkes, W.R.

    1978-01-01

    A system called an Emergency Containment System (ECS) to be used for tertiary containment of tritium was reported at the 13th Air Cleaning Conference. This system was part of the Tritium Effluent Control Laboratory then under construction at Mound Facility. A series of experiments has recently been conducted to evaluate the performance of an ECS in capturing tritium accidentally released into an operating laboratory. The ECS is an automatically actuated laboratory air detritiation system utilizing a catalytic oxidation reactor and presaturated oxide adsorption/exchange columns. In the event of an accidental release of tritium into the laboratory, the ECS is automatically activated, and quick-acting pneumatic dampers divert the laboratory air supply and exhaust through the ECS until room concentrations are returned to safe operating levels. The results of the experiments have shown that a tertiary containment of tritium is feasible. In the event of a catastrophic accident, the ECS is capable of preventing the release of a large quantity of tritium to the environment

  12. Seoul's greenbelt: an experiment in urban containment

    Science.gov (United States)

    David N. Bengston; Youn Yeo-Chang

    2005-01-01

    Urban containment policies are considered by some to be a promising approach to growth management. The greenbelt-based urban containment policy of Seoul, Republic of Korea is examined as a case study. Seoul's greenbelt has generated both significant social costs and benefits. Korea's greenbelt policy is currently being revised, largely due to pressure from...

  13. AREVA’s Containment Venting Technologies and Experience Worldwide

    Energy Technology Data Exchange (ETDEWEB)

    Welker, M.

    2015-07-01

    The AREVA Filtered Containment Venting System (FCVS) is a product family that minimizes the environmental impact in case of a severe accident in a nuclear power plant (NPP). Our experience is based on a large-scale test and qualification program as well as on the design, licensing and installation of more than 80 projects worldwide. The product family provides flexibility regarding the adaptation to respective accident scenarios, applicable codes and standards, seismic design, supply chain, implementation and localization. AREVA has broad experience of managing fleet supplies, successful support of licensing and cooperating with original equipment manufacturers (OEMs) of pressurized and boiling water reactors (PWR and BWR). (Author)

  14. Explicit Finite Element Modeling of Multilayer Composite Fabric for Gas Turbine Engine Containment Systems, Phase II. Part 3; Material Model Development and Simulation of Experiments

    Science.gov (United States)

    Simmons, J.; Erlich, D.; Shockey, D.

    2009-01-01

    A team consisting of Arizona State University, Honeywell Engines, Systems & Services, the National Aeronautics and Space Administration Glenn Research Center, and SRI International collaborated to develop computational models and verification testing for designing and evaluating turbine engine fan blade fabric containment structures. This research was conducted under the Federal Aviation Administration Airworthiness Assurance Center of Excellence and was sponsored by the Aircraft Catastrophic Failure Prevention Program. The research was directed toward improving the modeling of a turbine engine fabric containment structure for an engine blade-out containment demonstration test required for certification of aircraft engines. The research conducted in Phase II began a new level of capability to design and develop fan blade containment systems for turbine engines. Significant progress was made in three areas: (1) further development of the ballistic fabric model to increase confidence and robustness in the material models for the Kevlar(TradeName) and Zylon(TradeName) material models developed in Phase I, (2) the capability was improved for finite element modeling of multiple layers of fabric using multiple layers of shell elements, and (3) large-scale simulations were performed. This report concentrates on the material model development and simulations of the impact tests.

  15. The buffer/container experiment design and construction report

    Energy Technology Data Exchange (ETDEWEB)

    Chandler, N.A.; Wan, A.W.L.; Roach, P.J

    1998-03-01

    The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment

  16. The buffer/container experiment design and construction report

    International Nuclear Information System (INIS)

    Chandler, N.A.; Wan, A.W.L.; Roach, P.J.

    1998-03-01

    The Buffer/Container Experiment was a full-scale in situ experiment, installed at a depth of 240 m in granitic rock at AECL's Underground Research Laboratory (URL). The experiment was designed to examine the performance of a compacted sand-bentonite buffer material under the influences of elevated temperature and in situ moisture conditions. Buffer material was compacted in situ into a 5-m-deep, 1.24-m-diameter borehole drilled into the floor of an excavation. A 2.3-m long heater, representative of a nuclear fuel waste container, was placed within the buffer, and instrumentation was installed to monitor changes in buffer moisture conditions, temperature and stress. The experiment was sealed at the top of the borehole and restrained against vertical displacement. Instrumentation in the rock monitored pore pressures, temperatures and rock displacement. The heater was operated at a constant power of 1200 W, which provided a heater skin temperature of approximately 85 degrees C. Experiment construction and installation required two years, followed by two and a half years of heater operation and two years of monitoring the rock conditions during cooling. The construction phase of the experiment included the design, construction and testing of a segmental heater and controller, geological and hydrogeological characterization of the rock, excavation of the experiment room, drilling of the emplacement borehole using high pressure water, mixing and in situ compaction of buffer material, installation of instrumentation in the rock, buffer and on the heater, and the construction of concrete curb and steel vertical restraint system at the top of emplacement borehole. Upon completion of the experiment, decommissioning sampling equipment was designed and constructed and sampling methods were developed which allowed approximately 2000 samples of buffer material to be taken over a 12-day period. Quality assurance procedures were developed for all aspects of experiment construction

  17. Understanding aging in containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.J.

    1993-01-01

    A study has been performed to assess the effects of aging in nuclear power plant containment cooling systems. Failure records from national databases, as well as plant specific data were reviewed and analyzed to identify aging characteristics for this system. The predominant aging mechanisms were determined, along with the most frequently failed components and their associated failure modes. This paper discusses the aging mechanisms present in the containment spray system and the containment fan cooler system, which are two systems used to provide the containment cooling function. The failure modes, along with the relative frequency of each is also discussed

  18. FFTF-containment air-cleaning system

    International Nuclear Information System (INIS)

    Mahaffey, M.K.; Stepnewski, D.D.

    1981-01-01

    The FFTF Containment can accommodate all design basis events and the hypothetical core disruptive accident with adequate margin without a venting or purging system; however, in concert with the development objective, a system was designed and constructed to evaluate technology related to containment atmosphere venting and cleanup functions. The system can be used to purge high H 2 concentrations or to vent excessive containment pressure. In either case containment atmosphere is exhausted through an aqueous scrubber system consisting of a venturi scrubber and fibrous filter bank

  19. Remote container monitoring and surveillance systems

    International Nuclear Information System (INIS)

    Resnik, W.M.; Kadner, S.P.

    1995-01-01

    Aquila Technologies Group is developing a monitoring and surveillance system to monitor containers of nuclear materials. The system will both visually and physically monitor the containers. The system is based on the combination of Aquila's Gemini All-Digital Surveillance System and on Aquila's AssetLAN trademark asset tracking technology. This paper discusses the Gemini Digital Surveillance system as well as AssetLAN technology. The Gemini architecture with emphasis on anti-tamper security features is also described. The importance of all-digital surveillance versus other surveillance methods is also discussed. AssetLAN trademark technology is described, emphasizing the ability to continually track containers (as assets) by location utilizing touch memory technology. Touch memory technology provides unique container identification, as well as the ability to store and retrieve digital information on the container. This information may relate to container maintenance, inspection schedules, and other information. Finally, this paper describes the combination of the Gemini system with AssetLAN technology, yielding a self contained, container monitoring and area/container surveillance system. Secure container fixture design considerations are discussed. Basic surveillance review functions are also discussed

  20. Igloo containment system for improvised explosive devices

    International Nuclear Information System (INIS)

    Dyckes, G.W.

    1980-09-01

    A method for containing or partially containing the blast and dispersal of radioactive particulate from improvised explosive devices is described. The containment system is restricted to devices located in fairly open areas at ground level, e.g., devices concealed in trucks, vans, transportainers, or small buildings which are accessible from all sides

  1. NucleDyne's passive containment system

    International Nuclear Information System (INIS)

    Falls, O.B. Jr.; Kleimola, F.W.

    1987-01-01

    A simple definition of the passive containment system is that it is a total safeguards system for light water reactors designed to prevent and contain any accidental release of radioactivity. Its passive features utilize the natural laws of physics and thermodynamics. The system encompasses three basic containments constructed as one integrated structure on the reactor building foundation. The primary containment encloses the reactor pressure vessel and coolant system and passive engineered safety systems and components. Auxiliary containment enclosures house auxiliary systems and components. Secondary containment (the reactor building), housing the primary and auxiliary containment structures, provides a second containment barrier as added defense-in-depth against leakage of radioactivity for all accidents assumed by the industry. The generic features of the passive containment system are applicable to both the boiling water reactors and the pressurized water reactors as standardized features for all power ranges. These features provide for a zero source term, the industry's ultimate safety goal. This paper relates to a four-loop pressurized water reactor

  2. Containments for consolidated nuclear steam systems

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1978-01-01

    A containment system for a consolidated nuclear steam system incorporating a nuclear core, steam generator and reactor coolant pumps within a single pressure vessel is described which is designed to provide radiation shielding and pressure suppression. Design details, including those for the dry well and wet well of the containment, are given. (UK)

  3. Status of the LWR aerosol containment experiments (LACE) program

    International Nuclear Information System (INIS)

    Bloom, G.R.; Dickinson, D.R.; Hilliard, R.K.; McCormack, J.D.; Muhlestein, L.D.; Rahn, F.J.

    1985-01-01

    The LACE program, sponsored by an international consortium, is investigating inherent aerosol behavior for three postulated high consequence accident sequences; the containment bypass or V-sequence, failure to isolate containment, and delayed containment failure. Six large-scale tests are described which focus on these accident situations and which will be completed in the Containment Systems Test Facility at the Hanford Engineering Development Laboratory. The aerosol generation systems used to generate soluble and insoluble aerosols for the large-scale tests are described. The report then focuses on those tests which deal with the containment bypass accident sequence. Test results are presented and discussed for three containment bypass scoping tests

  4. 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

  5. The Soviet RBMK-1000 containment system

    International Nuclear Information System (INIS)

    Joosten, J.K.

    1988-01-01

    Following the accident in April, 1986, considerable attention was focused on the failure of the containment at the Chernobyl RBMK-1000 nuclear power plant. Conflicting statements arose regarding the nature of the plant's containment system primarily because of terminology differences, translation difficulties and lack of reliable information. This article, based on reports and briefings by the Soviet delegation, during the post-accident review meetings in Vienna and prior publications is intended to clarify perceptions of the Soviet RMBK-1000 nuclear power plant containment system design, and its relevance to containment management concepts. (author)

  6. MELCOR 1.8.3 assessment: CSE containment spray experiments

    International Nuclear Information System (INIS)

    Kmetyk, L.N.

    1994-12-01

    MELCOR is a fully integrated, engineering-level computer code, being developed at Sandia National Laboratories for the USNRC, that models the entire spectrum of severe accident phenomena in a unified framework for both BWRs and PWRS. As part, of an ongoing assessment program, the MELCOR computer code has been used to analyze a series of containment spray tests performed in the Containment Systems Experiment (CSE) vessel to evaluate the performance of aqueous sprays as a means of decontaminating containment atmospheres. Basecase MELCOR results are compared with test data, and a number of sensitivity studies on input modelling parameters and options in both the spray package and the associated aerosol washout and atmosphere decontamination by sprays modelled in the radionuclide package have been done. Time-step and machine-dependency calculations were done to identify whether any numeric effects exist in these CSE assessment analyses. A significant time-step dependency due to an error in the spray package coding was identified and eliminated. A number of other code deficiencies and inconveniences also are noted

  7. System for inspection of stacked cargo containers

    Science.gov (United States)

    Derenzo, Stephen [Pinole, CA

    2011-08-16

    The present invention relates to a system for inspection of stacked cargo containers. One embodiment of the invention generally comprises a plurality of stacked cargo containers arranged in rows or tiers, each container having a top, a bottom a first side, a second side, a front end, and a back end; a plurality of spacers arranged in rows or tiers; one or more mobile inspection devices for inspecting the cargo containers, wherein the one or more inspection devices are removeably disposed within the spacers, the inspection means configured to move through the spacers to detect radiation within the containers. The invented system can also be configured to inspect the cargo containers for a variety of other potentially hazardous materials including but not limited to explosive and chemical threats.

  8. Self-contained microfluidic systems: a review.

    Science.gov (United States)

    Boyd-Moss, Mitchell; Baratchi, Sara; Di Venere, Martina; Khoshmanesh, Khashayar

    2016-08-16

    Microfluidic systems enable rapid diagnosis, screening and monitoring of diseases and health conditions using small amounts of biological samples and reagents. Despite these remarkable features, conventional microfluidic systems rely on bulky expensive external equipment, which hinders their utility as powerful analysis tools outside of research laboratories. 'Self-contained' microfluidic systems, which contain all necessary components to facilitate a complete assay, have been developed to address this limitation. In this review, we provide an in-depth overview of self-contained microfluidic systems. We categorise these systems based on their operating mechanisms into three major groups: passive, hand-powered and active. Several examples are provided to discuss the structure, capabilities and shortcomings of each group. In particular, we discuss the self-contained microfluidic systems enabled by active mechanisms, due to their unique capability for running multi-step and highly controllable diagnostic assays. Integration of self-contained microfluidic systems with the image acquisition and processing capabilities of smartphones, especially those equipped with accessory optical components, enables highly sensitive and quantitative assays, which are discussed. Finally, the future trends and possible solutions to expand the versatility of self-contained, stand-alone microfluidic platforms are outlined.

  9. Passive containment system for a nuclear reactor

    International Nuclear Information System (INIS)

    Kleimola, F.W.

    1976-01-01

    A containment system is described 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 continuously maintained submerged in liquid. The primary containment vessel is restored to a high subatmospheric 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

  10. Simulation of containment atmosphere stratification experiment using local instantaneous description

    International Nuclear Information System (INIS)

    Babic, M.; Kljenak, I.

    2004-01-01

    An experiment on mixing and stratification in the atmosphere of a nuclear power plant containment at accident conditions was simulated with the CFD code CFX4.4. The original experiment was performed in the TOSQAN experimental facility. Simulated nonhomogeneous temperature, species concentration and velocity fields are compared to experimental results. (author)

  11. Passive containment system in high earthquake motion

    International Nuclear Information System (INIS)

    Kleimola, F.W.; Falls, O.B. Jr.

    1977-01-01

    High earthquake motion necessitates major design modifications in the complex of plant structures, systems and components in a nuclear power plant. Distinctive features imposed by seismic category, safety class and quality classification requirements for the high seismic ground acceleration loadings significantly reflect in plant costs. The design features in the Passive Containment System (PCS) responding to high earthquake ground motion are described

  12. Commissioning Ventilated Containment Systems in the Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    2008-08-01

    This Best Practices Guide focuses on the specialized approaches required for ventilated containment systems, understood to be all components that drive and control ventilated enclosures and local exhaust systems within the laboratory. Geared toward architects, engineers, and facility managers, this guide provides information about technologies and practices to use in designing, constructing, and operating operating safe, sustainable, high-performance laboratories.

  13. Active containment systems incorporating modified pillared clays

    International Nuclear Information System (INIS)

    Lundie, P.; McLeod, N.

    1997-01-01

    The application of treatment technologies in active containment systems provides a more advanced and effective method for the remediation of contaminated sites. These treatment technologies can be applied in permeable reactive walls and/or funnel and gate systems. The application of modified pillared clays in active containment systems provides a mechanism for producing permeable reactive walls with versatile properties. These pillared clays are suitably modified to incorporate reactive intercalatants capable of reacting with both a broad range of organic pollutants of varying molecular size, polarity and reactivity. Heavy metals can be removed from contaminated water by conventional ion-exchange and other reactive processes within the clay structure. Complex contamination problems can be addressed by the application of more than one modified clay on a site specific basis. This paper briefly describes the active containment system and the structure/chemistry of the modified pillared clay technology, illustrating potential applications of the in-situ treatment process for contaminated site remediation

  14. Evaluation of the nucledyne passive containment system

    International Nuclear Information System (INIS)

    1981-04-01

    This reports contains: (1) an evaluation by Gilbert/Commonwealth (G/C) of the NucleDyne passive Containment System (PCS) as that conceptual design is applied to a Westinghouse, two loop, Pressurized Water Reactor; (2) an evaluation by Westinghouse of two questions about the impact of the PCS on the Nuclear Steam Supply System (NSSS), which were posed by G/C and best answered by an NSSS vendor; and (3) replies to both the Gilbert/Commonwealth report and the Westinghoue report by NucleDyne Engineering Corporation

  15. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  16. The Mistra experiment for field containment code validation first results

    International Nuclear Information System (INIS)

    Caron-Charles, M.; Blumenfeld, L.

    2001-01-01

    The MISTRA facility is a large scale experiment, designed for the purpose of thermal-hydraulics multi-D codes validation. A short description of the facility, the set up of the instrumentation and the test program are presented. Then, the first experimental results, studying helium injection in the containment and their calculations are detailed. (author)

  17. Tritium-containment systems: a tradeoff study

    International Nuclear Information System (INIS)

    Folkers, C.L.; Cena, R.J.

    1978-01-01

    Various design parameters are evaluated that affect the performance of tritium-containment systems for fusion reactors. Our study included a review of such parameters as tritium forms, impurities, catalysts, adsorbents, getters, and as low as reasonably achievable principles. We organized these schemes, which can be considered for treating either air or inert atmospheres, so one could easily make orderly choices and tradeoffs for optimum performance. The relationships examined involved purification-system decontamination factors, flow rates, recycling and leakage, and environmental losses

  18. Review on experiments relating to primary containment vessel failure

    International Nuclear Information System (INIS)

    Suzuki, Hiroyuki; Okada, Hidetoshi; Uchida, Sunsuke; Naitoh, Masanori

    2015-01-01

    Experiments regarding failures of primary containment vessels (PCVs) are reviewed and remained issues to be investigated in the future are discussed. Experiments are categorized as those relating to criteria of PCV failures and to FP releases through breaches on PCV boundaries. In the experiments categorized as those relating to criteria of PCV failures, experiments with full-scale, scale models, and compounds used for sealing are surveyed. Experiments relating to an amount of radioactive fission products (FPs) trapped at breaches on PCV boundaries are also reviewed. As remained issues to be investigated in the future, two items are pointed out: Evaluating degradation behavior of PCV boundaries exposed to temperature and pressure from the failure onset criteria to far above them, and evaluating an amount of FPs trapped at breaches on PCV boundaries. (author)

  19. Study on vent containment filtering for the Spanish NPPS systems

    International Nuclear Information System (INIS)

    Peinado, A.; Serrano, C.; Garcia-Serrano, J. L.

    2013-01-01

    The study discusses filtering systems on the market, and its suppliers, taking into account aspects such as ease of integration into the current plant design, characteristics of the process of filtering, operational range, autonomy of the system, maintenance, qualification and proven experiences, among others. The study, also contains an analysis of sequences kind of accident that serve to define the design parameters of the system.

  20. ACE puts containment venting systems to the test

    International Nuclear Information System (INIS)

    Merilo, M.

    1990-01-01

    Filtered venting of reactor containments has received considerable attention recently as a method for avoiding containment failure due to overpressure during severe accidents. Several proposed filtration devices have been tested in the internationally sponsored Advanced Containment Experiments (ACE) programme, such that a self consistent comparison of the aerosol removal characteristics of these systems could be obtained. Considering the different design, requirements and operating conditions of the filter devices, a direct comparison is not possible, nor appropriate. Nevertheless, large scale models, using full scale elements of the various devices whenever feasible, have been tested with consistent mixtures of aerosols and carrier gases. (author)

  1. Detector correction in large container inspection systems

    CERN Document Server

    Kang Ke Jun; Chen Zhi Qiang

    2002-01-01

    In large container inspection systems, the image is constructed by parallel scanning with a one-dimensional detector array with a linac used as the X-ray source. The linear nonuniformity and nonlinearity of multiple detectors and the nonuniform intensity distribution of the X-ray sector beam result in horizontal striations in the scan image. This greatly impairs the image quality, so the image needs to be corrected. The correction parameters are determined experimentally by scaling the detector responses at multiple points with logarithm interpolation of the results. The horizontal striations are eliminated by modifying the original image data with the correction parameters. This method has proven to be effective and applicable in large container inspection systems

  2. Comparison of ANL containment codes with SNR-300 simulation experiments

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Wang, C.Y.; Fistedis, S.H.

    1976-01-01

    A comparison of REXCO and ICECO code predictions is made with data obtained from experiments of LMFBR excursion models. The comparisons are based on published results of tests conducted for the safety analysis of the SNR-300 fast breeder. The test configurations consist of a centrally located spherical source immersed in a pool of water which is encased in a cylindrical container. The cylinical walls of the container are prestressed by holddown bolts which span the two rigid ends. The space above the surface of the water within the container is occupied by air. Although certain aspects of the tests could not be simulated by the analytical models exactly, the comparison of results shows quite close agreement. The fact that the REXCO and ICECO codes involve different analytical formulations, their own close correspondence of results lends added credence to the value of analytical predictions

  3. DISPOSAL CONTAINER HANDLING SYSTEM DESCRIPTION DOCUMENT

    Energy Technology Data Exchange (ETDEWEB)

    E. F. Loros

    2000-06-30

    The Disposal Container Handling System receives and prepares new disposal containers (DCs) and transfers them to the Assembly Transfer System (ATS) or Canister Transfer System (CTS) for loading. The system receives the loaded DCs from ATS or CTS and welds the lids. When the welds are accepted the DCs are termed waste packages (WPs). The system may stage the WP for later transfer or transfer the WP directly to the Waste Emplacement/Retrieval System. The system can also transfer DCs/WPs to/from the Waste Package Remediation System. The Disposal Container Handling System begins with new DC preparation, which includes installing collars, tilting the DC upright, and outfitting the container for the specific fuel it is to receive. DCs and their lids are staged in the receipt area for transfer to the needed location. When called for, a DC is put on a cart and sent through an airlock into a hot cell. From this point on, all processes are done remotely. The DC transfer operation moves the DC to the ATS or CTS for loading and then receives the DC for welding. The DC welding operation receives loaded DCs directly from the waste handling lines or from interim lag storage for welding of the lids. The welding operation includes mounting the DC on a turntable, removing lid seals, and installing and welding the inner and outer lids. After the weld process and non-destructive examination are successfully completed, the WP is either staged or transferred to a tilting station. At the tilting station, the WP is tilted horizontally onto a cart and the collars removed. The cart is taken through an air lock where the WP is lifted, surveyed, decontaminated if required, and then moved into the Waste Emplacement/Retrieval System. DCs that do not meet the welding non-destructive examination criteria are transferred to the Waste Package Remediation System for weld preparation or removal of the lids. The Disposal Container Handling System is contained within the Waste Handling Building System

  4. Explosion testing for the container venting system

    International Nuclear Information System (INIS)

    Cashdollar, K.L.; Green, G.M.; Thomas, R.A.; Demiter, J.A.

    1993-01-01

    As part of the study of the hazards of inspecting nuclear waste stored at the Hanford Site, the US Department of Energy and Westinghouse Hanford Company have developed a container venting system to sample the gases that may be present in various metal drums and other containers. In support of this work, the US Bureau of Mines has studied the probability of ignition while drilling into drums and other containers that may contain flammable gas mixtures. The Westinghouse Hanford Company drilling procedure was simulated by tests conducted in the Bureau's 8-liter chamber, using the same type of pneumatic drill that will be used at the Hanford Site. There were no ignitions of near-stoichiometric hydrogen-air or methane-air mixtures during the drilling tests. The temperatures of the drill bits and lids were measured by an infrared video camera during the drilling tests. These measured temperatures are significantly lower than the ∼500 degree C autoignition temperature of uniformly heated hydrogen-air or the ∼600 degree C autoignition temperature of uniformly heated methane-air. The temperatures are substantially lower than the 750 degree C ignition temperature of hydrogen-air and 1,220 degree C temperature of methane-air when heated by a 1-m-diameter wire

  5. GOTHIC Simulation of Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Ha, Huiun; Kim, Hangon

    2013-01-01

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR

  6. GOTHIC Simulation of Passive Containment Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Kim, Hangon [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2013-05-15

    The performance of this system depends on the condensation of steam moving downward inside externally cooled vertical tubes. AES-2006: During a DBA, heat is removed by internally cooled vertical tubes, which are located in containment. We are currently developing the conceptual design of Innovative PWR, which is will be equipped with various passive safety features, including PCCS. We have plan to use internal heat exchanger (HX) type PCCS with concrete containment. In this case, the elevation of HXs is important to ensure the heat removal during accidents. In general, steam is lighter than air mixture in containment. So, steam may be collected at the upper side of containment. It means that higher elevation of HXs, larger heat removal efficiency of those. So, the aim of the present paper is to give preliminary study on variation of heat removal performance according to elevation of HXs. With reference to the design specification of the current reactors including APR+, we had determined conceptual design of PCCS. Using it, we developed a GOTHIC model of the APR1400 containment was adopted PCCS. This calculation model is described herein and representative results of calculation are presented. APR 1400 GOTHIC model was developed for PCCS performance calculation and sensitivity test according to installation elevation of PCCXs. Calculation results confirm that PCCS is working properly. It is found that the difference due to the installation elevation of PCCXs is insignificant at this preliminary analysis, however, further studies should be performed to confirm final performance of PCCS according to the installation elevation. These insights are important for developing the PCCS of Innovative PWR.

  7. Database management system for large container inspection system

    International Nuclear Information System (INIS)

    Gao Wenhuan; Li Zheng; Kang Kejun; Song Binshan; Liu Fang

    1998-01-01

    Large Container Inspection System (LCIS) based on radiation imaging technology is a powerful tool for the Customs to check the contents inside a large container without opening it. The author has discussed a database application system, as a part of Signal and Image System (SIS), for the LCIS. The basic requirements analysis was done first. Then the selections of computer hardware, operating system, and database management system were made according to the technology and market products circumstance. Based on the above considerations, a database application system with central management and distributed operation features has been implemented

  8. LWR aerosol containment experiments (LACE) program and initial test results

    International Nuclear Information System (INIS)

    Muhlestein, L.D.; Hilliard, R.K.; Bloom, G.R.; McCormack, J.D.; Rahn, F.J.

    1985-01-01

    The LWR aerosol containment experiments (LACE) program is described. The LACE program is being performed at the Hanford Engineer Development Laboratory (operated by Westinghouse Hanford Company) and the initial tests are sponsored by EPRI. The objectives of the LACE program are: to demonstrate, at large-scale, inherent radioactive aerosol retention behavior for postulated high consequence LWR accident situations; and to provide a data base to be used for aerosol behavior . Test results from the first phase of the LACE program are presented and discussed. Three large-scale scoping tests, simulating a containment bypass accident sequence, demonstrated the extent of agglomeration and deposition of aerosols occurring in the pipe pathway and vented auxiliary building under realistic accident conditions. Parameters varied during the scoping tests were aerosol type and steam condensation

  9. The buffer/container experiment: results, synthesis, issues

    International Nuclear Information System (INIS)

    Graham, J.; Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L.

    1997-12-01

    A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes operating

  10. The buffer/container experiment: results, synthesis, issues

    Energy Technology Data Exchange (ETDEWEB)

    Graham, J. [Univ. of Manitoba, Dept. of Civil Engineering, Winnipeg, MB (Canada); Chandler, N.A.; Dixon, D.A.; Roach, P.J.; To, T.; Wan, A.W.L

    1997-12-01

    A large in-ground experiment has examined how heat affects the performance of the dense sand bentonite 'buffer' that has been proposed for use in the Canadian Nuclear Fuel Waste Management Program. The experiment was performed by Atomic Energy of Canada Limited at its Underground Research Laboratory, Lac du Bonnet, Manitoba between 1991 and 1994. The experiment placed a full-size heater representing a container of nuclear fuel waste in a 1.24-m diameter borehole filled with buffer below the floor of a room excavated at 240-m depth in granitic rock of the Canadian Shield. The buffer and surrounding rock were extensively instrumented for temperatures, total pressures, water pressures, suctions, and rock displacements. Power was provided to the heater for almost 900 days. The experiment showed that good rock conditions can be pre-selected, a borehole can be drilled, and buffer can be placed at controlled densities and water contents. The instrumentation generally worked well, and an extensive data base was successfully organized. Drying was observed in buffer close to the heater. This caused some desiccation cracking. However the cracks only extended approximately one third of the distance to the buffer-rock interface and did not form an advective pathway. Following sampling at the time of decommissioning, cracked samples of buffer were transported to the laboratory and given access to water. The hydraulic conductivities and swelling pressures of these resaturated samples were very similar to those of uncracked buffer. A good balance was achieved between the mass of water flowing into the experiment from the surrounding rock and the increased mass of water in the buffer. A good understanding was developed of the relationships between suctions, water contents, and total pressures in buffer near the buffer-rock interface. Comparisons between measurements and predictions of measured parameters show that a good understanding has been developed of the processes

  11. Fire protection countermeasures for containment ventilation systems

    International Nuclear Information System (INIS)

    Alvares, N.; Beason, D.; Bergman, V.; Creighton, J.; Ford, H.; Lipska, A.

    1980-01-01

    The goal of this project is to find countermeasures to protect High Efficiency Particulate Air (HEPA) filters, in exit ventilation ducts, from the heat and smoke generated by fire. Initially, methods were developed to cool fire-heated air by fine water spray upstream of the filters. It was recognized that smoke aerosol exposure to HEPA filters could also cause disruption of the containment system. Through testing and analysis, several methods to partially mitigate the smoke exposure to the HEPA filters were identified. A continuous, movable, high-efficiency prefilter using modified commercial equipment was designed. The technique is capable of protecting HEPA filters over the total time duration of the test fires. The reason for success involved the modification of the prefiltration media. Commercially available filter media has particle sorption efficiency that is inversely proportional to media strength. To achieve properties of both efficiency and strength, rolling filter media were laminated with the desired properties. The approach was Edisonian, but truncation in short order to a combination of prefilters was effective. The application of this technique was qualified, since it is of use only to protect HEPA filters from fire-generated smoke aerosols. It is not believed that this technique is cost effective in the total spectrum of containment systems, especially if standard fire protection systems are available in the space. But in areas of high-fire risk, where the potential fuel load is large and ignition sources are plentiful, the complication of a rolling prefilter in exit ventilation ducts to protect HEPA filters from smoke aerosols is definitely justified

  12. A review on leakage rate tests for containment isolation systems

    International Nuclear Information System (INIS)

    Kim, In Goo; Kim, Hho Jung

    1992-01-01

    Wide experiences in operating containment isolation systems have been accumulated in Korea since 1978. Hence, it becomes necessary to review the operating data in order to confirm the integrity of containments with about 50 reactor-years of experience and to establish the future direction to the containment test program. The objectives of present work are to collect, consolidate and assess the leakage rate data, and then to find out dominant leakage paths and factors affecting integrated leakage rate test. General trends of overall leakage show that more careful surveillance during pre-operational test can reduce the containment leakage. Dominant leakage paths are found to be through air locks and large-sized valves, such as butterfly valves of purge lines, so that weighted surveillance and inspection on these dominant leakage paths can considerably reduce the containment leakage. The atmosphere stabilization are found to be the most important to obtain the reliable result. In order to get well stabilized atmosphere, temperature and flow rate of compressed air should be kept constant and it is preferable not to operate fan cooler during pressurizing the containment for test

  13. Buried waste containment system materials. Final Report

    International Nuclear Information System (INIS)

    Weidner, J.R.; Shaw, P.G.

    1997-10-01

    This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

  14. Containment and surveillance systems for international safeguards

    International Nuclear Information System (INIS)

    Ney, J.F.

    1978-01-01

    Important criteria in measuring the effectiveness of IAEA safeguards include timeliness of detection of diversion, timeliness of reporting such detections, and confidence in determining the amount of material diverted. Optimum use of IAEA inspectors, combined with adequate instrumentation, can provide a practical means for achieving these criteria. System studies are being carried out for different types of facilities that may come under IAEA safeguards to determine the proper balance between inspector's efforts and the use of safeguards instrumentation. A description of a typical study is presented. Based on the results of these studies, the program undertaken to develop those containment and surveillance subsystems for which the technical feasibility and operational acceptability need to be established is described

  15. Positron-containing systems and positron diagnostics

    International Nuclear Information System (INIS)

    1978-01-01

    The results of the experimental and theoretical investigations are presented. Considered are quantum-mechanical calculations of wave functions describing the states of positron-containing atomic systems and of cross-sections of the processes characterizing different interactions, and also the calculations of the behaviour of positrons in gases in the presence of an electric field. The results of experimental tests are presented by the data describing the behaviour of positrons and positronium in liquids, polymers and elastomers, complex oxides and in different solids. New equipment and systems developed on the basis of current studies are described. Examined is a possibility of applying the methods of model and effective potentials for studying the bound states of positron systems and for calculating cross-sections of elementary processes of elastic and inelastic collisions with a positron involved. The experimental works described indicate new possibilities of the positron diagnosis method: investigation of thin layers and films of semiconductor materials, defining the nature of chemical bonds in semiconductors, determination of the dislocation density in deformed semiconductors, derivation of important quantitative information of the energy states of radiation defects in them

  16. Tracer verification and monitoring of containment systems

    International Nuclear Information System (INIS)

    Lowry, W.; Dunn, S.D.; Williams, C.

    1996-01-01

    In-situ barrier emplacement techniques and materials for the containment of high-risk contaminants in soils are currently being developed by the Department of Energy (DOE). Because of their relatively high cost, the barriers are intended to be used in cases where the risk is too great to remove the contaminants, the contaminants are too difficult to remove with current technologies, or the potential for movement of the contaminants to the water table is so high that immediate action needs to be taken to reduce health risks. Consequently, barriers are primarily intended for use in high-risk sites where few viable alternatives exist to stop the movement of contaminants in the near term. Assessing the integrity of the barrier once it is emplaced, and during its anticipated life, is a very difficult but necessary requirement. Existing surface-based and borehole geophysical techniques do not provide the degree of resolution required to assure the formation of an integral in-situ barrier. Science and Engineering Associates, Inc., (SEA) and Sandia National Laboratories (SNL) are developing a quantitative subsurface barrier assessment system using gaseous tracers. Called SEAtrace trademark, this system integrates an autonomous, multipoint soil vapor sampling and analysis system with a global optimization modeling methodology to pinpoint leak sources and sizes in real time. SEAtrace trademark is applicable to impermeable barrier emplacements above the water table, providing a conservative assessment of barrier integrity after emplacement, as well as a long term integrity monitoring function. The SEAtrace trademark system is being developed under funding by the DOE-EM Subsurface Contaminant Focus Area

  17. Results from the DCH-1 [Direct Containment Heating] experiment

    International Nuclear Information System (INIS)

    Tarbell, W.W.; Brockmann, J.E.; Pilch, M.; Ross, J.E.; Oliver, M.S.; Lucero, D.A.; Kerley, T.E.; Arellano, F.E.; Gomez, R.D.

    1987-05-01

    The DCH-1 (Direct Containment Heating) test was the first experiment performed in the Surtsey Direct Heating Test Facility. The test involved 20 kg of molten core debris simulant ejected into a 1:10 scale model of the Zion reactor cavity. The melt was produced by a metallothermic reaction of iron oxide and aluminum powders to yield molten iron and alumina. The cavity model was placed so that the emerging debris propagated directly upwards along the vertical centerline of the chamber. Results from the experiment showed that the molten material was ejected from the caviity as a cloud of particles and aerosol. The dispersed debris caused a rapid pressurization of the 103-m 3 chamber atmosphere. Peak pressure from the six transducers ranged from 0.09 to 0.13 MPa (13.4 to 19.4 psig) above the initial value in the chamber. Posttest debris collection yielded 11.6 kg of material outside the cavity, of which approximately 1.6 kg was attributed to the uptake of oxygen by the iron particles. Mechanical sieving of the recovered debris showed a lognormal size distribution with a mass mean size of 0.55 mm. Aerosol measurements indicated a subsantial portion (2 to 16%) of the ejected mass was in the size range less than 10 m aerodynamic equivalent diameter

  18. Design of the containment spray system

    International Nuclear Information System (INIS)

    1985-12-01

    RFS or Regles Fondamentales de Surete (Basic Safety Rules) applicable to certain types of nuclear facilities lay down requirements with which compliance, for the type of facilities and within the scope of application covered by the RFS, is considered to be equivalent to compliance with technical French regulatory practice. The object of the RFS is to take advantage of standardization in the field of safety, while allowing for technical progress in that field. They are designed to enable the operating utility and contractors to know the rules pertaining to various subjects which are considered to be acceptable by the Service Central de Surete des Installations Nucleaires, or the SCSIN (Central Department for the Safety of Nuclear Facilities). These RFS should make safety analysis easier and lead to better understanding between experts and individuals concerned with the problems of nuclear safety. The SCSIN reserves the right to modify, when considered necessary, any RFS and specify, if need be, the terms under which a modification is deemed retroactive. The present RFS defines the functional requirements of the containment spray system and proposes certain complementary criteria or methods to be used in its equipment design

  19. Fire protection countermeasures for containment ventilation systems

    International Nuclear Information System (INIS)

    Alvares, N.J.; Beason, D.G.; Bergman, W.; Ford, H.W.; Lipska, A.E.

    1980-01-01

    The goal of this project is to find countermeasures to protect HEPA filters in exit ventilation ducts from the heat and smoke generated by fire. Several methods for partially mitigating the smoke exposure to the HEPA filters were identified through testing and analysis. These independently involve controlling the fuel, controlling the fire, and intercepting the smoke aerosol prior to its sorption on the HEPA filter. Exit duct treatment of aerosols is not unusual in industrial applications and involves the use of scrubbers, prefilters, and inertial impaction, depending on the size, distribution, and concentration of the subject aerosol. However, when these unmodified techniques were applied to smoke aerosols from fires on materials, common to experimental laboratories of LLNL, it was found they offered minimal protection to the HEPA filters. Ultimately, a continuous, movable, high-efficiency prefilter using modified commercial equipment was designed. This technique is capable of protecting HEPA filters over the total duration of the test fires. The reason for success involved the modificaton of the prefiltration media. Commercially available filter media has a particle sorption efficiency that is inversely proportional to media strength. To achieve properties of both efficiency and strength, we laminated rolling filter media with the desired properties. It is not true that the use of rolling prefilters solely to protect HEPA filters from fire-generated smoke aerosols is cost effective in every type of containment system, especially if standard fire-protection systems are available in the space. But in areas of high fire risk, where the potential fuel load is large and ignition sources are plentiful, the complication of a rolling prefilter in exit ventilation ducts to protect HEPA filters from smoke aerosols is definitely justified

  20. Containment integrity and leak testing. Procedures applied and experiences gained in European countries

    International Nuclear Information System (INIS)

    1987-01-01

    Containment systems are the ultimate safety barrier for preventing the escape of gaseous, liquid and solid radioactive materials produced in normal operation, not retained in process systems, and for keeping back radioactive materials released by system malfunction or equipment failure. A primary element of the containment shell is therefore its leak-tight design. The report describes the present containment concepts mostly used in European countries. The leak-testing procedures applied and the experiences gained in their application are also discussed. The report refers more particularly to pre-operational testing, periodic testing and extrapolation methods of leak rates measured at test conditions to expected leak rates at calculated accident conditions. The actual problems in periodic containment leak rate testing are critically reviewed. In the appendix to the report a summary is given of the regulations and specifications applied in different member countries

  1. Mathematical structure of ocean container transport systems; Kaiyo container yuso system no suriteki kozo ni tsuite

    Energy Technology Data Exchange (ETDEWEB)

    Shinkai, A [Kyushu University, Fukuoka (Japan). Faculty of Engineering; Chikushi, Y [Nippon Telegraph and Telephone Corp., Tokyo (Japan)

    1997-10-01

    Mathematical structure of a vessel arrangement program was discussed in order to learn roles of container ships in ocean transport systems among China, NIES/ASEAN countries and Japan. Formulation is possible on a mathematical handling method for sailing route connection diagrams between ports, a transport network to indicate container movements, a service network to indicate sailing routes, and a network generalizing them. This paper describes an analysis made on the container transport system between Japan and China, taken as an example. Four ports were selected each from Japan and China, and the statistical database for fiscals 1996 and 1994 was utilized to set models for: (a) the liner network system with transshipment at the port of Shanghai and (b) the cruising route system going through the ports of Yokohama, Nagoya and Kobe. A hypothesis was set that a consortium (coordinated ship allocation) can be implemented ideally and completely. The transport network (a) is lower by 10% in total cost than the transport network (b), resulting in 1.6 times greater productivity. Actual service network is closer to the network (b), but the system can be utilized for discussing guidelines on vessel arrangement programs with which shipping companies pursue better management efficiency under a condition that the consortium can be formed. 10 refs., 6 figs., 2 tabs.

  2. Chemistry experiences from a containment fire at Ringhals unit 2

    International Nuclear Information System (INIS)

    Arvidsson, Bengt; Svanberg, Pernilla; Bengtsson, Bernt

    2012-09-01

    containment, together with 1000 smear test (cotton pads) for chloride analysis in the chemistry laboratory to evaluate contamination levels and verify the cleaning procedures and results. The main chemistry issues and concerns have been related to surface and water contamination of chloride, bromide, carbon, lead, copper and zinc from corrosion point of view. Lack of specification and guidelines for several of this parameters forced Ringhals to establish some internal guidelines and technical basis for clean up and restart of the plant. The solubility of soot particles was found to be very low and more adhesive to surfaces at high temperature, this caused some concerns and actions to clean up reactor coolant from soot particles before fuel reload and heating. An extensive review of stainless steel Outer Diameter Stress Corrosion Cracking (ODSCC) was performed independently from the fire incident during the outage, indicating a high number of crack indications of 1-3 mm depth, all within acceptance criteria for material thickness and operation. The indications are more likely to be addressed to almost 40 years of operation in marine atmosphere then the fire itself, even if the chloride contamination from fire may have supported some propagation. All found cracks were grinded according to authority requirements and no pipes needed to be replaced. The heating and start-up of Ringhals 2 could be done successfully without any water chemistry deviations due to the fire and the following cycle have been normal. The cleanness of R2 containment surfaces are now highly improved compared to earlier outages or other sea-cooled power plants. However, an extended program has been introduced to follow external surface chloride contamination built up in containment more frequently, together with inspections of ODSCC. The workload from the containment fire has been extreme and the chemistry and corrosion experiences several. This paper gives a summary of the results, challenges, solutions and

  3. State system experience with safeguarding power reactors

    International Nuclear Information System (INIS)

    Roehnsch, W.

    1982-01-01

    This session describes the development and operation of the State System of Accountancy and Control in the German Democratic Republic, and summarizes operating experience with safeguards at power reactor facilities. Overall organization and responsibilities, containment and surveillance measures, materials accounting, and inspection procedures will be outlined. Cooperation between the IAEA, State system, facility, and supplier authorities will also be addressed

  4. Containment atmosphere cooling system for experimental fast reactor 'JOYO'

    International Nuclear Information System (INIS)

    Sasaki, Mikio; Hoshi, Akio; Sato, Morihiko; Takeuchi, Kaoru

    1979-01-01

    The experimental fast reactor ''JOYO'', the first sodium-cooled fast reactor in Japan, achieved the initially licensed full power operation (50 MW) in July 1978 and is now under steady operation. Toshiba has participated in the construction of this reactor as a leading manufacturer and supplied various systems. This article outlines the design philosophy, system concepts and the operating experience of the containment atmosphere cooling system which has many design interfaces throughout the whole plant and requires especially high reliability. The successful performance of this system during the reactor full-power operation owes to the spot cooling design philosophy and to the preoperational adjustment of heat load during the preheating period of reactor cooling system peculiar to FBR. (author)

  5. DABASCO Experiment Data Acquisition and Control System

    International Nuclear Information System (INIS)

    Alberdi Primicia, J.; Artigao Arteaga, A.; Barcala Rieveira, J. M.; Oller Gonzalez, J. C.

    2000-01-01

    DABASCO experiment wants to study the thermohydraulic phenomena produced into the containment area for a severe accident in a nuclear power facility. This document describes the characteristics of the data acquisition and control system used in the experiment. The main elements of the system were a data acquisition board, PCI-MIO-16E-4, and an application written with LaB View. (Author) 5 refs

  6. Management system for the SND experiments

    International Nuclear Information System (INIS)

    Pugachev, K.; Korol, A.

    2017-01-01

    A new management system for the SND detector experiments (at VEPP-2000 collider in Novosibirsk) is developed. We describe here the interaction between a user and the SND databases. These databases contain experiment configuration, conditions and metadata. The new system is designed in client-server architecture. It has several logical layers corresponding to the users roles. A new template engine is created. A web application is implemented using Node.js framework. At the time the application provides: showing and editing configuration; showing experiment metadata and experiment conditions data index; showing SND log (prototype).

  7. Reactor containment purge and vent valve performance experiments

    International Nuclear Information System (INIS)

    Hunter, J.A.; Steele, R.; Watkins, J.C.

    1985-01-01

    Three nuclear-designed butterfly valves typical of those used in domestic nuclear power plant containment purge and vent applications were tested. For a comparison of responses, two eight-inch nominal pipe size valves with differing internal design were tested. For extrapolation insights, a 24-inch nominal pipe size valve was also tested. The valve experiments were performed with various piping configurations and valve disc orientations to the flow, to simulate various installation options in field application. As a standard for comparing the effects of the installation options, testing was also performed in a standard ANSI test section. Test cycles were performed at inlet pressures of 5 to 60 psig, while monitoring numerous test parameters, such as the valve disc position, valve shaft torque, mass flow rate, and the pressure and temperature at multiple locations throughout the test section. An experimental data base was developed to assist in the evaluation of the current analytical methods and to determine the influence of inlet pressure, inlet duct geometry, and valve orientation to the flow media on valve torque requirements, along with any resulting limitations to the extrapolation methods. 2 refs., 15 figs

  8. FCA containment and surveillance (C/S) system

    International Nuclear Information System (INIS)

    Ogawa, Hironobu; Mukaiyama, Takehiko; Yokota, Yasuhiro.

    1994-11-01

    The Fast Critical Assembly (FCA) facility of the Japan Atomic Energy Research Inst. (JAERI) is internationally recognized as one of the most sensitive facility in the world from the viewpoint of international safeguards, because the facility possesses a large amount of metallic uranium and metallic plutonium, which are needed to perform various physical experiments. These material are subject to frequent verifications by the inspectorate, the International Atomic Energy Agency (IAEA) and the domestic authority (Science and Technology Agency of Japan, STA). Those verifications require inspectors to access to these materials for measurements and applications of seals. Human resources increase of irradiations and restrictions on the freedom of physical experiments, that are inevitably associated with these inspection activities, have been a serious problem that causes significant burdens for all relating parties. To decrease these burdens without any confliction with the inspection goals, an advanced comprehensive system of containment and surveillance has been developed. The FCA Containment and Surveillance (C/S) System consists of tow independent subsystems, i. e. Portal Monitor (P/M) and Penetration Monitor(PN/M). In this system the internal wall of the reactor building is used as a part of containment for the safeguards purpose, which enables the portal, that is installed at the internal wall of the reactor building, to be used as an area for monitoring of any removal of nuclear material. A metal detector of high sensitivity has been selected for the system since all nuclear materials possessed by the FCA has metallic forms. The internal wall has several penetrations for utility purposes, which should also be monitored for the purpose of detecting any removal of nuclear material from the reactor core area. A penetration monitor system has been developed for this purpose. This report describes functions of the system and their operation procedures. (author)

  9. FFTF control system experience

    International Nuclear Information System (INIS)

    Warrick, R.P.

    1981-01-01

    The FFTF control systems provide control equipment for safe and efficient operation of the plant. For convenience, these systems will be divided into three parts for discussions: (1) Plant Protection System (PPS); (2) Plant Control System (PCS); and (3) General Observations. Performance of each of these systems is discussed

  10. Model Based Control of Reefer Container Systems

    DEFF Research Database (Denmark)

    Sørensen, Kresten Kjær

    This thesis is concerned with the development of model based control for the Star Cool refrigerated container (reefer) with the objective of reducing energy consumption. This project has been carried out under the Danish Industrial PhD programme and has been financed by Lodam together with the Da......This thesis is concerned with the development of model based control for the Star Cool refrigerated container (reefer) with the objective of reducing energy consumption. This project has been carried out under the Danish Industrial PhD programme and has been financed by Lodam together...

  11. Experience representation in information systems

    OpenAIRE

    Kaczmarek, Jan

    2014-01-01

    This thesis looks into the ways subjective dimension of experience could be represented in artificial, non-biological systems, in particular information systems. The pivotal assumption is that experience as opposed to mainstream thinking in information science is not equal to knowledge, so that experience is a broader term which encapsulates both knowledge and subjective, affective component of experience, which so far has not been properly embraced by knowledge representation theories. This ...

  12. Experience representation in information systems

    OpenAIRE

    Kaczmarek, Jan

    2014-01-01

    This thesis looks into the ways subjective dimension of experience could be represented in artificial, non-biological systems, in particular information systems. The pivotal assumption is that experience as opposed to mainstream thinking in information science is not equal to knowledge, so that experience is a broader term which encapsulates both knowledge and subjective, affective component of experience, which so far has not been properly embraced by knowledge representation theories. Th...

  13. Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

    International Nuclear Information System (INIS)

    Binder, J.L.; McUmber, L.M.; Spencer, B.W.

    1993-01-01

    Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

  14. Application of a CFD based containment model to different large-scale hydrogen distribution experiments

    International Nuclear Information System (INIS)

    Visser, D.C.; Siccama, N.B.; Jayaraju, S.T.; Komen, E.M.J.

    2014-01-01

    Highlights: • A CFD based model developed in ANSYS-FLUENT for simulating the distribution of hydrogen in the containment of a nuclear power plant during a severe accident is validated against four large-scale experiments. • The successive formation and mixing of a stratified gas-layer in experiments performed in the THAI and PANDA facilities are predicted well by the CFD model. • The pressure evolution and related condensation rate during different mixed convection flow conditions in the TOSQAN facility are predicted well by the CFD model. • The results give confidence in the general applicability of the CFD model and model settings. - Abstract: In the event of core degradation during a severe accident in water-cooled nuclear power plants (NPPs), large amounts of hydrogen are generated that may be released into the reactor containment. As the hydrogen mixes with the air in the containment, it can form a flammable mixture. Upon ignition it can damage relevant safety systems and put the integrity of the containment at risk. Despite the installation of mitigation measures, it has been recognized that the temporary existence of combustible or explosive gas clouds cannot be fully excluded during certain postulated accident scenarios. The distribution of hydrogen in the containment and mitigation of the risk are, therefore, important safety issues for NPPs. Complementary to lumped parameter code modelling, Computational Fluid Dynamics (CFD) modelling is needed for the detailed assessment of the hydrogen risk in the containment and for the optimal design of hydrogen mitigation systems in order to reduce this risk as far as possible. The CFD model applied by NRG makes use of the well-developed basic features of the commercial CFD package ANSYS-FLUENT. This general purpose CFD package is complemented with specific user-defined sub-models required to capture the relevant thermal-hydraulic phenomena in the containment during a severe accident as well as the effect of

  15. Application of a CFD based containment model to different large-scale hydrogen distribution experiments

    Energy Technology Data Exchange (ETDEWEB)

    Visser, D.C., E-mail: visser@nrg.eu; Siccama, N.B.; Jayaraju, S.T.; Komen, E.M.J.

    2014-10-15

    Highlights: • A CFD based model developed in ANSYS-FLUENT for simulating the distribution of hydrogen in the containment of a nuclear power plant during a severe accident is validated against four large-scale experiments. • The successive formation and mixing of a stratified gas-layer in experiments performed in the THAI and PANDA facilities are predicted well by the CFD model. • The pressure evolution and related condensation rate during different mixed convection flow conditions in the TOSQAN facility are predicted well by the CFD model. • The results give confidence in the general applicability of the CFD model and model settings. - Abstract: In the event of core degradation during a severe accident in water-cooled nuclear power plants (NPPs), large amounts of hydrogen are generated that may be released into the reactor containment. As the hydrogen mixes with the air in the containment, it can form a flammable mixture. Upon ignition it can damage relevant safety systems and put the integrity of the containment at risk. Despite the installation of mitigation measures, it has been recognized that the temporary existence of combustible or explosive gas clouds cannot be fully excluded during certain postulated accident scenarios. The distribution of hydrogen in the containment and mitigation of the risk are, therefore, important safety issues for NPPs. Complementary to lumped parameter code modelling, Computational Fluid Dynamics (CFD) modelling is needed for the detailed assessment of the hydrogen risk in the containment and for the optimal design of hydrogen mitigation systems in order to reduce this risk as far as possible. The CFD model applied by NRG makes use of the well-developed basic features of the commercial CFD package ANSYS-FLUENT. This general purpose CFD package is complemented with specific user-defined sub-models required to capture the relevant thermal-hydraulic phenomena in the containment during a severe accident as well as the effect of

  16. Experimental lithium system experience

    International Nuclear Information System (INIS)

    Atwood, J.M.; Berg, J.D.; Kolowith, R.; Miller, W.C.

    1984-01-01

    The Experimental Lithium System is a test loop built to support design and operation of the Fusion Materials Irradiation Test Facility. ELS has achieved over 15,000 hours of safe and reliable operation. An extensive test program has demonstrated satisfactory performance of the system components, including an electromagnetic pump, lithium jet target, and vacuum system. Data on materials corrosion and behavior of lithium impurities are also presented. (author)

  17. Scaling for Mixed Convection Heat Transfer in Passive Containments and Experiment Design

    International Nuclear Information System (INIS)

    Wang, Shengfei; Yu, Yu; Lv, Xuefeng; Niu, Fenglei; Yan, Xiuping

    2012-01-01

    Most of the advanced nuclear reactor design utilizes passive systems to remove heat from the core by natural circulation. The passive systems will be widely used in generation III pressurized water reactor. One of the typical passive systems is passive containment cooling system (PCCS), which is a passive condenser system designed to remove heat from the containment for long term cooling after a postulated reactor accident. In order to establish empirical correlations and develop simulation models, a scaling analysis is performed in designing an experiment for the prototype PCCS. This paper presents a scaling method and the design of the experimental facility. The key dimensionless parameters governing the dominant processes are given at last

  18. Containment systems for uranium-mill tailings

    International Nuclear Information System (INIS)

    Hartley, J.N.; Buelt, J.L.

    1982-11-01

    Cover and liner systems for uranium mill tailings in the United States must satisfy stringent requirements regarding long-term stability, radon control, and radionuclide and hazardous chemical migration. The cover and liner technology discussed in this paper involves: (1) single and multilayer earthen cover systems; (2) asphalt emulsion radon barrier systems; and (3) asphalt, clay, and synthetic liner systems. These systems have been field tested at the Grand Junction, Colorado, tailings pile, where they have been shown to effectively reduce radon releases and radionuclide and chemical migration

  19. Transport in porous media containing residual hydrocarbon. 2: Experiments

    International Nuclear Information System (INIS)

    Hatfield, K.; Ziegler, J.; Burris, D.R.

    1993-01-01

    When liquid hydrocarbons or nonaqueous-phase liquids (NAPLs) become entrapped below the water table, flowing ground waters carry soluble NAPL components away from the spill zone. Transport of these dissolved NAPL components is controlled by several processes including advection, dispersion, sorption to aquifer materials, and liquid-liquid partitioning. To better understand these processes, miscible displacement experiments were conducted to generate break-through curves (BTCs) of pentafluorobenzoic acid (PFBA), benzene, and toluene on sand column with and without a fixed decane residual. A departure from equilibrium transport is observed in BTCs from the sand-decane system. These BTCs show characteristics of early breakthrough, asymmetry, and tailing. The cause of nonequilibrium is hypothesized to be rate-limited solute exchange between decane and water. A new transport model, capable of handling time-dependent exchange processes, is successfully applied to reproduce experimental BATCs. Results indicate that time-dependent partitioning becomes increasingly important as the solute decane-water partition coefficient and the aqueous-phase fluid velocity increase

  20. The application of PLC in 60Co container inspection system

    International Nuclear Information System (INIS)

    Huang Yibin; Xiang Xincheng

    2001-01-01

    The author discusses the interlock technique of 60 Co container inspection system, and introduces the hardware structure and program of interlock control system using PLC. Due to adopting PLC distributed control, the system works stably and reliably. The successful application of PLC in 60 Co container inspection system has some use for reference in nuclear technology field

  1. 46 CFR 56.50-60 - Systems containing oil.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 2 2010-10-01 2010-10-01 false Systems containing oil. 56.50-60 Section 56.50-60... APPURTENANCES Design Requirements Pertaining to Specific Systems § 56.50-60 Systems containing oil. (a)(1) Oil-piping systems for the transfer or discharge of cargo or fuel oil must be separate from other piping...

  2. Container corporation: The financial, management, and operating implications of a universal beverage container recovery system: Executive summary

    Energy Technology Data Exchange (ETDEWEB)

    1991-01-01

    This report contains an overview of the system. It discusses containers and container materials, elements of the system, and the container recovery cost structure. It includes a financial evaluation and addresses issues of implementation.

  3. Leakage detection system in nuclear reactor container

    International Nuclear Information System (INIS)

    Kurosawa, Masahiko.

    1993-01-01

    The present invention comprises an injection means for adding radioactive materials to coolants in a container cooler, a gamma ray amplitude analyzer connected to coolant pipelines and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to a drain water sump and a means for recording/transferring the data of the result of the measurement, a gamma ray amplitude analyzer connected to various kinds of pipelines and a means for recording/transferring the data of the result of the measurement, and a data processing means for comparing and analyzing the measured data of each of the gamma ray amplitude analyzers inputted from each of date recording/transferring means. The gamma ray amplitude analysis for each of the pipelines and drain water sump are conducted at an appropriate frequency, and the measured data are compared and analyzed, to improve the detection accuracy for a trace amount of leakage from each of the pressure pipelines and the container cooler coolant pipelines, thereby enabling to specify the pipeline having leakage. Maintenance efficiency is improved, and severe rupture accident in each of pressure pipelines is prevented previously. (N.H.)

  4. Tracking System : Suaineadh satellite experiment

    OpenAIRE

    Brengesjö, Carl; Selin, Martine

    2011-01-01

    The purpose of this bachelor thesis is to present a tracking system for the Suaineadh satellite experiment. The experiment is a part of the REXUS (Rocket EXperiments for University Students) program and the objective is to deploy a foldable web in space. The assignment of this thesis is to develop a tracking system to find the parts from the Suaineadh experiment that will land on Earth. It is important to find the parts and recover all the data that the experiment performed during the travel ...

  5. Development on design methodology of PWR passive containment system

    International Nuclear Information System (INIS)

    Lee, Seong Wook

    1998-02-01

    The containment is the most important barrier against the release of radioactive materials into the environment during accident conditions of nuclear power plants. Therefore the development of a reliable containment cooling system is one of key areas in advanced reactor development. To enhance the safety of the containment system, many new containment system designs have been proposed and developed in the world. Several passive containment cooling system (PCCS) concepts for both steel and concrete containment systems are overviewed and assessed comparatively. Major concepts considered are: (a) the spray of water on the outer surface of a steel containment from an elevated tank, (b) an external moat for a steel containment, (c) a suppression pool for a concrete containment, and (d) combination of the internal spray and internal or external condensers for a concrete containment. Emphasis is given to the heat removal principles, the required heat transfer area, system complexity and operational reliability. As one of conceptual design steps of containment, a methodology based on scaling principles is proposed to determine the containment size according to the power level. The AP600 containment system is selected as the reference containment to which the scaling laws are applied. Governing equations of containment pressure are set up in consideration of containment behavior in accident conditions. Then, the dimensionless numbers, which characterize the containment phenomena, are derived for the blowdown dominant and decay heat dominant stage, respectively. The important phenomena in blowdown stage are mass and energy sources and their absorption in containment atmosphere or containment structure, while heat transfer to the outer environment becomes important in decay heat stage. Based on their similarity between the prototype and the model, the containment sizes are determined for higher power levels and are compared with the SPWR containment design values available

  6. Pharmaceutical cost-containment policies and sustainability: recent Irish experience.

    Science.gov (United States)

    Kenneally, Martin; Walshe, Valerie

    2012-01-01

    Our objective is to review and assess the main pharmaceutical cost-containment policies used in Ireland in recent years, and to highlight how a policy that improved fiscal sustainability but worsened economic sustainability could have improved both if an option-based approach was implemented. The main public pharmaceutical cost-containment policy measures including reducing the ex-factory price of drugs, pharmacy dispensing fees and community drug scheme coverage, and increasing patient copayments are outlined along with the resulting savings. We quantify the cost implications of a new policy that restricts the entitlement to free prescription drugs of persons older than 70 years and propose an alternative option-based policy that reduces the total cost to both the state and the patient. This set of policy measures reduced public spending on community drugs by an estimated €380m in 2011. The policy restricting free prescription drugs for persons older than 70 years, though effective in reducing public cost, increased the total cost of the drugs supplied. The policy-induced cost increase stems from a fees anomaly between the two main community drugs schemes which is circumvented by our alternative option-based policy. Our findings highlight the need for policymakers, even when absorbed with reducing cost, to design cost-containment policies that are both fiscally and economically sustainable. Copyright © 2012 International Society for Pharmacoeconomics and Outcomes Research (ISPOR). Published by Elsevier Inc. All rights reserved.

  7. Particles growth by steam nucleation in a containment. The PITEAS experiment

    International Nuclear Information System (INIS)

    Layly, V.D.

    1993-01-01

    One of the major issues of the fission products inventory in the containment of a nuclear power plant during the few hours following the initiating phase of a severe accident (involving core degradation and fission products release from the fuel) is the physical behaviour of the aerosols suspended in the containment atmosphere. The aerosol mass concentration versus time is controlled by agglomeration, sedimentation, deposition on walls and steam nucleation phenomena. In order to assess the Nuclear Safety codes dealing with such phenomena, analytical experiments are necessary. PITEAS is an analytical experiment, on the behaviour of soluble aerosols in humid atmosphere. The PITEAS program covers three topics: diffusiophoresis, agglomeration and particle growth by steam nucleation. In the present work, we analyse the results of the tests devoted to the soluble particle growth and how such results are used to assess the IPSN code AEROSOL-B2 code, part of the System of codes ESCADRE. (author)

  8. Contained fission explosion breeder reactor system

    International Nuclear Information System (INIS)

    Juhl, N.H.; Marwick, E.F.

    1983-01-01

    A reactor system for producing useful thermal energy and valuable isotopes, such as plutonium-239, uranium-233, and/or tritium, in which a pair of sub-critical masses of fissile and fertile actinide slugs are propelled into an ellipsoidal pressure vessel. The propelled slugs intercept near the center of the chamber where the concurring slugs become a more than prompt configuration thereby producing a fission explosion. Re-useable accelerating mechanisms are provided external of the vessel for propelling the slugs at predetermined time intervals into the vessel. A working fluid of lean molten metal slurry is injected into the chamber prior to each explosion for the attenuation of the explosion's effects, for the protection of the chamber's walls, and for the absorbtion of thermal energy and debris from the explosion. The working fluid is injected into the chamber in a pattern so as not to interfere with the flight paths of the slugs and to maximize the concentration of working fluid near the chamber's center. The heated working fluid is drained from the vessel and is used to perform useful work. Most of the debris from the explosion is collected as precipitate and is used for the manufacture of new slugs

  9. Optimized Experiment Design for Marine Systems Identification

    DEFF Research Database (Denmark)

    Blanke, M.; Knudsen, Morten

    1999-01-01

    Simulation of maneuvring and design of motion controls for marine systems require non-linear mathematical models, which often have more than one-hundred parameters. Model identification is hence an extremely difficult task. This paper discusses experiment design for marine systems identification...... and proposes a sensitivity approach to solve the practical experiment design problem. The applicability of the sensitivity approach is demonstrated on a large non-linear model of surge, sway, roll and yaw of a ship. The use of the method is illustrated for a container-ship where both model and full-scale tests...

  10. FFTF in-containment cell liner design and installation experience

    International Nuclear Information System (INIS)

    Umek, A.M.; Swenson, L.D.

    1980-01-01

    Design features and liner construction techniques are discussed. Cell leak-rate tests and the methods used to locate and repair leaks are described. A brief analysis of the overall experience at FFTF is provided, with recommendations for future plant designs

  11. LCA comparison of container systems in municipal solid waste management

    International Nuclear Information System (INIS)

    Rives, Jesus; Rieradevall, Joan; Gabarrell, Xavier

    2010-01-01

    The planning and design of integrated municipal solid waste management (MSWM) systems requires accurate environmental impact evaluation of the systems and their components. This research assessed, quantified and compared the environmental impact of the first stage of the most used MSW container systems. The comparison was based on factors such as the volume of the containers, from small bins of 60-80 l to containers of 2400 l, and on the manufactured materials, steel and high-density polyethylene (HDPE). Also, some parameters such as frequency of collections, waste generation, filling percentage and waste container contents, were established to obtain comparable systems. The methodological framework of the analysis was the life cycle assessment (LCA), and the impact assessment method was based on CML 2 baseline 2000. Results indicated that, for the same volume, the collection systems that use HDPE waste containers had more of an impact than those using steel waste containers, in terms of abiotic depletion, global warming, ozone layer depletion, acidification, eutrophication, photochemical oxidation, human toxicity and terrestrial ecotoxicity. Besides, the collection systems using small HDPE bins (60 l or 80 l) had most impact while systems using big steel containers (2400 l) had less impact. Subsequent sensitivity analysis about the parameters established demonstrated that they could change the ultimate environmental impact of each waste container collection system, but that the comparative relationship between systems was similar.

  12. Containment

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

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

  13. Plutonium active operation of the Winfrith modular containment system

    International Nuclear Information System (INIS)

    Sanders, M.J.; Pengelly, M.G.A.; McSherry, K.

    1985-01-01

    Three gloveboxes contaminated with plutonium have been dismantled inside the Winfrith Modular Containment System. This system is a portable, demountable pressurised suit area with its own filters and shower entry tunnel. Details of the operation are given. (U.K.)

  14. An advanced dispatching technology for large container inspection system

    International Nuclear Information System (INIS)

    Chen Zhiqiang; Zhang Li; Kang Kejun; Gao Wenhuan

    2001-01-01

    The author describes the transmitting and dispatching technology of large container inspection system. It introduces the structure of the double buffer graded pipe lining used in the system. Strategies of queue mechanism and waiting dispatch policy are illustrated

  15. Report on container technology for the ATLAS TDAQ system

    CERN Document Server

    Gadirov, Hamid

    2016-01-01

    My summer student project "Container technology for the Upgrade of the ATLAS Trigger and Data Acquisition (TDAQ) system" focused on the research of container-based (operating system-level) virtualization for TDAQ software. Several tests were performed on Docker platform, all of them showed compatibility for TDAQ software.

  16. Throughput maximization of parcel sorter systems by scheduling inbound containers

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Schutten, Johannes M.J.; Fikse, K.; Clausen, Uwe; ten Hompel, Michael; Meier, J. Fabian

    2013-01-01

    This paper addresses the inbound container scheduling problem for automated sorter systems in express parcel sorting. The purpose is to analyze which container scheduling approaches maximize the throughput of sorter systems. We build on existing literature, particularly on the dynamic load balancing

  17. Post Fukushima requirement of containment filtered venting system in NPPS

    International Nuclear Information System (INIS)

    Deo, Anuj Kumar; Bera, S.; Nagrale, D.B.; Lakshmanan, S.P.; Baburajan, P.K.; Paul, U.K.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima safety enhancement through provision of an additional layer of Defence-in-Depth in the existing and new Indian nuclear power plants has led to the need of containment filtered venting system (CFVS). The regulatory review of the design of CFVS is in progress. In order to assess the same, the regulatory knowledge base had to be generated on the current state of the art of the design of such a system by study of the international experience on this system available in the open literature. The regulatory stand on requirements and implementation status of the CFVS in various countries were also studied. The information available on design features of various kinds of venting systems, relevant design basis and/or acceptance criteria were collected for supporting the design safety review of the Indian CFVS under consideration. During the on-going regulatory review process several analyses have been carried out, some more are in progress, to support the deliberations and decision making. This paper presents the above mentioned information and the summary of the analyses carried out including the status and outcome. Important aspects of the design review and associated analyses are also presented in this paper which includes the descriptions of the work on CFD study of venturi atomization, thermal hydraulics studies, shielding analysis and source term estimation studies carried out by the regulatory body. (author)

  18. Light and heavy water replacing system in reactor container

    International Nuclear Information System (INIS)

    Miyamoto, Keiji.

    1979-01-01

    Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

  19. Gross Containment Leakage Monitoring System (GCLM) applied to accidental impairment of containment integrity determination

    International Nuclear Information System (INIS)

    Dinu, Camelia; Talpalariu, A.; Constantinescu, G.

    2007-01-01

    The Prioritization of Generic Safety Issues (NUREG-0933 of October 2006), section 1 task II.E.4 item II.E.4.3 recommends that a method of periodic or continuous testing has to be available, in order to detect unknown gross openings in the nuclear power plants containment structure. The Palisades incident and three other incidents are exemplified, when the reactor was operated for about 1.5 years, while the containment isolation valves in a purge system bypass line were unknowingly locked in the open position. It was estimated that the presence of a GCLM system could identify an unknown breach and reduce the expected unavailability of containment due to containment integrity breach events, to a 1.6x10 -3 /year demand. (authors)

  20. Container code recognition in information auto collection system of container inspection

    International Nuclear Information System (INIS)

    Su Jianping; Chen Zhiqiang; Zhang Li; Gao Wenhuan; Kang Kejun

    2003-01-01

    Now custom needs electrical application and automatic detection. Container inspection should not only give the image of the goods, but also auto-attain container's code and weight. Its function and track control, information transfer make up the Information Auto Collection system of Container Inspection. Code Recognition is the point. The article is based on model match, the close property of character, and uses it to recognize. Base on checkout rule, design the adjustment arithmetic, form the whole recognition strategy. This strategy can achieve high recognition ratio and robust property

  1. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms [IPWF]) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. US Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as co-disposal. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister inserted in the center and/or one or more DOE SNF canisters displacing a HLW canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by

  2. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  3. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  4. Expert system for controlling plant growth in a contained environment

    Science.gov (United States)

    May, George A. (Inventor); Lanoue, Mark Allen (Inventor); Bethel, Matthew (Inventor); Ryan, Robert E. (Inventor)

    2011-01-01

    In a system for optimizing crop growth, vegetation is cultivated in a contained environment, such as a greenhouse, an underground cavern or other enclosed space. Imaging equipment is positioned within or about the contained environment, to acquire spatially distributed crop growth information, and environmental sensors are provided to acquire data regarding multiple environmental conditions that can affect crop development. Illumination within the contained environment, and the addition of essential nutrients and chemicals are in turn controlled in response to data acquired by the imaging apparatus and environmental sensors, by an "expert system" which is trained to analyze and evaluate crop conditions. The expert system controls the spatial and temporal lighting pattern within the contained area, and the timing and allocation of nutrients and chemicals to achieve optimized crop development. A user can access the "expert system" remotely, to assess activity within the growth chamber, and can override the "expert system".

  5. COMMIX analysis of AP-600 Passive Containment Cooling System

    International Nuclear Information System (INIS)

    Chang, J.F.C.; Chien, T.H.; Ding, J.; Sun, J.G.; Sha, W.T.

    1992-01-01

    COMMIX modeling and basic concepts that relate components, i.e., containment, water film cooling, and natural draft air flow systems. of the AP-600 Passive Containment Cooling System are discussed. The critical safety issues during a postulated accident have been identified as (1) maintaining the liquid film outside the steel containment vessel, (2) ensuring the natural convection in the air annulus. and (3) quantifying both heat and mass transfer accurately for the system. The lack of appropriate heat and mass transfer models in the present analysis is addressed. and additional assessment and validation of the proposed models is proposed

  6. Defense High Level Waste Disposal Container System Description Document

    International Nuclear Information System (INIS)

    2000-01-01

    The Defense High Level Waste Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded and sealed in the surface waste handling facilities, transferred to the underground through the accesses using a rail mounted transporter, and emplaced in emplacement drifts. The defense high level waste (HLW) disposal container provides long-term confinement of the commercial HLW and defense HLW (including immobilized plutonium waste forms (IPWF)) placed within disposable canisters, and withstands the loading, transfer, emplacement, and retrieval loads and environments. U.S. Department of Energy (DOE)-owned spent nuclear fuel (SNF) in disposable canisters may also be placed in a defense HLW disposal container along with commercial HLW waste forms, which is known as 'co-disposal'. The Defense High Level Waste Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container/waste package maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual canister temperatures after emplacement, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident. Defense HLW disposal containers for HLW disposal will hold up to five HLW canisters. Defense HLW disposal containers for co-disposal will hold up to five HLW canisters arranged in a ring and one DOE SNF canister in the ring. Defense HLW disposal containers also will hold two Multi-Canister Overpacks (MCOs) and two HLW canisters in one disposal container. The disposal container will include outer and inner cylinders, outer and inner cylinder lids, and may include a canister guide. An exterior label will provide a means by which to identify the disposal container and its contents. Different materials

  7. Mass extraction container closure integrity physical testing method development for parenteral container closure systems.

    Science.gov (United States)

    Yoon, Seung-Yil; Sagi, Hemi; Goldhammer, Craig; Li, Lei

    2012-01-01

    Container closure integrity (CCI) is a critical factor to ensure that product sterility is maintained over its entire shelf life. Assuring the CCI during container closure (C/C) system qualification, routine manufacturing and stability is important. FDA guidance also encourages industry to develop a CCI physical testing method in lieu of sterility testing in a stability program. A mass extraction system has been developed to check CCI for a variety of container closure systems such as vials, syringes, and cartridges. Various types of defects (e.g., glass micropipette, laser drill, wire) were created and used to demonstrate a detection limit. Leakage, detected as mass flow in this study, changes as a function of defect length and diameter. Therefore, the morphology of defects has been examined in detail with fluid theories. This study demonstrated that a mass extraction system was able to distinguish between intact samples and samples with 2 μm defects reliably when the defect was exposed to air, water, placebo, or drug product (3 mg/mL concentration) solution. Also, it has been verified that the method was robust, and capable of determining the acceptance limit using 3σ for syringes and 6σ for vials. Sterile products must maintain their sterility over their entire shelf life. Container closure systems such as those found in syringes and vials provide a seal between rubber and glass containers. This seal must be ensured to maintain product sterility. A mass extraction system has been developed to check container closure integrity for a variety of container closure systems such as vials, syringes, and cartridges. In order to demonstrate the method's capability, various types of defects (e.g., glass micropipette, laser drill, wire) were created in syringes and vials and were tested. This study demonstrated that a mass extraction system was able to distinguish between intact samples and samples with 2 μm defects reliably when the defect was exposed to air, water

  8. Requirements for containment system components in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term 'components' includes non registered items

  9. Requirements for containment system components in CANDU nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-02-01

    This Standard specifies the requirements and establishes the rules for design, fabrication, and installation of pressure-retaining containment system components. In this Standard the term `components` includes non registered items.

  10. Introduction to the controlled nuclear fusion (magnetic containment systems)

    International Nuclear Information System (INIS)

    Cabrera, J.A.; Guasp, J.; Martin, R.

    1975-01-01

    The magnetic containment systems, their more important features, and their potentiality to became thermonuclear reactors is described. The work is based upon the first part of a set of lectures dedicated to Plasma and Fusion Physics. (author)

  11. Uncanistered Spent Nuclear fuel Disposal Container System Description Document

    International Nuclear Information System (INIS)

    Pettit, N. E.

    2001-01-01

    The Uncanistered Spent Nuclear Fuel (SNF) Disposal Container System supports the confinement and isolation of waste within the Engineered Barrier System of the Monitored Geologic Repository (MGR). Disposal containers are loaded with intact uncanistered assemblies and/or individually canistered SNF assemblies and sealed in the surface waste handling facilities, transferred to the underground through the access drifts, and emplaced in emplacement drifts. The Uncanistered SNF Disposal Container provides long-term confinement of the commercial SNF placed inside, and withstands the loading, transfer, emplacement, and retrieval loads and environments. The Uncanistered SNF Disposal Container System provides containment of waste for a designated period of time, and limits radionuclide release. The disposal container maintains the waste in a designated configuration, withstands maximum handling and rockfall loads, limits the individual SNF assembly temperatures after emplacement, limits the introduction of moderator into the disposal container during the criticality control period, resists corrosion in the expected handling and repository environments, and provides containment of waste in the event of an accident

  12. DYMAC system: status and experience

    International Nuclear Information System (INIS)

    Malanify, J.J.

    1980-01-01

    The Dynamic Materials Accountability (DYMAC) system has been in operation at the Plutonium Processing Facility at the Los Alamos Scientific Laboratory since January 1978. The current status of this development and demonstration program, as well as the achieved operational experience, is reported

  13. Simplified safety and containment systems for the iris reactor

    International Nuclear Information System (INIS)

    Conway, L.E.; Lombardi, C.; Ricotti, M.; Oriani, L.

    2001-01-01

    The IRIS (International Reactor Innovative and Secure) is a 100 - 300 MW modular type pressurized water reactor supported by the U.S. DOE NERI Program. IRIS features a long-life core to provide proliferation resistance and to reduce the volume of spent fuel, as well as reduce maintenance requirements. IRIS utilizes an integral reactor vessel that contains all major primary system components. This integral reactor vessel makes it possible to reduce containment size; making the IRIS more cost competitive. IRIS is being designed to enhance reactor safety, and therefore a key aspect of the IRIS program is the development of the safety and containment systems. These systems are being designed to maximize containment integrity, prevent core uncover following postulated accidents, minimize the probability and consequences of severe accidents, and provide a significant simplification over current safety system designs. The design of the IRIS containment and safety systems has been identified and preliminary analyses have been completed. The IRIS safety concept employs some unique features that minimize the consequences of postulated design basis events. This paper will provide a description of the containment design and safety systems, and will summarize the analysis results. (author)

  14. Assessment of GOTHIC and TRACE codes against selected PANDA experiments on a Passive Containment Condenser

    Energy Technology Data Exchange (ETDEWEB)

    Papini, Davide, E-mail: davide.papini@psi.ch; Adamsson, Carl; Andreani, Michele; Prasser, Horst-Michael

    2014-10-15

    Highlights: • Code comparison on the performance of a Passive Containment Condenser. • Simulation of separate effect tests with pure steam and non-condensable gases. • Role of the secondary side and accuracy of pool boiling models are discussed. • GOTHIC and TRACE predict the experimental performance with slight underestimation. • Recirculatory flow pattern with injection of light non-condensable gas is inferred. - Abstract: Typical passive safety systems for ALWRs (Advanced Light Water Reactors) rely on the condensation of steam to remove the decay heat from the core or the containment. In the present paper the three-dimensional containment code GOTHIC and the one-dimensional system code TRACE are compared on the calculation of a variety of phenomena characterizing the response of a passive condenser submerged in a boiling pool. The investigation addresses the conditions of interest for the Passive Containment Cooling System (PCCS) proposed for the ESBWR (Economic Simplified Boiling Water Reactor). The analysis of selected separate effect tests carried out on a PCC (Passive Containment Condenser) unit in the PANDA large-scale thermal-hydraulic facility is presented to assess the code predictions. Both pure steam conditions (operating pressure of 3 bar, 6 bar and 9 bar) and the effect on the condensation heat transfer of non-condensable gases heavier than steam (air) and lighter than steam (helium) are considered. The role of the secondary side (pool side) heat transfer on the condenser performance is examined too. In general, this study shows that both the GOTHIC and TRACE codes are able to reasonably predict the heat transfer capability of the PCC as well as the influence of non-condensable gas on the system. A slight underestimation of the condenser performance is obtained with both codes. For those tests where the experimental and simulated efficiencies agree better the possibility of compensating errors among different parts of the heat transfer

  15. Analysis of Depressurization Performance in Containment of Wolsong NPP Unit 1 through Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Lee, Sunghan; Kim, Jinhyuck; Suh, Nam Duk; Cho, Songwon

    2014-01-01

    Containment filtered venting system (CFVS) is designed to open and to close isolation valves passively by an operator. CFVS is operated when the containment pressure exceeds the design pressure (225 kPa(a)) and is closed when the containment pressure decreases below 151 kPa(a). The aim of this study is to analyze the depressurization performance of Wolsong unit 1 through CFVS during SBO. The thermal-hydraulic behavior in containment of Wolsong unit 1 was evaluated using the MELCOR 1.8.6 code developed at Sandia National Laboratories (SNL) for the U.S. Nuclear Regulatory Commission (NRC). In addition, in order to evaluate the effects of the CFVS according to the venting area, a sensitivity study depending on different venting area of the CFVS was conducted. Finally, an analysis of the effects of filtering and scrubbing of radioactive material for CFVS is important but not treated in this paper. The SBO accident is chosen to analyze the thermal-hydraulic behavior of Wolsong unit 1. During SBO, the analysis of CFVS affecting on the depressurization of the containment was conducted using MELCOR 1.8.6 code. Also, a sensitivity study was carried out to evaluate the depressurization performance according to the venting area of CFVS. The results show that the containment pressure is considerably decreased and the integrity of the containment could be maintained in case of CFVS operating. Therefore, CFVS has the capacity to keep the containment pressure below the design pressure during SBO. In addition, there are large differences in the containment pressure depending on venting area. We found that the decreasing rate of the pressure in the containment and water level in CFVS depends on the venting area. In the future, a proper requirement for CFVS sizing criteria according to accident scenarios such as LBLOCA, SBLOCA and SGTR, etc. should be evaluated in order to review the licensing for CFVS. Finally, analyses of aerosols, fission product, and radioactive material

  16. Permanent monitoring of containment integrity: the sexten system

    International Nuclear Information System (INIS)

    Germain, J.L.; Janneteau, E.

    1993-01-01

    Reactor containment integrity is of prime importance to the safety of PWR units. It is checked by means of tests performed at high pressure during the containment building pressure tests. These periodical tests are supplemented in France by permanent monitoring using the SEXTEN system. First feasibility tests for this system were carried out in 1980. The encouraging results obtained led to the development of a prototype, followed by an industrial system which has since been installed in all French PWR units. This system measures the containment leak rate, with corrections for the compressed air intakes used by the air-operated valves. Leaktightness is expressed in terms of the leak rate for a 60 mbar overpressure. If the leak rate exceeds a fixed limit value, leak detection operations are initiated, using SEXTEN. A new version of the system, known as SEXTEN 2 is being developed. (authors). 2 figs

  17. Alternatives for high-level waste forms, containers, and container processing systems

    International Nuclear Information System (INIS)

    Crawford, T.W.

    1995-01-01

    This study evaluates alternatives for high-level waste forms, containers, container processing systems, and onsite interim storage. Glass waste forms considered are cullet, marbles, gems, and monolithic glass. Small and large containers configured with several combinations of overpack confinement and shield casks are evaluated for these waste forms. Onsite interim storage concepts including canister storage building, bore holes, and storage pad were configured with various glass forms and canister alternatives. All favorable options include the monolithic glass production process as the waste form. Of the favorable options the unshielded 4- and 7-canister overpack options have the greatest technical assurance associated with their design concepts due to their process packaging and storage methods. These canisters are 0.68 m and 0.54 m in diameter respectively and 4.57 m tall. Life-cycle costs are not a discriminating factor in most cases, varying typically less than 15 percent

  18. HUMOS monitoring system of leaks into the containment atmosphere

    International Nuclear Information System (INIS)

    Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

    1997-01-01

    The detection and monitoring of coolant leaks into the containment atmosphere during reactor operation is a major safety measure. Using the HUMOS monitoring system, leaks can be detected in pressure tests of integrity and in any other mode of operation when the reactor ventilation system is operating and the primary circuit and its components are pressurized. Performance tests, the design, hardware and software of the HUMOS system are briefly described. A test was performed to demonstrate that a small amount of humidity released by leakage into the containment air can be detected. (M.D.)

  19. Information managing in 60Co container inspection system

    International Nuclear Information System (INIS)

    Wu Zhifang; Gu Bohua; Zhou Liye; An Jigang; Liu Yisi

    1998-01-01

    The design, maintenance and realization of information managing database in 60 Co container inspection system made by INET of Tsinghua University is introduced. The technique of Open Database Connectivity (ODBC) is adopted to develop a general format database including text and graphic information. The database application is developed with Visual C ++ 5.0 programming language to run in 32-bit Windows operation system. It conforms to Client/Server model and supports network communication. It works very well in the laboratory emulator of 60 Co container inspection system

  20. Experiments to investigate direct containment heating phenomena with scaled models of the Surry Nuclear Power Plant

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Allen, M.D.; Pilch, M.M.

    1994-01-01

    The Containment Technology Test Facility (CTTF) and the Surtsey Test Facility at Sandia National Laboratories (SNL) are used to perform scaled experiments for the Nuclear Regulatory Commission (NRC) that simulate High Pressure Melt Ejection (HPME) accidents in a nuclear power plant (NPP). These experiments are designed to investigate the effects of direct containment heating (DCH) phenomena on the containment load. High-temperature, chemically reactive melt is ejected by high-pressure steam into a scale model of a reactor cavity. Debris is entrained by the steam blowdown into a containment model where specific phenomena, such as the effect of subcompartment structures, prototypic atmospheres, and hydrogen generation and combustion, can be studied

  1. Design of double containment canister cask storage system

    International Nuclear Information System (INIS)

    Asami, M.; Matsumoto, T.; Oohama, T.; Kuriyama, K.; Kawakami, K.

    2004-01-01

    Spent fuels discharged from Japanese LWR will be stored as recycled-fuel-resources in interim storage facilities. The concrete cask storage system is one of important forms for the spent fuel interim storage. In Japan, the interim storage facility will be located near the coast, therefore it is important to prevent SCC (Stress Corrosion Cracking) caused by sea salt particles and to assure the containment integrity of the canister which contains spent fuels. KEPCO, NFT and OCL have designed the double containment canister cask storage system that can assure the long-term containment integrity and monitor the containment performance without storage capacity decrease. Major features of the combined canister cask system are shown as follows: This system can survey containment integrity of dual canisters by monitoring the pressure of the gap between canisters. The primary canister has dual lids sealed by welding. The secondary canister has single lid tightened by bolts and sealed by metallic gaskets. The primary canister is contained in the transport cask during transportation, and the gap between the primary canister and the transport cask is filled with He gas. Under storage condition in the concrete cask, the primary canister is contained in the secondary canister, and the gap between these canisters is filled with helium gas. Hence this system can prevent the primary canister to contact sea salt particle in the air and from SCC. Decrease of cooling performance because of the double canister is compensated by fins fitted on the secondary canister surface. Then, this system can prevent the decrease of storage capacity determined by the fuel temperature limit. This system can assure that the primary canister will keep intact for long term storage. Therefore, in the case of pressure down of the gap between canisters, it can be considered that the secondary canister containment is damaged, and the primary canister will be transferred to another secondary canister at the

  2. Thermal hydraulic analysis of BWR containment venting system

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Sharma, Prashant; Paul, U.K.; Gaikwad, Avinash

    2015-01-01

    Installation of additional containment filtered venting system (CFVS) is necessary to depressurize the containment to maintain its mechanical integrity due to over pressurization during severe accident condition. A typical venting system for BWR is modelled using RELAP5 and analysed to investigate the effect of various thermal hydraulic parameters on the operational parameters of the venting system. The venting system consists of piping from the containment to the scrubber tank and exit line from the scrubber tank. The scrubber tank is partially filled with water to enable the scrubbing action to remove the particulate radionuclides from the incoming containment air. The pipe line from the containment is connected to the venturi inlet and the throat of the venturi is open to the scrubber tank water inventory at designed submergence level. The exit of the venturi is open to scrubber tank water. Filters are used in the upper air space of the scrubber tank as mist separator before venting out the air into the atmosphere through the exit vent line. The effect of thermal hydraulic parameters such as inlet fluid temperature, inlet steam content and venturi submergence in the scrubber tank on the venting flow rate, exit steam content, scrubber tank inventory, overflow line and siphon breaker flow rate is analysed. Results show that inlet steam content and the venturi nozzle submergence influence the venting system parameters. (author)

  3. Status of advanced containment systems for next generation water reactors

    International Nuclear Information System (INIS)

    1994-06-01

    The present IAEA status report is intended to provide information on the current status and development of containment systems of the next generation reactors for electricity production and, particularly, to highlight features which may be considered advanced, i.e. which present improved performance with evolutionary or innovative design solutions or new design approaches. The objectives of the present status report are: To present, on a concise and consistent basis, selected containment designs currently being developed in the world; to review and compare new approaches to the design bases for the containments, in order to identify common trends, that may eventually lead to greater worldwide consensus, to identify, list and compare existing design objectives for advanced containments, related to safety, availability, maintainability, plant life, decommissioning, economics, etc.; to describe the general approaches adopted in different advanced containments to cope with various identified challenges, both those included in the current design bases and those related to new events considered in the design; to briefly identify recent achievements and future needs for new or improved computer codes, standards, experimental research, prototype testing, etc. related to containment systems; to describe the outstanding features of some containments or specific solutions proposed by different parties and which are generally interesting to the international scientific community. 36 refs, 27 figs, 1 tab

  4. System of two containers contaminated on the inside

    International Nuclear Information System (INIS)

    Hager, L.; Heller, G.

    1983-01-01

    Two lids coupled together of a system of two containers contaminated on the inside form a frustrum of a cone with an outer surface decreasing smoothly in the same direction. The seats for these lids form two openings of the containers of the frustrum of a cone-shaped jacket matched to the jacket of the two coupled lids. The outsides of the two lids and the outsides of the containers form a surface in the same plane as the openings in the annular regions of at least one of the jacket surfaces of the frustrum of a cone or the seats of the frustrum of a cone jacket. (orig./HP) [de

  5. Emergency air cleaning system development for LMFBR containments

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.; Muhlestein, L.D.

    1975-01-01

    Criteria for evaluating the various types of Emergency Air Cleaning Systems which may be used in LMFBR plants have been established for both single containment and containment-confinement arrangements. These two plant arrangements have quite different air cleaning requirements for postulated design base accident conditions. Work is currently in progress to select from a list of candidate air cleaning systems those which best meet the criteria requirements. By means of a weighted rating system, areas of strength or weakness can be found and the conceptual system design then optimized. The final system arrangements will be ranked and several of the most promising systems selected for large-scale tests in the former CSE vessel at Hanford. 8 references. (U.S.)

  6. The directors’ roles in containing the Robben Island Diversity Experience (RIDE

    Directory of Open Access Journals (Sweden)

    Frans Cilliers

    2012-03-01

    Research purpose: The purpose of the research was to describe the experiences of the directors of RIDE in the last 10 years. Motivation for the study: Of the many and different diversity events that South African organisations present, RIDE is the only systems psycho-dynamically designed and presented event. This research was an effort to explore the nature of the directors’ roles in working with unconscious diversity dynamics in such a provocative venue. Research design, approach and method: The researchers conducted qualitative, descriptive and double hermeneutic research. The various RIDE events served as case studies. The data consisted of researcher field notes collected during the 10 years. Thematic analysis resulted in four themes, for which the researchers formulated working hypotheses. They integrated them into the research hypothesis. Main findings: Four themes emerged. They were the diversity characteristics of the directors as containers, working on the boundary between RIDE and the macro role players, attacks on the programme as container and challenges from participants. Practical/managerial implications: The research highlighted the important roles of directors’ authorisation as a resilience factor in containing RIDE. Contribution/value-add: The research contributed towards the awareness of intergroup relations between role players during diversity dynamic events and of how authorisation cements relationships.

  7. Large scale fire experiments in the HDR containment as a basis for fire code development

    International Nuclear Information System (INIS)

    Hosser, D.; Dobbernack, R.

    1993-01-01

    Between 1984 and 1991 7 different series of large scale fire experiments and related numerical and theoretical investigations have been performed in the containment of a high pressure reactor in Germany (known as HDR plant). The experimental part included: gas burner tests for checking the containment behaviour; naturally ventilated fires with wood cribs; naturally and forced ventilated oil pool fires; naturally and forced ventilated cable fires. Many results of the oil pool and cable fires can directly be applied to predict the impact of real fires at different locations in a containment on mechanical or structural components as well as on plant personnel. But the main advantage of the measurements and observations was to serve as a basis for fire code development and validation. Different types of fire codes have been used to predict in advance or evaluate afterwards the test results: zone models for single room and multiple room configurations; system codes for multiple room configurations; field models for complex single room configurations. Finally, there exist codes of varying degree of specialization which have proven their power and sufficient exactness to predict fire effects as a basis for optimum fire protection design. (author)

  8. Development of Wireless System for Containment Integrated Leakage Rate Test

    International Nuclear Information System (INIS)

    Lee, Kwang-Dae; Oh, Eung-Se; Yang, Seung-Ok

    2006-01-01

    The containment system leakage rate should be estimated periodically with reliable test equipment. In light-water reactor nuclear power plants, ANSI/ANS- 56.8 is a basis for determining leakage rates. Two types of data acquisition system, centralized type and networked type, has been used. In centralized type, all sensors are connected directly from sensors in the containment to the measuring equipment outside the building. The other hand, the networked type has several branch chains which connect one group of the network-sensors together. To test leakage rate, more than 20 temperature sensors and 6 humidity sensors, which are different for each plant, should be installed on a specific level in the containment. A wireless technology gives the benefits such as reducing installation efforts, making pretest easy, so it is widely used more and more in the plant monitoring. As the containment system has many kinds of complex barriers to the radio frequency, the radio power and frequency band for better transmission rate as well as the interference by the radio frequency should be considered. The overview of the wireless sensor system for the containment leakage rate test is described here and the test results on Yonggwang unit 4 PWR plant is presented

  9. System for cooling the containment vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Costes, Didier.

    1982-01-01

    The invention concerns a post-accidental cooling system for a nuclear reactor containment vessel. This system includes in series a turbine fed by the moist air contained in the vessel, a condenser in which the air is dried and cooled, a compressor actuated by the turbine and a cooling exchanger. The cold water flowing through the condenser and in the exchanger is taken from a tank outside the vessel and injected by a pump actuated by the turbine. The application is for nuclear reactors under pressure [fr

  10. Corrosion of copper alloys in sulphide containing district heting systems

    DEFF Research Database (Denmark)

    Thorarinsdottir, R.I.; Maahn, Ernst Emanuel

    1999-01-01

    Copper and some copper alloys are prone to corrosion in sulphide containing geothermal water analogous to corrosion observed in district heating systems containing sulphide due to sulphate reducing bacteria. In order to study the corrosion of copper alloys under practical conditions a test...... was carried out at four sites in the Reykjavik District Heating System. The geothermal water chemistry is different at each site. The corrosion rate and the amount and chemical composition of deposits on weight loss coupons of six different copper alloys are described after exposure of 12 and 18 months......, respectively. Some major differences in scaling composition and the degree of corrosion attack are observed between alloys and water types....

  11. Preserving experience through expert systems

    International Nuclear Information System (INIS)

    Jelinek, J.B.; Weidman, S.H.

    1989-01-01

    Expert systems technology, one of the branches in the field of computerized artificial intelligence, has existed for >30 yr but only recently has been made available on commercially standard hardware and software platforms. An expert system can be defined as any method of encoding knowledge by representing that knowledge as a collection of facts or objects. Decisions are made by the expert program by obtaining data about the problem or situation and correlating encoded facts (knowledge) to the data until a conclusion can be reached. Such conclusions can be relayed to the end user as expert advice. Realizing the potential of this technology, General Electric (GE) Nuclear Energy (GENE) has initiated a development program in expert systems applications; this technology offers the potential for packaging, distributing, and preserving nuclear experience in a software form. The paper discusses application fields, effective applications, and knowledge acquisition and knowledge verification

  12. APPLICATION OF A GEOGRAPHIC INFORMATION SYSTEM FOR A CONTAINMENT SYSTEM LEAK DETECTION

    Science.gov (United States)

    The use of physical and hydraulic containment systems for the isolation of contaminated ground water associated with hazardous waste sites has increased during the last decade. Existing methodologies for monitoring and evaluating leakage from hazardous waste containment systems ...

  13. Effects of aging in containment spray injection system of PWR reactor containment

    International Nuclear Information System (INIS)

    Borges, Diogo da S.; Lava, Deise D.; Affonso, Renato R.W.; Guimaraes, Antonio C.F.; Moreira, Maria de L.

    2014-01-01

    This paper presents a contribution to the study of the components aging process in commercial plants of Pressurized Water Reactors (PWR). The analysis is done by applying the method of Fault trees, Monte Carlo Method and Fussell-Vesely Importance Measurement. The study on the aging of nuclear plants, is related to economic factors involved directly with the extent of their operational life, and also provides important data on issues of safety. The most recent case involving the process of extending the life of a PWR plant can be seen in Angra 1 Nuclear Power Plant by investing $ 27 million in the installation of a new reactor cover. The corrective action generated an extension of the useful life of Angra 1 estimated in twenty years, and a great savings compared to the cost of building a new plant and the decommissioning of the first, if it had reached the operation time out 40 years. The extension of the lifetime of a nuclear power plant must be accompanied by special attention from the most sensitive components of the systems to the aging process. After the application of the methodology (aging analysis of Containment Spray Injection System (CSIS)) proposed in this paper, it can be seen that increasing the probability of failure of each component, due to the aging process, generate an increased general unavailability of the system that contains these basic components. The final results obtained were as expected and can contribute to the maintenance policy, preventing premature aging in nuclear power systems

  14. Data acquisition system issues for large experiments

    International Nuclear Information System (INIS)

    Siskind, E.J.

    2007-01-01

    This talk consists of personal observations on two classes of data acquisition ('DAQ') systems for Silicon trackers in large experiments with which the author has been concerned over the last three or more years. The first half is a classic 'lessons learned' recital based on experience with the high-level debug and configuration of the DAQ system for the GLAST LAT detector. The second half is concerned with a discussion of the promises and pitfalls of using modern (and future) generations of 'system-on-a-chip' ('SOC') or 'platform' field-programmable gate arrays ('FPGAs') in future large DAQ systems. The DAQ system pipeline for the 864k channels of Si tracker in the GLAST LAT consists of five tiers of hardware buffers which ultimately feed into the main memory of the (two-active-node) level-3 trigger processor farm. The data formats and buffer volumes of these tiers are briefly described, as well as the flow control employed between successive tiers. Lessons learned regarding data formats, buffer volumes, and flow control/data discard policy are discussed. The continued development of platform FPGAs containing large amounts of configurable logic fabric, embedded PowerPC hard processor cores, digital signal processing components, large volumes of on-chip buffer memory, and multi-gigabit serial I/O capability permits DAQ system designers to vastly increase the amount of data preprocessing that can be performed in parallel within the DAQ pipeline for detector systems in large experiments. The capabilities of some currently available FPGA families are reviewed, along with the prospects for next-generation families of announced, but not yet available, platform FPGAs. Some experience with an actual implementation is presented, and reconciliation between advertised and achievable specifications is attempted. The prospects for applying these components to space-borne Si tracker detectors are briefly discussed

  15. Numerical Analysis of a Passive Containment Filtered Venting System

    International Nuclear Information System (INIS)

    Kim, Taejoon; Ha, Huiun; Heo, Sun

    2014-01-01

    The passive Containment Filtered Venting system (CFVS) does not have principally any kind of isolation valves or filtering devices which need periodic maintenance. In this study, the hydro-thermal analysis is presented to investigate the existence of flow instability in the passive CFVS and its performance under the pressure change of APR+ containment building with LB-LOCA M/E data. The Passive Containment Filtered Venting System was suggested as a part in i-Power development project and the operation mechanism was investigated by numerical modeling and simulation using GOTHIC8.0 system code. There are four Phases for consideration to investigate the pressurization of the containment building, loss of hydrostatic head in the pipe line of CFVS, opening of pipe line and gas ejection to the coolant tank, and the head recovery inside the pipe as the containment gas exhausted. The simulation results show that gas generation rate determine the timing of head recovery in the CFVS pipe line and that the equipment of various devices inducing pressure loss at the pipe can give the capacity of Phase control of the passive CFVS operation

  16. Development of an accident management expert system for containment assessment

    International Nuclear Information System (INIS)

    Nelson, W.R.; Sebo, D.E.; Haney, L.N.

    1987-01-01

    The United States Nuclear Regulatory Commission (USNRSC) is sponsoring a program at the Idaho National Engineering Laboratory (INEL) to develop an accident management expert system. The intended users of the system are the personnel of the NRC Operations Center in Washington, D.C. The expert system will be used to help NRC personnel monitor and evaluate the status and management of the containment during a severe reactor accident. The knowledge base will include severe accident knowledge regarding the maintenance of the critical safety functions, especially containment integrity, during an accident. This paper summarizes the concepts that have been developed for the accident management expert system, and the plans that have been developed for its implementation

  17. Improving the performance of sorter systems by scheduling inbound containers

    NARCIS (Netherlands)

    Haneyah, S.W.A.; Schutten, Johannes M.J.; Fikse, K.

    2013-01-01

    This paper addresses the inbound containers scheduling problem for automated sorter systems in two different industrial sectors: parcel & postal sorting and baggage handling. We build on existing literature, particularly on the dynamic load balancing algorithm designed for the parcel hub scheduling

  18. Structure-rheology relations in sodium caseinate containing systems

    NARCIS (Netherlands)

    Ruis, H.G.M.

    2007-01-01

    The general aim of the work described in this thesis was to investigate structure-rheologyrelations for dairy related products, focusing on model systems containing sodium caseinate. The acid inducedgelationof sodium caseinate, of sodium caseinate stabilized emulsions, and the effect of shear on the

  19. Safety of systems for the retention of wastes containing radionuclides

    International Nuclear Information System (INIS)

    1980-11-01

    Information and minimal requirements demanded by CNEN for the emission of the Approval Certificate of the Safety Analysis Report related to system for the retention of wastes containing radionuclide, are established, aiming to assure low radioactivity levels to the environment. (E.G.) [pt

  20. Microparticulate drug delivery system containing tramadol hydrochloride for pain treatment.

    Science.gov (United States)

    Ciurba, Adriana; Todoran, Nicoleta; Vari, C E; Lazăr, Luminita; Al Hussein, Stela; Hancu, G

    2014-01-01

    The current trend of replacing conventional pharmaceutical forms is justified because most substances administered in this form give fluctuations of therapeutic concentrations and often outside the therapeutic range. In addition, these formulations offer a reduction in the dose or the number of administrations, thus increasing patient compliance. In the experiment, we developed an appropriate technology for the preparation of gelatin microspheres containing tramadol hydrochloride by emulsification/cross-linking method. The formulated microspheres were characterized by product yield, size distribution, encapsulation efficiency and in vitro release of tramadol hydrochloride. Data obtained from in vitro release studies were fitted to various mathematical models to elucidate the transport mechanisms. The kinetic models used were zero-order, first-order, Higuchi Korsmeyer-Peppas and Hopfenberg. Spherical microspheres were obtained, with free-flowing properties. The entrapment efficiency of tramadol hydrochloride in microparticles was 79.91% and product yield -94.92%. As the microsphere size was increased, the entrapment efficiency increased. This was 67.56, 70.03, 79.91% for formulations MT80-250, MT8-500 and, MT250-500. High entrapment efficiency was observed for MT250-500 formulation. The gelatin microspheres had particle sizes ranging from 80 to 500 microm. The drug was released for a period of 12 hours with a maximum release of 96.02%. Of the three proposed formulations, MT250-500 presented desirable properties and optimal characteristics for the therapy of pain. Release of tramadol hydrochloridi was best fitted to Korsmeyer-Peppas equation because the Akaike Information Criterion had the lowest values for this kinetic model. These results suggest the opportunity to influence the therapeutic characteristics of gelatin microspheres to obtain a suitable drug delivery system for the oral administration of tramadol hydrochloride.

  1. An experimental investigation of natural circulated air flow in the passive containment cooling system

    International Nuclear Information System (INIS)

    Ryu, S.H.; Oh, S.M.; Park, G.C.

    2004-01-01

    The objective of this study is to investigate the effects of air inlet position and external conditions on the natural circulated air flow rate in a passive containment cooling system of the advanced passive reactor. Experiments have been performed with 1/36 scaled segment type passive containment test facility. The air velocities and temperatures are measured through the air flow path. Also, the experimental results are compared with numerical calculations and show good agreement. (author)

  2. Optimal design of passive containment cooling system for innovative PWR

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon [Central Research Institute, Korea Hydro and Nuclear Power, Ltd., Daejeon (Korea, Republic of)

    2017-08-15

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  3. Optimal design of passive containment cooling system for innovative PWR

    Directory of Open Access Journals (Sweden)

    Huiun Ha

    2017-08-01

    Full Text Available Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS of an innovative pressurized water reactor (PWR. A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT geometry, PCCS heat exchanger (PCCX location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed.

  4. Optimal design of passive containment cooling system for innovative PWR

    International Nuclear Information System (INIS)

    Ha, Huiun; Lee, Sang Won; Kim, Hangon

    2017-01-01

    Using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) code, thermal-hydraulic phenomena that occur inside the containment have been investigated, along with the preliminary design of the passive containment cooling system (PCCS) of an innovative pressurized water reactor (PWR). A GOTHIC containment model was constructed with reference to the design data of the Advanced Power Reactor 1400, and report related PCCS. The effects of the design parameters were evaluated for passive containment cooling tank (PCCT) geometry, PCCS heat exchanger (PCCX) location, and surface area. The analyzed results, obtained using the single PCCT, showed that repressurization and reheating phenomena had occurred. To resolve these problems, a coupled PCCT concept was suggested and was found to continually decrease the containment pressure and temperature without repressurization and reheating. If the installation level of the PCCX is higher than that of the PCCT, it may affect the PCCS performance. Additionally, it was confirmed that various means of increasing the external surface area of the PCCX, such as fins, could help improve the energy removal performance of the PCCS. To improve the PCCS design and investigate its performance, further studies are needed

  5. Test plan for buried waste containment system materials

    International Nuclear Information System (INIS)

    Weidner, J.; Shaw, P.

    1997-03-01

    The objectives of the FY 1997 barrier material work at the Idaho National Engineering and Environmental Laboratory are to (1) select a waste barrier material and verify that it is compatible with the Buried Waste Containment System Process, and (2) determine if, and how, the Buried Waste Containment System emplacement process affects the material properties and performance (on proof of principle scale). This test plan describes a set of measurements and procedures used to validate a waste barrier material for the Buried Waste Containment System. A latex modified proprietary cement manufactured by CTS Cement Manufacturing Company will be tested. Emplacement properties required for the Buried Waste Containment System process are: slump between 8 and 10 in., set time between 15 and 30 minutes, compressive strength at set of 20 psi minimum, and set temperature less than 100 degrees C. Durability properties include resistance to degradation from carbonate, sulfate, and waste-site soil leachates. A set of baseline barrier material properties will be determined to provide a data base for comparison with the barrier materials when tested in the field. The measurements include permeability, petrographic analysis to determine separation and/or segregation of mix components, and a set of mechanical properties. The measurements will be repeated on specimens from the field test material. The data will be used to determine if the Buried Waste Containment System equipment changes the material. The emplacement properties will be determined using standard laboratory procedures and instruments. Durability of the barrier material will be evaluated by determining the effect of carbonate, sulfate, and waste-site soil leachates on the compressive strength of the barrier material. The baseline properties will be determined using standard ASTM procedures. 9 refs., 1 fig., 2 tabs

  6. Reliability assessment of passive containment isolation system using APSRA methodology

    International Nuclear Information System (INIS)

    Nayak, A.K.; Jain, Vikas; Gartia, M.R.; Srivastava, A.; Prasad, Hari; Anthony, A.; Gaikwad, A.J.; Bhatia, S.; Sinha, R.K.

    2008-01-01

    In this paper, a methodology known as APSRA (Assessment of Passive System ReliAbility) has been employed for evaluation of the reliability of passive systems. The methodology has been applied to the passive containment isolation system (PCIS) of the Indian advanced heavy water reactor (AHWR). In the APSRA methodology, the passive system reliability evaluation is based on the failure probability of the system to carryout the desired function. The methodology first determines the operational characteristics of the system and the failure conditions by assigning a predetermined failure criterion. The failure surface is predicted using a best estimate code considering deviations of the operating parameters from their nominal states, which affect the PCIS performance. APSRA proposes to compare the code predictions with the test data to generate the uncertainties on the failure parameter prediction, which is later considered in the code for accurate prediction of failure surface of the system. Once the failure surface of the system is predicted, the cause of failure is examined through root diagnosis, which occurs mainly due to failure of mechanical components. The failure probability of these components is evaluated through a classical PSA treatment using the generic data. The reliability of the PCIS is evaluated from the probability of availability of the components for the success of the passive containment isolation system

  7. Experiences made with tritium-containing water used as tracer in laboratory experiments with fluvioglacial gravels

    International Nuclear Information System (INIS)

    Klotz, D.; Rauert, W.

    1982-01-01

    Batch tests performed on 11 different Bavarian fluvioglacial gravels led to tritium distribution coefficients, which deviated not or only insignificantly from zero within the range of experimental accuracy applied to routine testings. The result of nine flow experiments in a gravelfilled column was a mean retardation factor of 1.01 +- 0.01. These experiments thus showed - as it had been expected - that 3 HHO is not significantly delayed with regard to the flow or movement of the water. (orig.) [de

  8. CONTEMPT4/MOD2: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Metcalfe, L.J.; Mings, W.J.; Hartman, J.E.; Crail, A.C.

    1978-02-01

    CONTEMPT4/MOD2 is a digital computer program, written in FORTRAN IV, which describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental 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 intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat conducting structures, sump drain, and PWR ice condensers. Dynamic storage allocations (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

  9. CONTEMPT 4/MOD 3: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Cheng, T.C.; Metcalfe, L.J.; Hartman, J.E.; Mings, W.J.; Crail, A.C.

    1982-12-01

    CONTEMPT4/MOD3 is a digital computer program, written in FORTRAN IV, that describes the behavior of multicompartment pressurized water reactor (PWR) containment systems and experimental containment systems subjected to postulated loss-of-coolant accident (LOCA) conditons. The program calculates the time variation of compartment pressures, temperatures, mass and energy inventories, heat structure temperature distributions, and intercompartment mass and energy exchange based on user-supplied values for compartment descriptions, time step and edit controls, and selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, fan coolers, heat-conducting structures, sump drains, and PWR ice condensers. Dynamic stroage allocation (DSA) is used to limit the amount of computer core used for each problem. Optional automatic time step control allows the code to determine time step sizes within limits dictated by the user. Multicompartment capability (up to 999 individual compartments) and generalized, user-oriented input-data descriptions permit improved flexibility over previous codes in the CONTEMPT series. Analytical model descriptions, input instructions, and sample problem results are presented

  10. CONTEMPT4/MOD6: a multicompartment containment system analysis program

    International Nuclear Information System (INIS)

    Lin, C.C.; Economos, C.; Lehner, J.R.; Maise, G.

    1986-03-01

    CONTEMPT4/MOD6 is a digital computer program that describes the response of multicompartment containment system subjected to postulated loss-of-coolant accident (LOCA) conditions. The program is written in FORTRAN IV and can accomodate both pressurized water reactor (PWR) and boiling water reactor (BWR) containment systems. Also, both design basis accident (DBA) and degraded core type LOCA conditions can be analyzed. The program calculates the time variation of compartment pressures, temperatures and mass and energy inventories due to intercompartment mass and energy exchange taking into account user supplied descriptions of compartments, intercompartment junction flow areas, LOCA source terms and user selected problem features. Analytical models available to describe containment systems include models for containment fans and pumps, cooling sprays, heat conducting structures, sump drains, PWR ice condensers and BWR pressure suppression systems. To accommodate degraded core type accidents, analytical models for hydrogen and carbon monoxide combustion within compartments and energy transfer due to gas radiation are also provided. Dynamic storage allocation (DSA) is used to limit the amount of computer core used for each problem. The flexibility needed to more realistically model the complexity of prototypical containments is provided by the multicompartment capability (up to 999 individual compartments) and generalized user oriented input data descriptions. The program employs an implicit algorithm to compute junction flow when numerically induced flow oscillations are encountered. This capability provides significant reduction of computer run time relative to previous codes in the CONTEMPT series. Descriptions of these analytical models are presented, together with input instructions for the CONTEMPT4/MOD6 program and sample problem results. 23 refs., 62 figs

  11. High temperature performance limit of containment system of transport cask

    International Nuclear Information System (INIS)

    Kato, Osamu; Saegusa, Toshiari

    1998-01-01

    The containment performance of a containment system using elastomer gaskets for transport casks under a high temperature and high pressure was clarified. Major results are as follows; (1) The deformation characteristics of the gaskets were represented by the compressive permanent strain rate (Dp). The temperature and time dependence was shown by Larson-Miller Parameter (LMP). (2) Generally, the high temperature performance limit is obtained by a value of LMP when the Dp value reaches 80%. However, the gaskets (FKM, VMQ, EPDM) used for real transport casks were not damaged and the containment performance was not deteriorated as a conservative condition. (3) Assuming that the service period of the gaskets for transport casks is 3 months or 1 year, the high temperature performance limit of the gasket made of fluorine rubber (FKM) is 202degC or 182degC, respectively, which includes safety margin. (author)

  12. Report on further development of the Winfrith Modular Containment System and associated equipment

    International Nuclear Information System (INIS)

    Sanders, M.J.; Pengelly, M.G.A.

    1988-03-01

    As a result of operational experience gained with the Winfrith Modular Containment, the need for a lifting aid to facilitate the decommissioning of tall plant, a 2-stage mobile ventilation system and an improved shower entry tunnel was identified. Improved plant and equipment has been designed, constructed and tested and the results are presented here. (author)

  13. Piping systems, containment pre-stressing and steel ventilation chimney

    International Nuclear Information System (INIS)

    Stuessi, U.

    1996-01-01

    Units 5 and 6 of NPP Kozloduy have been designed initially for seismic levels which are considered too low today. In the frame of an IAEA Coordinated Research Programme, a Swiss team has been commissioned by Natsionalna Elektricheska Kompania, Sofia, to analyse the relevant piping system, the containment prestressing and the steel ventilation chimney and to recommend upgrade measures for adequate seismic capacity where applicable. Seismic input had been specified by and agreed upon earlier by IAEA experts. The necessary investigations have been performed in 1995 and discussed with internationally recognized experts. The main results may be summarized as follows: Upgrades are necessary at different piping sy ports (additional snubbers or viscous dampers). These fixes can be done easily at low cost. The containment prestressing tendons are adequately designed for the specified load combinations. However, unfavourable construction features endanger the reliability. It is therefore strongly recommended to replace the tendons stepwise and to upgrade the existing monitoring system. Finally, the steel ventilation chimney may not withstand a seismic event, however the containment and diesel generator building will not be destroyed at possible impact by the chimney. On the other hand the roof of the main building has to be reinforced partially. It is recommended to continue the project for 1996 and 1997 to implement the upgrade measures mentioned above, to analyse the remaining piping systems and to consolidate all results obtained by different research groups of the IAEA programme with respect to piping systems including components and tanks

  14. DHCVIM - a direct heating containment vessel interactions module: applications to Sandia National Laboratories Surtsey experiments

    International Nuclear Information System (INIS)

    Ginsberg, T.; Tutu, N.K.

    1987-01-01

    Direct containment heating is the mechanism of severe nuclear reactor accident containment loading that results from transfer of thermal and chemical energy from high-temperature, finely divided, molten core material to the containment atmosphere. The direct heating containment vessel interactions module (DHCVIM) has been developed at Brookhaven National Laboratory to model the mechanisms of containment loading resulting from the direct heating accident sequence. The calculational procedure is being used at present to model the Sandia National Laboratories one-tenth-scale Surtsey direct containment heating experiments. The objective of the code is to provide a test bed for detailed modeling of various aspects of the thermal, chemical, and hydrodynamic interactions that are expected to occur in three regions of a containment building: reactor cavity, intermediate subcompartments, and containment dome. Major emphasis is placed on the description of reactor cavity dynamics. This paper summarizes the modeling principles that are incorporated in DHCVIM and presents a prediction of the Surtsey Test DCH-2 that was made prior to execution of the experiment

  15. Thermodynamics of organic mixtures containing amines. VIII. Systems with quinoline

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez, Juan Antonio [G.E.T.E.F., Grupo Especializado en Termodinamica de Equilibrio entre Fases, Departamento de Fisica Aplicada, Facultad de Ciencias, Universidad de Valladolid, E-47071 Valladolid (Spain)], E-mail: jagl@termo.uva.es; Domanska, Urszula; Zawadzki, Maciej [Physical Chemistry Division, Faculty of Chemistry, Warsaw University of Technology, 00-664 Warsaw (Poland)

    2008-08-15

    (Solid + liquid) equilibrium temperatures for mixtures containing quinoline and 1-dodecanol, 1-hexadecanol, or 1-octadecanol have been measured using a dynamic method. (Quinoline + benzene, +alkane, or +1-alkanol) systems were investigated using DISQUAC. The corresponding interaction parameters are reported. The model yields a good representation of molar excess Gibbs free energies, G{sup E}, molar excess enthalpies, H{sup E}, and of the (solid + liquid) equilibria, SLE. Interactional and structural effects were analysed comparing H{sup E} and the molar excess internal energy at constant volume, U{sub V}{sup E}. It was encountered that structural effects are very important in systems involving alkanes or 1-alkanols. Interactions between amine molecules are stronger in mixtures with quinoline than in those containing pyridine, which was ascribed to the higher polarizability of quinoline.

  16. 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.

  17. LWR severe accident simulation: Iodine behaviour in FPT2 experiment and advances on containment iodine chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Girault, N., E-mail: nathalie.girault@irsn.fr [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Bosland, L. [Institut de Radioprotection et de Surete Nucleaire (IRSN), BP3 - 13115 St.-Paul-lez-Durance (France); Dickinson, S. [National Nuclear Laboratory, Harwell, Oxon OX11 0QT (United Kingdom); Funke, F. [AREVA NP Gmbh, PO Box 1109, 91001 Erlangen (Germany); Guentay, S. [Paul Scherrer Institut, 5232 Villigen PSI (Switzerland); Herranz, L.E. [Centro des Investigaciones Energeticas, MedioAmbiantales y Tecnologicas, av. Complutense 2, 28040 Madrid (Spain); Powers, D. [Sandia National Laboratories, New Mexico, PO Box 5800, Albuquerque, NM 87185 (United States)

    2012-02-15

    the iodine behaviour in the FPT2 containment vessel based on dedicated small-scale analytical experiments and computer codes calculations. Other investigations dealing with primary circuit and sump chemistry are also reported. These could help to scale the results of Phebus-FP tests to reactor accidents. Modelling studies were generally successful when a gaseous iodine injection from the primary circuit was assumed. Indeed, though each of the iodine codes has specific iodine chemistry features that should be further developed and each approach to the modelling is distinct, the overall iodine behaviour in the FPT2 containment is generally well reproduced by the models that predict: Bullet a low final gaseous iodine concentration in the containment atmosphere, Bullet a predominant iodine concentration in the sump and to a lesser extent a significant iodine deposition on containment surfaces. The main code-to-code differences, in the results obtained in gaseous iodine speciation, come from the various treatments of gaseous radiolytic reactions. Calculations that include the radiolytic conversion of volatile iodine into iodine oxide particulate show there is a persistence of both gaseous iodine and iodine oxide particles in the atmosphere. There are also some variations between the predicted organic iodine concentrations that depend mainly on the initial assumptions. A key aspect of the Phebus FPT2 test interpretation is that the long term iodine behaviour in the containment can be explained by exchanges between organic iodide released from painted surfaces and inorganic iodine released from deposited aerosol on the containment walls. Further studies of regulatory significance on sump chemistry showed that the gaseous iodine control that was evident in the Phebus tests through silver release and/or alkaline buffered sump solutions may not be assured. As most of the past iodine aqueous chemistry studies were done with rather pure systems and because of the uncompleted

  18. Safety Research Experiment Facility Project. Conceptual design report. Volume IV. Reactor containment

    International Nuclear Information System (INIS)

    1975-12-01

    The principal purpose of the SAREF Reactor Containment Building (RCB) is to prevent the uncontrolled release of radioactive materials to the atmosphere as a result of accidental occurrences inside the containment. The RCB houses numerous reactor systems and components including the Prestressed Concrete Reactor Vessel (PCRV). The design of the RCB is of reinforced concrete (steel-lined). The containment building is embedded nearly 100 feet in lava rock. It has therefore been necessary to independently formulate an appropriate and conservative design approach

  19. Structure-rheology relations in sodium caseinate containing systems

    OpenAIRE

    Ruis, H.G.M.

    2007-01-01

    The general aim of the work described in this thesis was to investigate structure-rheologyrelations for dairy related products, focusing on model systems containing sodium caseinate. The acid inducedgelationof sodium caseinate, of sodium caseinate stabilized emulsions, and the effect of shear on the structure formation was characterized. Special attention was given to the sol-gel transition point, which was defined by a frequency independent loss tangent. It was shown that the sol-gel transit...

  20. Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

    International Nuclear Information System (INIS)

    Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

    1999-01-01

    A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

  1. LHCb : Full Experiment System Test

    CERN Multimedia

    Cattaneo, M

    2009-01-01

    LHCb had been planning to commission its High Level Trigger software and Data Quality monitoring procedures using real collisions data from the LHC pilot run. Following the LHC incident on 19th September 2008, it was decided to commission the system using simulated data. This “Full Experiment System Test” consists of: - Injection of simulated minimum bias events into the full HLT farm, after selection by a simulated Level 0 trigger. - Processing in the HLT farm to achieve the output rate expected for nominal LHC luminosity running, sustained over the typical duration of an LHC fill. - Real time Data Quality validation of the HLT output, validation of calibration and alignment parameters for use in the reconstruction. - Transmission of the event data, calibration data and book-keeping information to Tier1 sites and full reconstruction of the event data. - Data Quality validation of the reconstruction output. We will report on the preparations and results of FEST09, and on the status of commissioning for no...

  2. Secondary containment systems for bulk oil storage facilities

    International Nuclear Information System (INIS)

    Carr, B.A.

    1996-01-01

    The United States Environmental Protection Agency has conducted site inspections at several onshore bulk oil above ground storage facilities, to ensure that owners follow the spill prevention, control and countermeasure regulations. The four violations which were most frequently cited at these facilities were: (1) lack of a spill prevention plan, (2) lack of appropriate containment equipment to prevent discharged oil from reaching a navigable water course, (3) inadequate secondary containment structures, and (4) lack of an adequate quick drainage system in the facility tank loading/unloading area. Suggestions for feasible designs which would improve the impermeability of secondary containment for above ground storage tanks (AST) included the addition of a liner, retrofitting the bottom of an AST with a second steel plate, using a geosynthetic liner on top of the original bottom, installing a leak detection system in the interstitial space between the steel plates, or installing an under-tank liner with a leak detection system during construction of a new AST. 2 refs

  3. Disruption management for truck appointment system at a container terminal

    DEFF Research Database (Denmark)

    Li, N.; Chen, Gang; Jin, Z.

    2016-01-01

    -appointed arrivals at a container terminal that is running an appointment system. Second, we propose some response strategies to cope with different levels of disruptions, and evaluate their resilience ability with two Key Performance Indicators (KPIs): total waiting time of on-time trucks and total idling emissions...... of all trucks, in order to balance the service quality to punctual arrivals and green performance of the whole system. Third, we conduct a sensitivity analysis using a discrete event simulation to understand the performance of the proposed strategies. Considering both KPIs, the best strategy in most......-crane moving distance, especially when the first KPI is given lower weight than the second one....

  4. Simulation of International Standard Problem No. 44 'KAEVER' experiments on aerosol behaviour with the CONTAIN code

    International Nuclear Information System (INIS)

    Kljenak, I.

    2001-01-01

    Experiments on aerosol behavior in a vapor-saturated atmosphere, which were performed in the KAEVER experimental facility and proposed for the OECD International Standard Problem No. 44, were simulated with the CONTAIN thermal-hydraulic computer code. The purpose of the work was to assess the capability of the CONTAIN code to model aerosol condensation and deposition in a containment of a light-water-reactor nuclear power plant at severe accident conditions. Results of dry and wet aerosol concentrations are presented and analyzed.(author)

  5. The assessment of containment codes by experiments simulating severe accident scenarios

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

    Hitherto, a generally applicable validation matrix for codes simulating the containment behaviour under severe accident conditions did not exist. Past code applications have shown that most problems may be traced back to inaccurate thermalhydraulic parameters governing gas- or aerosol-distribution events. A provisional code-validation matrix is proposed, based on a careful selection of containment experiments performed during recent years in relevant test facilities under various operating conditions. The matrix focuses on the thermalhydraulic aspects of the containment behaviour after severe accidents as a first important step. It may be supplemented in the future by additional suitable tests

  6. Experiences in the emptying of waste silos containing solid nuclear waste from graphite- moderated reactors

    International Nuclear Information System (INIS)

    Wall, S.; Schwarz, T.

    2003-01-01

    Before reactor sites can be handed over for ultimate decommissioning, at some sites silos containing waste from operations need to be emptied. The form and physical condition of the waste demands sophisticated retrieval technologies taking into account the onsite situation in terms of infrastructure and silo geometry. Furthermore, in the case of graphite moderated reactors, this waste usually includes several tonnes of graphite waste requiring special HVAC and dust handling measures. RWE NUKEM Group has already performed several contracts dealing with such emptying tasks. Of particular interest for the upcoming decommissioning projects in France might be the activities at Vandellos, Spain and Trawsfynnyd, UK. Retrieval System for Vandellos NPP is discussed. Following an international competitive tender exercise, RWE NUKEM won the contract to provide a turn-key retrieval system. This involved the design, manufacture and installation of a system built around the modules of a 200 kg capacity version of the ARTISAN manipulator system. The ARTISAN 200 manipulator, with remote slave arm detach facility, was deployed on a telescopic mast inserted into the silos through the roof penetrations. The manipulator deployed a range of tools to gather the waste and load it into a transfer basket, deployed through an adjacent penetration. After commissioning, the system cleared the vaults in less than the scheduled period with no failures. At the Trawsfynnyd Magnox plants two types of intermediate level waste (ILW) accumulated on site; namely Miscellaneous Activated Components (MAC) and Fuel Element Debris (FED). MAC is predominantly components that have been activated by the reactor core and then discharged. FED mainly consists of fuel cladding produced when fuel elements were prepared for dispatch to the reprocessing facility. RWE NUKEM Ltd. was awarded a contract to design, supply, commission and operate equipment to retrieve, pack and immobilize the two waste streams. Major

  7. Containment hydrogen removal system for a nuclear power plant

    International Nuclear Information System (INIS)

    Callaghan, V.M.; Flynn, E.P.; Pokora, B.M.

    1984-01-01

    A hydrogen removal system (10) separates hydrogen from the containment atmosphere of a nuclear power plant using a hydrogen permeable membrane separator (30). Water vapor is removed by condenser (14) from a gas stream withdrawn from the containment atmosphere. The gas stream is then compressed by compressor (24) and cooled (28,34) to the operating temperature of the hydrogen permeable membrane separator (30). The separator (30) separates the gas stream into a first stream, rich in hydrogen permeate, and a second stream that is hydrogen depleted. The separated hydrogen is passed through a charcoal adsorber (48) to adsorb radioactive particles that have passed through the hydrogen permeable membrane (44). The hydrogen is then flared in gas burner (52) with atmospheric air and the combustion products vented to the plant vent. The hydrogen depleted stream is returned to containment through a regenerative heat exchanger (28) and expander (60). Energy is extracted from the expander (60) to drive the compressor (24) thereby reducing the energy input necessary to drive the compressor (24) and thus reducing the hydrogen removal system (10) power requirements

  8. Report on further development of the Winfrith Modular Containment System and associated equipment

    International Nuclear Information System (INIS)

    Sanders, M.J.; Pengelly, M.G.A.

    1987-12-01

    The Winfrith modular containment system was developed to enable redundant plutonium processing plant to be safely decommissioned. As a result of operational experience the need for a lifting aid to facilitate the decommissioning of tall plant, a 2-stage mobile ventilation system and an improved shower entry tunnel was identified. Improved plant and equipment has been designed, constructed and tested and the results are presented here. (author)

  9. Secondary containment system for a high tritium research cryostat

    International Nuclear Information System (INIS)

    Tsugawa, R.T.; Fearon, D.; Souers, P.C.; Hickman, R.G.; Roberts, P.E.

    1976-01-01

    A 4.2- to 300-K liquid helium cryostat has been constructed for cryogenic samples of D--T containing up to 4 x 10 14 dis/s (10,000 Ci) of tritium radioactivity. The cryostat is enclosed in a secondary box, which acts as the ultimate container in case of a tritium release. Dry argon is flushed through the box, and the box atmosphere is monitored for tritium, oxygen, and water vapor. A rupture disk and abort tank protect the box atmosphere in case the sample cell breaks. If tritium breaks into the box, a powdered uranium getter trap reduces the 4 x 10 14 dis/s (10,000 Ci) to 4 x 10 9 dis/s (0.1 Ci) in 24 h. A backup palladium--zeolite getter system goes into operation if an overabundance of oxygen contaminates the uranium getter

  10. Simulation of Molecular Transport in Systems Containing Mobile Obstacles.

    Science.gov (United States)

    Polanowski, Piotr; Sikorski, Andrzej

    2016-08-04

    In this paper, we investigate the movement of molecules in crowded environments with obstacles undergoing Brownian motion by means of extensive Monte Carlo simulations. Our investigations were performed using the dynamic lattice liquid model, which was based on the cooperative movement concept and allowed to mimic systems at high densities where the motion of all elements (obstacles as well as moving particles) were highly correlated. The crowded environments are modeled on a two-dimensional triangular lattice containing obstacles (particles whose mobility was significantly reduced) moving by a Brownian motion. The subdiffusive motion of both elements in the system was analyzed. It was shown that the percolation transition does not exist in such systems in spite of the cooperative character of the particles' motion. The reduction of the obstacle mobility leads to the longer caging of liquid particles by mobile obstacles.

  11. Degradation and failure characteristics of NPP containment protective coating systems

    Energy Technology Data Exchange (ETDEWEB)

    Sindelar, R.L.

    2000-03-30

    A research program to investigate the performance and potential for failure of Service Level 1 coating systems used in nuclear power plant containment is in progress. The research activities are aligned to address phenomena important to cause failure as identified by the industry coatings expert panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are discussed in this report and the application of these elements to the System 5 coating system (polyamide epoxy primer, carbon steel substrate) is used to evaluate performance.

  12. Degradation and failure characteristics of NPP containment protective coating systems

    International Nuclear Information System (INIS)

    Sindelar, R.L.

    2000-01-01

    A research program to investigate the performance and potential for failure of Service Level 1 coating systems used in nuclear power plant containment is in progress. The research activities are aligned to address phenomena important to cause failure as identified by the industry coatings expert panel. The period of interest for performance covers the time from application of the coating through 40 years of service, followed by a medium-to-large break loss-of-coolant accident scenario, which is a design basis accident (DBA) scenario. The interactive program elements are discussed in this report and the application of these elements to the System 5 coating system (polyamide epoxy primer, carbon steel substrate) is used to evaluate performance

  13. Design compliance matrix waste sample container filling system for nested, fixed-depth sampling system

    International Nuclear Information System (INIS)

    BOGER, R.M.

    1999-01-01

    This design compliance matrix document provides specific design related functional characteristics, constraints, and requirements for the container filling system that is part of the nested, fixed-depth sampling system. This document addresses performance, external interfaces, ALARA, Authorization Basis, environmental and design code requirements for the container filling system. The container filling system will interface with the waste stream from the fluidic pumping channels of the nested, fixed-depth sampling system and will fill containers with waste that meet the Resource Conservation and Recovery Act (RCRA) criteria for waste that contains volatile and semi-volatile organic materials. The specifications for the nested, fixed-depth sampling system are described in a Level 2 Specification document (HNF-3483, Rev. 1). The basis for this design compliance matrix document is the Tank Waste Remediation System (TWRS) desk instructions for design Compliance matrix documents (PI-CP-008-00, Rev. 0)

  14. Calculations of the SNL experiments Sup1 and Sup2 with CONTAIN 2

    International Nuclear Information System (INIS)

    Jacobs, G.; Noebel, R.; Wendlandt, T.

    2000-01-01

    Post-test calculations using the CONTAIN code were performed for the SNL melt dispersal/DCH tests SUP-1 und SUP-2, resulting in a workable input model for future applications to high-temperature melt dispersal experiments as well as for prototypes with tight annular reactor cavity geometries. (orig.) [de

  15. HECTR [Hydrogen Event: Containment Transient Response] analyses of the Nevada Test Site (NTS) premixed combustion experiments

    International Nuclear Information System (INIS)

    Wong, C.C.

    1988-11-01

    The HECTR (Hydrogen Event: Containment Transient Response) computer code has been developed at Sandia National Laboratories to predict the transient pressure and temperature responses within reactor containments for hypothetical accidents involving the transport and combustion of hydrogen. Although HECTR was designed primarily to investigate these phenomena in LWRs, it may also be used to analyze hydrogen transport and combustion experiments as well. It is in this manner that HECTR is assessed and empirical correlations, such as the combustion completeness and flame speed correlations for the hydrogen combustion model, if necessary, are upgraded. In this report, we present HECTR analyses of the large-scale premixed hydrogen combustion experiments at the Nevada Test Site (NTS) and comparison with the test results. The existing correlations in HECTR version 1.0, under certain conditions, have difficulty in predicting accurately the combustion completeness and burn time for the NTS experiments. By combining the combustion data obtained from the NTS experiments with other experimental data (FITS, VGES, ACUREX, and Whiteshell), a set of new and better combustion correlations was generated. HECTR prediction of the containment responses, using a single-compartment model and EPRI-provided combustion completeness and burn time, compares reasonably well against the test results. However, HECTR prediction of the containment responses using a multicompartment model does not compare well with the test results. This discrepancy shows the deficiency of the homogeneous burning model used in HECTR. To overcome this deficiency, a flame propagation model is highly recommended. 16 refs., 84 figs., 5 tabs

  16. Decision support system for containment and release management

    Energy Technology Data Exchange (ETDEWEB)

    Oosterhuis, B [Twente Univ., Enschede (Netherlands). Computer Science Dept.

    1995-09-01

    The Containment and Release Management project was carried out within the Reinforced Concerted Action Programme for Accident Management Support and partly financed by the European Union. In this report a prototype of an accident management support system is presented. The support system integrates several concepts from accident management research, like safety objective trees, severe accident phenomena, calculation models and an emergency response data system. These concepts are provided by the prototype in such a way that the decision making process of accident management is supported. The prototype application is demonstrated by process data taken from a severe accident scenario for a pressurized water reactor (PWR) that was simulated with the thermohydraulic computer program MAAP. The prototype was derived from a decision support framework based on a decision theory. For established and innovative concepts from accident management research it is pointed out in which way these concepts can support accident management and how these concepts can be integrated. This approach is generic in two ways; it applies to both pressurized and boiling water reactors and it applies to both in vessel management and containment and release management. The prototype application was developed in Multimedia Toolbox 3.0 and requires at least a 386 PC with 4 MB memory, 6 MB free disk space and MS Windows 3.1. (orig.).

  17. Decision support system for containment and release management

    International Nuclear Information System (INIS)

    Oosterhuis, B.

    1995-09-01

    The Containment and Release Management project was carried out within the Reinforced Concerted Action Programme for Accident Management Support and partly financed by the European Union. In this report a prototype of an accident management support system is presented. The support system integrates several concepts from accident management research, like safety objective trees, severe accident phenomena, calculation models and an emergency response data system. These concepts are provided by the prototype in such a way that the decision making process of accident management is supported. The prototype application is demonstrated by process data taken from a severe accident scenario for a pressurized water reactor (PWR) that was simulated with the thermohydraulic computer program MAAP. The prototype was derived from a decision support framework based on a decision theory. For established and innovative concepts from accident management research it is pointed out in which way these concepts can support accident management and how these concepts can be integrated. This approach is generic in two ways; it applies to both pressurized and boiling water reactors and it applies to both in vessel management and containment and release management. The prototype application was developed in Multimedia Toolbox 3.0 and requires at least a 386 PC with 4 MB memory, 6 MB free disk space and MS Windows 3.1. (orig.)

  18. Performance assessment of containment filtered venting system with Venturi scrubber

    International Nuclear Information System (INIS)

    Adinarayna, K.N.V.; Ali, Seik Mansoor; Balasubramaniyan, V.

    2015-01-01

    Venting through appropriate filtration systems is now being considered as a severe accident management strategy for maintaining the containment integrity and also as a means to reduce the radiological consequences to the public and environment. The option of filtered containment venting appears to have assumed significance in the post- Fukushima accident backdrop. Back-fitting of a suitable Venturi scrubber based CFVS for the Indian BWRs (TAPS- 1 and 2) at Tarapur is now being contemplated. Several key issues need to be carefully addressed for ensuring the desired functional capability of such a system. At the outset, this paper highlights a few thermal hydraulic issues that are of interest from regulatory perspective. This is followed by a detailed description of the mathematical models developed for assessing the depressurization characteristics of CFVS, energy absorption capacity of the Scrubber Tank (ST) water inventory, iodine removal and aerosol retention capability etc. Finally, application of these models to investigate the response of CFVS under twin unit SBO conditions in TAPS-1 and 2 is presented. The studies presented here give insight into the key variables affecting the CFVS performance and would be useful to both the system designer as well as the regulator. (author)

  19. Westinghouse containment filtered venting system wet scrubber technology

    International Nuclear Information System (INIS)

    Kristensson, S.; Nilsson, P-O.

    2014-01-01

    Following the Fukushima event Westinghouse has further developed and enhanced its filtered containment venting system (FCVS) product line. The filtration efficiency of the proven FILTRA-MVSS system installed at all Swedish NPPs as well as at the Muhelberg plant in Switzerland has been enhanced and a new wet scrubber design, SVEN (Safety Venting), based on the FILTRA-MVSS tradition, developed. To meet increased filtration requirements for organic iodine these two wet scrubber products have been complemented with a zeolite module. The offering of a select choice of products allows for a better adjustment to the specific constraints and needs of each nuclear power station that is planning for the installation of such a system. The FILTRA-MVSS (MVSS=Multi Venturi Scrubber System) is a wet containment filtered vent system that uses multiple venturies to create an interaction between the vent gases and the scrubber media allowing for removal of aerosols and gaseous iodines in a very efficient manner. The FILTRA-MVSS was originally developed to meet stringent requirements on autonomy and maintained filtration efficiency over a wide range of venting conditions. The system was jointly developed in the late 80's by ABB Atom and ABB Flaekt, today Westinghouse and Alstom. Following installations in Sweden and Switzerland the system was further developed by replacement of the gravel-bed moisture separator with a standard demister and by addition of a set of sintered metal fibre filter cartridges placed after the moisture separator step. The system is today offered as a modular steel tank design to simplify installation at site. To reduce complexity and delivery time Westinghouse has developed an alternative design in which the venturi module is replaced by a submerged metal fibre filter cartridges module. This new wet scrubber design, SVEN (patent pending), provides a flexible, compact, and lower weight system, while still preserving and even enhancing the filtration

  20. A remote inspection system for use inside reactor containment vessels

    International Nuclear Information System (INIS)

    Aoki, Toshihiko; Kashiwai, Jun-ichi; Yamamoto, Ikuo; Fukada, Koichi; Yamanaka, Yoshinobu.

    1985-01-01

    The harsh environment in the reactor-containment vesels of pressurized-water reactor nuclear-power plants precludes the possibility of direct circuit inspection; a remote-inspection system is essential. A robot for performing this task must not only be able to withstand the harsh conditions but must also be small and maneuverable enough to function effectively within complex and confined spaces. The article describes a monorail-type remote-inspection robot developed by Mitsubishi Electric to meet these needs, which is now under trial production and testing. (author)

  1. A stochastic killing system for biological containment of Escherichia coli

    DEFF Research Database (Denmark)

    Klemm, P.; Jensen, Lars Bogø; Molin, Søren

    1995-01-01

    Bacteria with a stochastic conditional lethal containment system have been constructed. The invertible switch promoter located upstream of the fimA gene from Escherichia coli was inserted as expression cassette in front of the Lethal gef gene deleted of its own natural promoter. The resulting...... fusion was placed on a plasmid and transformed to E. coli. The phenotype connected with the presence of such a plasmid was to reduce the population growth rate with increasing significance as the cell growth rate was reduced. In very fast growing cells, there was no measurable effect on growth rate. When...

  2. Container lid gasket protective strip for double door transfer system

    Science.gov (United States)

    Allen, Jr., Burgess M

    2013-02-19

    An apparatus and a process for forming a protective barrier seal along a "ring of concern" of a transfer container used with double door systems is provided. A protective substrate is supplied between a "ring of concern" and a safety cover in which an adhesive layer of the substrate engages the "ring of concern". A compressive foam strip along an opposite side of the substrate engages a safety cover such that a compressive force is maintained between the "ring of concern" and the adhesive layer of the substrate.

  3. System for indicating the level of material in a container

    International Nuclear Information System (INIS)

    Erb, T.L.

    1980-01-01

    In a radiation detecting system for controlling the level of material in a container, the first counter accumulates pulses generated by a geiger tube at a rate related to the level of material and a second counter accumulates clock pulses. A race condition is established between a NAND circuit indicating that the first counter has reached a predetermined total, and a NAND circuit indicating that the second counter has reached a second predetermined total representing a fixed counting interval. The first NAND circuit to respond to its predetermined total actuates a circuit to reset both counters and, if indicative of the material level being below a predetermined minimum, actuates an alarm or operates a control circuit to add material to the container. In the example shown, an additional NAND circuit responds to a different count in the first counter which count in the same time interval corresponds to a higher level, and when material is being added to the container, the race condition is between two NAND circuits. The effect of this is to provide a hysteresis effect preventing the circuit from 'hunting' around one level of material. (author)

  4. FIREX (Fire Influence on Regional and Global Environments Experiment): Measurements of Nitrogen Containing Volatile Organic Compounds

    Science.gov (United States)

    Warneke, C.; Schwarz, J. P.; Yokelson, R. J.; Roberts, J. M.; Koss, A.; Coggon, M.; Yuan, B.; Sekimoto, K.

    2017-12-01

    A combination of a warmer, drier climate with fire-control practices over the last century have produced a situation in which we can expect more frequent fires and fires of larger magnitude in the Western U.S. and Canada. There are urgent needs to better understand the impacts of wildfire and biomass burning (BB) on the atmosphere and climate system, and for policy-relevant science to aid in the process of managing fires. The FIREX (Fire Influence on Regional and Global Environment Experiment) research effort is a multi-year, multi-agency measurement campaign focused on the impact of BB on climate and air quality from western North American wild fires, where research takes place on scales ranging from the flame-front to the global atmosphere. FIREX includes methods development and small- and large-scale laboratory and field experiments. FIREX will include: emission factor measurements from typical North American fuels in the fire science laboratory in Missoula, Montana; mobile laboratory deployments; ground site measurements at sites influenced by BB from several western states. The main FIREX effort will be a large field study with multiple aircraft and mobile labs in the fire season of 2019. One of the main advances of FIREX is the availability of various new measurement techniques that allows for smoke evaluation in unprecedented detail. The first major effort of FIREX was the fire science laboratory measurements in October 2016, where a large number of previously understudied Nitrogen containing volatile organic compounds (NVOCs) were measured using H3O+CIMS and I-CIMS instruments. The contribution of NVOCs to the total reactive Nitrogen budget and the relationship to the Nitrogen content of the fuel are investigated.

  5. SWR 1000 related containment cooling system tests in PANDA

    International Nuclear Information System (INIS)

    Dreier, J.; Aubert, C.; Huggenberger, M.; Strassberger, H.J.; Yadigaroglu, G.

    2000-01-01

    Since 1991 the Paul Scherrer Institute has participated in the investigations of several of the new passive Advanced Light Water Reactor designs proposed world-wide. The current phase of the project, ALPHA-II, is focused on both the boiling water and the pressurized water reactor passive designs and consists of three projects under the sponsorship of the European Commission. The paper describes the performed PANDA transient system tests related to one of these projects, called 'BWR R and D Cluster for Innovative Passive Safety Systems (IPSS)', and details the PSI contribution to the experimental investigation of passive containment cooling by a Building Condenser system which is part of the advanced Boiling Water Reactor SWR 1000 designed by Siemens. First, a short description of the relevant systems of the SWR 1000 design and its simulation in the PANDA facility are presented. After the description of the experimental programme for the large-scale integral system test investigations in the PANDA facility, the main results of the performed tests are also given. Finally, the main conclusions, based on the to date available experimental results and their analysis, are summarised. (author)

  6. Controlled release systems containing solid dispersions: strategies and mechanisms.

    Science.gov (United States)

    Tran, Phuong Ha-Lien; Tran, Thao Truong-Dinh; Park, Jun Bom; Lee, Beom-Jin

    2011-10-01

    In addition to a number of highly soluble drugs, most new chemical entities under development are poorly water-soluble drugs generally characterized by an insufficient dissolution rate and a small absorption window, leading to the low bioavailability. Controlled-release (CR) formulations have several potential advantages over conventional dosage forms, such as providing a uniform and prolonged therapeutic effect to improve patient compliance, reducing the frequency of dosing, minimizing the number of side effects, and reducing the strength of the required dose while increasing the effectiveness of the drug. Solid dispersions (SD) can be used to enhance the dissolution rate of poorly water-soluble drugs and to sustain the drug release by choosing an appropriate carrier. Thus, a CR-SD comprises both functions of SD and CR for poorly water-soluble drugs. Such CR dosage forms containing SD provide an immediately available dose for an immediate action followed by a gradual and continuous release of subsequent doses to maintain the plasma concentration of poorly water-soluble drugs over an extended period of time. This review aims to summarize all currently known aspects of controlled release systems containing solid dispersions, focusing on the preparation methods, mechanisms of action and characterization of physicochemical properties of the system.

  7. Iodine removal in containment filtered venting system during nuclear accident

    International Nuclear Information System (INIS)

    Bera, Subrata; Deo, Anuj Kumar; Nagrale, D.B.; Paul, U.K.; Prasad, M.; Gaikwad, A.J.

    2015-01-01

    Post Fukushima nuclear accident, containment filtered venting system is being introduced in Indian nuclear power plant to strengthen the defense in depth safety barrier by depressurizing the containment building along with minimization of radioactivity release to environment during a severe accident. Radioactive iodine is one of the major contributors to radiation dose during early release phase of a severe accident. Physical and Chemical form of iodine and iodine bearing compounds includes particulates, elemental and organic. In the most efficient design of CFVS, wet scrubbing mechanism has been employed through use of venture scrubber. The Iodine removal process in wet scrubber involves two processes: chemical reaction in highly alkaline aqueous solution and impingement of particulates with water droplets produced in the venturi nozzle. In this paper, venturi has been modeled using the Calvert model. The variation of efficiency has been estimated for the different particle sizes. The impact of the shape parameter of log-normal distribution on the amount of scrubbed iodine has also been assessed. Release phase wise the scrubbed amount of iodine in the venturi based CFVS system has been estimated for a typical BWR. (author)

  8. AC-600 passive containment cooling system performance research

    International Nuclear Information System (INIS)

    Jia Baoshan; Yu Jiyang; Shi Junying

    1997-01-01

    a code named PCCSAC which is able to predict both the evaporating film on the outside surface of the vessel and the condensed film on its inside is developed successfully. It is a special software tool to analyze the passive containment cooling system (PCCS) performance in the design of AC-600. The author includes the establishment of physical models, selection of numerical methods, debugging and verification of the code and application of the code in the AC-600 PCCS. In physical models, the fundamental conservation equations about various areas and heat conduction equations are established. In order to make the equations to meet the closed form of solution, a lot of structure formulae are complemented. After repeated selection and demonstration of the numerical methods, the backward difference method Gear which is generally used for stiff problem is chosen for the solution of ordinary differential equations derived from the physical models. The results of standard example calculated by the PCCSAC code and the COMMIX code which is used to analyze westinghouse AP-600 are same in the main. The reliability and validity are verified from the calculations. The PCCSAC code is applied in the calculations of two important LOCA used in the containment safety analyses. The sensitivity of main parameters in the system based on LOCA are studied. All the results are reasonable and in agreement with the theoretical analyses. It can be concluded that the PCCSAC code is able to be used for the analyses of AC-600 PCCS performance

  9. Large-scale experiments on aerosol behavior in light water reactor containments

    International Nuclear Information System (INIS)

    Schock, W.; Bunz, H.; Adams, R.E.; Tobias, M.L.; Rahn, F.J.

    1988-01-01

    Recently, three large-scale experimental programs were carried out dealing with the behavior of aerosols during core-melt accidents in light water reactors (LWRs). In the Nuclear Safety Pilot Plant (NSPP) program, the principal behaviors of different insoluble aerosols and of mixed aerosols were measured in dry air atmospheres and in condensing steam-air atmospheres contained in a 38-m/sup 3/ steel vessel. The Demonstration of Nuclear Aerosol Behavior (DEMONA) program used a 640-m/sup 3/ concrete containment model to simulate typical accident sequence conditions, and measured the behavior of different insoluble aerosols and mixed aerosols in condensing and transient atmospheric conditions. Part of the LWR Aerosol Containment Experiments (LACE) program was also devoted to aerosol behavior in containment; and 852-m/sup 3/ steel vessel was used, and the aerosols were composed of mixtures of insoluble and soluble species. The results of these experiments provide a suitable data base for validation of aerosol behavior codes. Fundamental insight into details of aerosol behavior in condensing environments has been gained through the results of the NSPP tests. Code comparisons have been and are being performed in the DEMONA and LACE experiments

  10. Criticality experiments with annular cylinders containing plutonium solutions; Experiences de criticite sur des cylindres annulaires contenant des solutions de plutonium

    Energy Technology Data Exchange (ETDEWEB)

    Molbert, M; Sauve, A; Houelle, M; Deilgat, E [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1964-07-01

    The criticality station of Dijon involves three cells, shielded by concrete walls of 1.46 meter thickness. Those cells are designed to contain the criticality experiment apparatus. The engineering building is also involving: one chemical laboratory where plutonium solutions are prepared, one analysis laboratory, several activated solutions storages, several control rooms, One cell contains the B system, which is designed to study: annular cylindrical geometries, slab of 10 cm thickness, interaction between annular cylinders. This report includes the first results given by experiments on annular cylinders defined by their own geometry (outer and inner diameter of ring containing plutonium solutions). Those results have been plotted in curves, for several concentrations and for different reflection conditions (outer or inner light water reflector, cadmium screen), H{sub c} and M{sub c} = f (c) (where H{sub c} is the critical height of solution, M{sub c} is the critical mass, c is the plutonium concentration: 42,3 g/lexperiments on this cylinder being unfinished to the date of this present report publication. On this miscellaneous results, we have following informations know: - Screen effect of light water in central hole. Strengthened effect by cadmium foil on the inside wall. - Normalized interaction curves ( {alpha}*H{sub c}/H{sub c{infinity}} ) versus the distance between the two vessels, where H{sub c{infinity}} critical height of an insulated cylinder, shows that: 1) In light water, two cylinders set aside from 15 cm, can be considers like separated. 2) For some configurations, {alpha} vary

  11. ORNL experiments to characterize fuel release from the reactor primary containment in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wright, A.L.; Kress, T.S.; Smith, A.M.

    1980-01-01

    This paper presents results from aerosol source term experiments performed in the ORNL Aerosol Release and Transport (ART) Program sponsored by the US NRC. The tests described were performed to provide information on fuel release from an LMFBR primary containment as a result of a hypothetical core-disruptive accident (HCDA). The release path investigated in these tests assumes that a fuel/sodium bubble is formed after disassembly that transports fuel and fission products through the sodium coolant and cover gas to be relased into the reactor secondary containment. Due to the excellent heat transfer characteristics of the sodium, there is potential for large attenuation of the maximum release

  12. Ventilation Systems Operating Experience Review for Fusion Applications

    International Nuclear Information System (INIS)

    Cadwallader, L.C.

    1999-01-01

    This report is a collection and review of system operation and failure experiences for air ventilation systems in nuclear facilities. These experiences are applicable for magnetic and inertial fusion facilities since air ventilation systems are support systems that can be considered generic to nuclear facilities. The report contains descriptions of ventilation system components, operating experiences with these systems, component failure rates, and component repair times. Since ventilation systems have a role in mitigating accident releases in nuclear facilities, these data are useful in safety analysis and risk assessment of public safety. An effort has also been given to identifying any safety issues with personnel operating or maintaining ventilation systems. Finally, the recommended failure data were compared to an independent data set to determine the accuracy of individual values. This comparison is useful for the International Energy Agency task on fusion component failure rate data collection

  13. STUDIES ON VINYL POLYMERIZATION WITH INITIATION SYSTEM CONTAINING AMINE DERIVATIVES

    Institute of Scientific and Technical Information of China (English)

    QIU Kunyuan; ZHANG Jingyi; FENG Xinde(S. T. Voong)

    1983-01-01

    Two main types of amine-containing initiation systems were studied in this work. In the case of MMA polymerization initiated by BPO-amine (DMT, DHET, DMA) redox systems, it was found that the polymerization rate and colour stability of the polymer for different amine systems were in the following order: DMT≈DHET>DMA. Accordingly, BPO-DMT and BPO-DHET are effective initiators. In the case of MEMA polymerization by amine (DMT, DHET, DMA) alone, it was found that the polymerization rate and the percentage of conversion for these different amine systems were in the following order: DMT≥DHET>DMA. The polymerization rate and the percentage of conversion also increased with the increase of DMT concentration. From the kinetic investigation the rate equation of Rp=K [DMT]1/2 [MEMA]3/2 was obtained, and the overall activation energy of polymerization was calculated to be 34.3 KJ/mol (8.2 Kcal/mol). Moreover, the polymerization of MEMA in the presence of DMT was strongly inhibited by hydroquinone, indicating the polymerization being free radical in nature. From these results, the mechanism of MEMA polymerization initiated by amine was proposed.

  14. Cross System Extensions (CSE) experience

    International Nuclear Information System (INIS)

    Johnston, T.Y.

    1990-08-01

    Cross System Extension (CSE) provides VM/XA systems with the ability to share minidisks and spool in loosely coupled environment. CSE will also cooperate with the VM/HPO Inter System Facility (ISF) in sharing minidisks between VM/XA and VM/HPO to XA, reliability of CSE, and some operational considerations when running with it

  15. CONPAS 1.0 (CONtainment Performance Analysis System). User's manual

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Jin, Young Ho

    1996-04-01

    CONPAS (CONtainment Performance Analysis System) is a verified computer code package to integrate the numerical, graphical, and results-operation aspects of Level 2 probabilistic safety assessments (PSA) for nuclear power plants automatically under a PC window environment. Compared with the existing DOS-based computer codes for Level 2 PSA, the most important merit of the window-based computer code is that user can easily describe and quantify the accident progression models, and manipulate the resultant outputs in a variety of ways. As a main logic for accident progression analysis, CONPAS employs a concept of the small containment phenomenological event tree (CPET) helpful to trace out visually individual accident progressions and of the large supporting event tree (LSET) for its detailed quantification. For the integrated analysis of Level 2 PSA, the code utilizes four distinct, but closely related modules; (1) ET Editor for construction of several event tree models describing the accident progressions, (2) Computer for quantification of the constructed event trees and graphical display of the resultant outputs, (3) Text Editor for preparation of input decks for quanification and utilization of calculational results, and (4) Mechanistic Code Plotter for utilization of results obtained from severe accident analysis codes. Compared with other existing computer codes for Level 2 PSA, the CONPAS code provides several advanced features: computational aspects including systematic uncertainty analysis, importance analysis, sensitivity analysis and data interpretation, reporting aspects including tabling and graphic as well as user-friend interface. 10 refs. (Author) .new

  16. Model experiment and numerical simulation of drop impact response of multilayer-combinational container

    International Nuclear Information System (INIS)

    Xie Ruoze; Zhong Weizhou; Wan Qiang; Huang Xicheng; Zhang Fangju

    2015-01-01

    The drop impact process of multilayer-combinational container was simulated experimentally using a gas gun, and the normal impact and oblique impact of scaled models were tested. The experiments of scaled models were simulated numerically, and the stress distribution and plastic deformation in the tested structures during collision process were obtained. The results were compared with the experiment data. It was shown that the impact work mainly converted into plastic work due to the plastic deformation of the cushion wood and the plastic hinge in the buckled steel shell. The plastic deformation mainly happened at the collided end of the scaled models, and there was no plastic deformation found far from the collided end. The compressive stress-strain curve of the wood in texture direction can be used to simulate numerically the drop impact process of multilayer-combinational container. (authors)

  17. Supramolecular effects in dendritic systems containing photoactive groups

    Directory of Open Access Journals (Sweden)

    GIANLUCA CAMILLO AZZELLINI

    2000-03-01

    Full Text Available In this article are described dendritic structures containing photoactive groups at the surface or in the core. The observed supramolecular effects can be attributed to the nature of the photoactive group and their location in the dendritic architecture. The peripheric azobenzene groups in these dendrimeric compounds can be regarded as single residues that retain the spectroscopic and photochemical properties of free azobenzene moiety. The E and Z forms of higher generation dendrimer, functionalized with azobenzene groups, show different host ability towards eosin dye, suggesting the possibility of using such dendrimer in photocontrolled host-guest systems. The photophysical properties of many dendritic-bipyridine ruthenium complexes have been investigated. Particularly in aerated medium more intense emission and a longer excited-state lifetime are observed as compared to the parent unsubstituted bipyridine ruthenium complexes. These differences can be attributed to a shielding effect towards dioxygen quenching originated by the dendritic branches.

  18. Design and hydrodynamic testing of an oil slick containment system

    International Nuclear Information System (INIS)

    Allen-Jones, J.

    1997-01-01

    Aspects of mechanical containment of spilled oil were studied. The focus was on design problems and the development of a model for global loading on a horizontal catenary of a previously defined form. The result is then compared with existing theoretical formulations and an approximate model is developed for the effect of flow through the system in deep water. The modified result is again compared with accepted formulations and with sea-trial data. The leading edge of the skirt was observed to oscillate sinusoidally. Experimental results obtained from pressure transducer data and calibrated underwater video measurements show that the oscillation period diminishes with increases in tow speed. In contrast, the magnitude of the oscillation increases while mean deviation from datum draught returns to zero. 14 refs., 5 tabs., 31 figs

  19. Validation of NCSSHP for highly enriched uranium systems containing beryllium

    International Nuclear Information System (INIS)

    Krass, A.W.; Elliott, E.P.; Tollefson, D.A.

    1994-01-01

    This document describes the validation of KENO V.a using the 27-group ENDF/B-IV cross section library for highly enriched uranium and beryllium neutronic systems, and is in accordance with ANSI/ANS-8.1-1983(R1988) requirements for calculational methods. The validation has been performed on a Hewlett Packard 9000/Series 700 Workstation at the Oak Ridge Y-12 Plant Nuclear Criticality Safety Department using the Oak Ridge Y-12 Plant Nuclear Criticality Safety Software code package. Critical experiments from LA-2203, UCRL-4975, ORNL-2201, and ORNL/ENG-2 have been identified as having the constituents desired for this validation as well as sufficient experimental detail to allow accurate construction of KENO V.a calculational models. The results of these calculations establish the safety criteria to be employed in future calculational studies of these types of systems

  20. Loop containment (joint integrity) assessment Brayton Isotope Power System flight system

    International Nuclear Information System (INIS)

    1976-01-01

    The Brayton Isotope Power System (BIPS) contains a large number of joints. Since the failure of a joint would result in loss of the working fluid and consequential failure of the BIPS, the integrity of the joints is of paramount importance. The reliability of the ERDA BIPS loop containment (joint integrity) is evaluated. The conceptual flight system as presently configured is depicted. A brief description of the flight system is given

  1. The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Parozzi, Flavio, E-mail: flavio.parozzi@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Caracciolo, Eduardo D.J., E-mail: eduardo.caracciolo@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA Cadarache (France); Piluso, Pascal, E-mail: pascal.piluso@cea.fr [CEA Cadarache (France)

    2013-08-15

    Highlights: ► Experiment investigating aerosol retention within concrete containment cracks under nuclear severe accident conditions. ► Provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. ► Prototypical aerosol particles generated with a thermite reaction and transported through the crack sample reproducing surface characteristics, temperature, pressure drop and gas leakage. ► The results indicate the significant retention due to zig-zag path. -- Abstract: CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced. The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

  2. An Artificial Channel Experiment for Purifying Drainage Water Containing Arsenic by Using Eleocharis acicularis

    Science.gov (United States)

    Okazaki, Kenji; Yamazaki, Shusaku; Kurahashi, Toshiyuki; Sakakibara, Masayuki

    2017-06-01

    This paper reports the results of an artificial channel experiment in which water containing arsenic was purified by using Eleocharis acicularis. The experiment was conducted to investigate the feasibility of phytoremediation by Eleocharis acicularis in civil engineering projects. In the experiment, 15 m2 of Eleocharis acicularis mats were laid in an artificial channel. Three sessions of artificial flow were implemented by leading 100.0 L of river water containing 0.234 mg/L of arsenic into the channel each time. The arsenic concentration of the leachate from the channel was analyzed. As the results of experiment, the arsenic concentrations of the leachate for the three sessions were 0.045 mg/L, 0.133 mg/L, and 0.249 mg/L. This shows that the arsenic concentration decreased during the first two sessions, whose flow totaled 200 L. The arsenic concentrations in the Eleocharis acicularis were 0.87 mg/kg, 1.01 mg/kg, and 4.16 mg/kg, which show that the plant absorbs arsenic. Moreover, it was found that the amount of sample water was reduced through evapotranspiration from the plant and the artificial channel.

  3. Characterization and stability studies of emulsion systems containing pumice

    Directory of Open Access Journals (Sweden)

    Marilene Estanqueiro

    2014-04-01

    Full Text Available Emulsions are the most common form of skin care products. However, these systems may exhibit some instability. Therefore, when developing emulsions for topical application it is interesting to verify whether they have suitable physical and mechanical characteristics and further assess their stability. The aim of this work was to study the stability of emulsion systems, which varied in the proportion of the emulsifying agent cetearyl alcohol (and sodium lauryl sulfate (and sodium cetearyl sulfate (LSX, the nature of the oily phase (decyl oleate, cyclomethicone or dimethicone and the presence or absence of pumice (5% w/w. While maintaining the samples at room temperature, rheology studies, texture analysis and microscopic observation of formulations with and without pumice were performed. Samples were also submitted to an accelerated stability study by centrifugation and to a thermal stress test. Through the testing, it was found that the amount of emulsifying agent affects the consistency and textural properties such as firmness and adhesiveness. So, formulations containing LSX (5% w/w and decyl oleate or dimethicone as oily phase had a better consistency and remained stable with time, so exhibited the best features to be used for skin care products.

  4. TRAC analysis of passive containment cooling system performance

    International Nuclear Information System (INIS)

    Arai, Kenji; Kataoka, Kazuyoshi; Nagasaka, Hideo

    1993-01-01

    A passive containment cooling system (PCCS) is a promising concept to improve the reliability of decay heat removal during an accident. Toshiba has carried out analytical studies for PCCS development in addition to experimental studies, using a best estimate thermal hydraulic computer code TRAC. In order to establish an analytical model for the PCCS performance analysis, it is necessary for the analytical model to be qualified against experimental results and thoroughly address the phenomena important for PCCS performance analysis. In this paper, the TRAC qualification for PCCS application is reported. A TRAC model has been verified against a drain line break simulation test conducted at the PCCS integral test facility, GIRAFFE. The result shows that the TRAC model can accurately predict the major system response and the PCCS performance in the drain line break test. In addition, the results of several sensitivity analyses, showing various points concerning the modeling in the PCCS performance analysis, have been reported. The analyses have been carried out for the SBWR and the analytical points are closely related to important phenomena which can affect PCCS performance

  5. Humos monitoring system of leaks in to the containment atmosphere

    International Nuclear Information System (INIS)

    Matal, O.; Zaloudek, J.; Matal, O. Jr.; Klinga, J.; Brom, J.

    1997-01-01

    HUmidity MOnitoring System (HUMOS) has been developed and designed to detect the presence of leak in selected primary circuit high energy pipelines and components that are evaluated from the point of view of Leak Before Break (LBB) requirements. It also requires to apply technical tools for detection and identification of coolant leaks from primary circuit and components of PWRs reactors. Safety significant of leaks depend on: leak source (location); leak rate, and leak duration. Therefore to detect and monitor coolant leaks in to the containment atmosphere during reactor operation is one of important safety measures. As potential leak sources flange connection in the upper head region of WWER reactors can be considered. HUMOS does not rely on the release of radioactivity to detect leaks but rather the relies on detection of moisture in the air resulting from a primary boundary leak. Because HUMOS relies on moisture and temperature detection, leaks can be detected without requiring the reactor to be critical. Therefore leaks can be detected during integrity pressure tests and any other mode of operation provided the reactor ventilation system is operating and primary circuit and components are pressurized. 3 figs

  6. Cost Optimal System Identification Experiment Design

    DEFF Research Database (Denmark)

    Kirkegaard, Poul Henning

    A structural system identification experiment design method is formulated in the light of decision theory, structural reliability theory and optimization theory. The experiment design is based on a preposterior analysis, well-known from the classical decision theory. I.e. the decisions concerning...... reflecting the cost of the experiment and the value of obtained additional information. An example concerning design of an experiment for parametric identification of a single degree of freedom structural system shows the applicability of the experiment design method....... the experiment design are not based on obtained experimental data. Instead the decisions are based on the expected experimental data assumed to be obtained from the measurements, estimated based on prior information and engineering judgement. The design method provides a system identification experiment design...

  7. Containment analysis on the PHEBUS FPT-0, FPT-1 and FPT-2 experiments

    International Nuclear Information System (INIS)

    Gyenes, Gyorgy; Ammirabile, Luca

    2011-01-01

    Research highlights: → The CPA/ASTEC code can reproduce similar patterns of CFD-based codes. → The deposition on elliptic bottom and on the painted wet condenser are qualitatively predicted. → The gas circulation affects the quick mixing of aerosols in the containment atmosphere. → The flow fields in CPA/ASTEC have a medium impact on the airborne mass in the PHEBUS containment. - Abstract: In a severe accident, most of the fission-product species are already condensed in aerosols when they are released to the containment. The behaviour of these aerosol particles controls the fission-product transport into the containment and affects the global Source Term. The calculations presented here were performed using the CPA module (Containment Package implemented in the European integral code ASTEC) for the in-pile PHEBUS FPT-0, FPT-1 and FPT-2 experiments and are focused on the aerosol transport. A detailed thermal-hydraulic model was used in the CPA/ASTEC code to evaluate the gas circulation pattern in the closed containment volume. The comparison of ASTEC results showed that the patterns are similar to the ones predicted by the CFD-based codes. Good agreement was reached with the measured average thermo-hydraulic parameters such as containment gas pressure, temperature and the condensation rate on the condensers. The calculations with the detailed simulation of the flow in the PHEBUS containment qualitatively predicted the particle settling on the elliptic bottom and deposition on the painted wet condenser surfaces. It was shown that the influence of the gas circulation leads to a relatively quick mixing of aerosols in the containment atmosphere. In the tests investigated, the effect of the gas circulation on the airborne aerosol mass during the aerosol injection period is small because the injected mass flux is significantly higher compared to the deposition fluxes on the vessel surfaces. During the long-term aerosol deposition phase, the flow fields predicted

  8. Modelling of containment atmosphere mixing and stratification experiment using CFD approach

    International Nuclear Information System (INIS)

    Ivo Kljenak; Miroslav Babic; Borut Mavko; Ivan Bajsic

    2005-01-01

    An experiment on containment atmosphere mixing and stratification, which was originally performed in the TOSQAN facility in Saclay (France), was simulated with the Computational Fluid Dynamics code CFX. The TOSQAN facility consists of a large cylindrical vessel in which gases are injected. In the considered experiment, steam, air and helium were injected during different phases of the experiment, with steam condensing on vessel walls. Three intermediate steady states, which were obtained with different boundary conditions, were simulated independently. A two-dimensional axisymmetric model of the TOSQAN vessel for the CFX4.4 code was developed. The flow in the simulation domain was modelled as single-phase. Steam condensation on vessel walls was modelled as a sink of mass and energy. Calculated profiles of temperature, steam concentration, and velocity components are compared to experimental results. (authors)

  9. Experimental study on iodine chemistry (EXSI) - Containment experiments with methyl iodide

    Energy Technology Data Exchange (ETDEWEB)

    Holm, J.; Ekberg, C. (Chalmers Univ. of Technology, Goeteborg (Sweden)); Kaerkelae, T.; Auvinen, A. (VTT, Espoo (Finland)); Glaenneskog, H. (Vattenfall Power Consultant, Goeteborg (Sweden))

    2011-05-15

    An experimental study on radiolytic decomposition of methyl iodide was conducted in co-operation between VTT and Chalmers University of Technology as a part of the NKS-R programs. In year 2008 the NROI project, a Nordic collaboration studying iodine chemistry in the containment, was started. During year 2008 (NROI-1) the radiolytic oxidation of elemental iodine was investigated and during 2009 (NROI-2), the radiolytic oxidation of organic iodine was studied. This project (NROI-3) is a continuation of the investigation of the oxidation of organic iodine. The project has been divided into two parts. 1. The aims of the first part were to investigate the effect of ozone and UV-radiation, in dry and humid conditions, on methyl iodide. 2. The second project was about gamma radiation (approx20 kGy/h) and methyl iodide in dry and humid conditions. 1. Experimental results showed that the methyl iodide concentration in the facility was reduced with increasing temperature and increasing UV-radiation intensity. Similar behaviour occurred when ozone was present in the system. Formed organic gas species during the decomposition of methyl iodide was mainly formaldehyde and methanol. The particle formation was instant and extensive when methyl iodide was exposed to ozone and/or radiation at all temperatures. The size of the formed primary particles was about 10 nm and the size of secondary particles was between 50-200 nm. From the SEM-EDX analyses of the particles, the conclusion was drawn that these were some kind of iodine oxides (I{sub xO{sub y}). However, the correct speciation of the formed particles was difficult to obtain because the particles melted and fused together under the electron beam. 2. The results from this sub-project are more inconsistent and hard to interpret. The particle formation was significant lesser than corresponding experiments when ozone/UV-radiation was used instead of gamma radiation. The transport of gaseous methyl iodide through the facility was

  10. Warnings on alcohol containers and advertisements: international experience and evidence on effects.

    Science.gov (United States)

    Wilkinson, Claire; Room, Robin

    2009-07-01

    In light of possible introduction of alcohol warning labels in Australia and New Zealand, this paper discusses the international experience with and evidence of effects of alcohol warning labels. The report describes international experience with providing information and warnings concerning the promotion or sale of alcoholic beverages, and considers the evidence on the effects of such information and warnings. The experience with and evaluations of the effects of tobacco warning labels are also considered. The most methodologically sound evaluations of alcohol warning labels are based on the US experience. Although these evaluations find little evidence that the introduction of the warning label in the USA had an impact on drinking behaviour, there is evidence that they led to an increase in awareness of the message they contained. In contrast, evaluations of tobacco warning labels find clear evidence of effects on behaviour. There is a need and opportunity for a rigorous evaluation of the impacts of introducing alcohol warning labels to add to the published work on their effectiveness. The experience with tobacco labels might guide the way for more effective alcohol warning labels. Alcohol warning labels are an increasingly popular alcohol policy initiative. It is clear that warning labels can be ineffective, but the tobacco experience suggests that effective warning labels are possible. Any introduction of alcohol warning labels should be evaluated in terms of effects on attitudes and behaviour.

  11. Apollo experience report: Food systems

    Science.gov (United States)

    Smith, M. C., Jr.; Rapp, R. M.; Huber, C. S.; Rambaut, P. C.; Heidelbaugh, N. D.

    1974-01-01

    Development, delivery, and use of food systems in support of the Apollo 7 to 14 missions are discussed. Changes in design criteria for this unique program as mission requirements varied are traced from the baseline system that was established before the completion of the Gemini Program. Problems and progress in subsystem management, material selection, food packaging, development of new food items, menu design, and food-consumption methods under zero-gravity conditions are described. The effectiveness of various approaches in meeting food system objectives of providing flight crews with safe, nutritious, easy to prepare, and highly acceptable foods is considered. Nutritional quality and adequacy in maintaining crew health are discussed in relation to the establishment of nutritional criteria for future missions. Technological advances that have resulted from the design of separate food systems for the command module, the lunar module, The Mobile Quarantine Facility, and the Lunar Receiving Laboratory are presented for application to future manned spacecraft and to unique populations in earthbound situations.

  12. The nonlinear finite element analysis program NUCAS (NUclear Containment Analysis System) for reinforced concrete containment building

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Jin; Lee, Hong Pyo; Seo, Jeong Moon [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-03-01

    The maim goal of this research is to develop a nonlinear finite element analysis program NUCAS to accurately predict global and local failure modes of containment building subjected to internal pressure. In this report, we describe the techniques we developed throught this research. An adequate model to the analysis of containment building such as microscopic material model is adopted and it applied into the development Reissner-Mindlin degenerated shell element. To avoid finite element deficiencies, the substitute strains based on the assumed strain method is used in the shell formulation. Arc-length control method is also adopted to fully trace the peak load-displacement path due to crack formation. In addition, a benchmark test suite is developed to investigate the performance of NUCAS and proposed as the future benchmark tests for nonlinear analysis of reinforced concrete. Finally, the input format of NUCAS and the examples of input/output file are described. 39 refs., 65 figs., 8 tabs. (Author)

  13. Simulation of the containment spray system test PACOS PX2.2 with the integral code ASTEC and the containment code system COCOSYS

    International Nuclear Information System (INIS)

    Risken, Tobias; Koch, Marco K.

    2011-01-01

    The reactor safety research contains the analysis of postulated accidents in nuclear power plants (npp). These accidents may involve a loss of coolant from the nuclear plant's reactor coolant system, during which heat and pressure within the containment are increased. To handle these atmospheric conditions, containment spray systems are installed in various light water reactors (LWR) worldwide as a part of the accident management system. For the improvement and the safety ensurance in npp operation and accident management, numeric simulations of postulated accident scenarios are performed. The presented calculations regard the predictability of the containment spray system's effect with the integral code ASTEC and the containment code system COCOSYS, performed at Ruhr-Universitaet Bochum. Therefore the test PACOS Px2.2 is simulated, in which water is sprayed in the stratified containment atmosphere of the BMC (Battelle Modell-Containment). (orig.)

  14. Experiences with Ada in an embedded system

    Science.gov (United States)

    Labaugh, Robert J.

    1988-01-01

    Recent experiences with using Ada in a real time environment are described. The application was the control system for an experimental robotic arm. The objectives of the effort were to experiment with developing embedded applications in Ada, evaluating the suitability of the language for the application, and determining the performance of the system. Additional objectives were to develop a control system based on the NASA/NBS Standard Reference Model for Telerobot Control System Architecture (NASREM) in Ada, and to experiment with the control laws and how to incorporate them into the NASREM architecture.

  15. Criticality experiments with low enriched UO2 fuel rods in water containing dissolved gadolinium

    International Nuclear Information System (INIS)

    Bierman, S.R.; Murphy, E.S.; Clayton, E.D.; Keay, R.T.

    1984-02-01

    The results obtained in a criticality experiments program performed for British Nuclear Fuels, Ltd. (BNFL) under contract with the United States Department of Energy (USDOE) are presented in this report along with a complete description of the experiments. The experiments involved low enriched UO 2 and PuO 2 -UO 2 fuel rods in water containing dissolved gadolinium, and are in direct support of BNFL plans to use soluble compounds of the neutron poison gadolinium as a primary criticality safeguard in the reprocessing of low enriched nuclear fuels. The experiments were designed primarily to provide data for validating a calculation method being developed for BNFL design and safety assessments, and to obtain data for the use of gadolinium as a neutron poison in nuclear chemical plant operations - particularly fuel dissolution. The experiments program covers a wide range of neutron moderation (near optimum to very under-moderated) and a wide range of gadolinium concentration (zero to about 2.5 g Gd/l). The measurements provide critical and subcritical k/sub eff/ data (1 greater than or equal to k/sub eff/ greater than or equal to 0.87) on fuel-water assemblies of UO 2 rods at two enrichments (2.35 wt % and 4.31 wt % 235 U) and on mixed fuel-water assemblies of UO 2 and PuO 2 -UO 2 rods containing 4.31 wt % 235 U and 2 wt % PuO 2 in natural UO 2 respectively. Critical size of the lattices was determined with water containing no gadolinium and with water containing dissolved gadolinium nitrate. Pulsed neutron source measurements were performed to determine subcritical k/sub eff/ values as additional amounts of gadolinium were successively dissolved in the water of each critical assembly. Fission rate measurements in 235 U using solid state track recorders were made in each of the three unpoisoned critical assemblies, and in the near-optimum moderated and the close-packed poisoned assemblies of this fuel

  16. Smart integrated containment leakage rate test system using wireless communication

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Hwan; Lee, Sang Yong; Kim, Jung Sun; Kim, Gun Soo; Kim, Jong Myeong; Ahn, Jong Han [Research and Development Center, Ulsan (Korea, Republic of)

    2012-10-15

    Integrated Leakage Rate Test (ILRT) is the important test the confidentiality and integrity of the containment building, which is the last barrier when Design basis accidents (DBA) of Nuclear Power plant occur. Since the result of this test is the basis to guarantee the safety of nuclear power plants, the test process, test procedure, and the test equipment are required to have high reliability. The test devices previously used have been products of VOLUMERTRICS and GRAFTEL of USA. These devices have been inconvenient to calibrate and use. Thus improved devices needed to be developed to remove the inconveniences, to verify the safety of Korean nuclear power plants with Korea's own technology, and to secure core technology. A new leak test system was developed by domestic technology for that purpose and needed to be verified. In this paper, technical details of the newly developed easy to use and highly reliable measuring test device, which is in operation at the nuclear power plant sites, will be introduced. State of art technology was applied to the device to address the shortcomings of previous US made devices and the difficulties to use on site.

  17. Dynamic testing of MFTF containment-vessel structural system

    International Nuclear Information System (INIS)

    Weaver, H.J.; McCallen, D.B.; Eli, M.W.

    1982-01-01

    Dynamic (modal) testing was performed on the Magnetic Fusion Test Facility (MFTF) containment vessel. The seismic design of this vessel was heavily dependent upon the value of structural damping used in the analysis. Typically for welded steel vessels, a value of 2 to 3% of critical is used. However, due to the large mass of the vessel and magnet supported inside, we felt that the interaction between the structure and its foundation would be enhanced. This would result in a larger value of damping because vibrational energy in the structure would be transferred through the foundation into the surrounding soil. The dynamic test performed on this structure (with the magnet in place) confirmed this later theory and resulted in damping values of approximately 4 to 5% for the whole body modes. This report presents a brief description of dynamic testing emphasizing the specific test procedure used on the MFTF-A system. It also presents an interpretation of the damping mechanisms observed (material and geometric) based upon the spatial characteristics of the modal parameters

  18. Fuel salt and container material studies for MOSART transforming system

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

    2013-07-01

    A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

  19. Throughput Evaluation of an Autonomous Sustainment Cargo Container System

    National Research Council Canada - National Science Library

    Yeh, Mingtze

    2007-01-01

    .... Autonomous containers will play an essential role in the ability to deliver logistical supplies to waterborne littoral vessels enabling them to maintain station and complete there military operations...

  20. Design of database management system for 60Co container inspection system

    International Nuclear Information System (INIS)

    Liu Jinhui; Wu Zhifang

    2007-01-01

    The function of the database management system has been designed according to the features of cobalt-60 container inspection system. And the software related to the function has been constructed. The database querying and searching are included in the software. The database operation program is constructed based on Microsoft SQL server and Visual C ++ under Windows 2000. The software realizes database querying, image and graph displaying, statistic, report form and its printing, interface designing, etc. The software is powerful and flexible for operation and information querying. And it has been successfully used in the real database management system of cobalt-60 container inspection system. (authors)

  1. 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

  2. Swimming in a contained space: Understanding the experience of indoor lap swimmers.

    Science.gov (United States)

    Ward, Miranda

    2017-07-01

    Drawing on ethnographic work, this paper explores the convergence of bodies, materialities and practices found at the indoor swimming pool - a space that has not often been the subject of geographical study, in spite of the fact that swimming is one of the most popular forms of exercise in countries such as the UK. The paper focuses on the "contained" nature of the indoor pool environment, examining the distinct experience this can create for lap swimmers. This focus is placed in the context of a broader politics of exercise, with an emphasis on the popularity and potential benefits of swimming, as well as less encouraging facts about participation and facility provision, suggesting that in order to encourage further uptake of swimming and preservation of swimming facilities the voices and experiences of regular swimmers should be considered. Copyright © 2016 Elsevier Ltd. All rights reserved.

  3. The TN-GEMINI: experience on a versatile alpha waste transport container

    International Nuclear Information System (INIS)

    Roland, V.; Chanzy, Y.

    2001-01-01

    The present paper discusses experience gained in moving alpha wastes and its teachings regarding transport aspects of D and D. Alpha wastes are generated in fuel cycle facilities such as those involved in reprocessing, in manufacture of mixed oxide fuel, and by research laboratories. If a significant amount of wastes has to be transported, then a Type B packaging is required. Developed by Transnucleaire and COGEMA, the TN GEMINI container enables nuclear facilities operators to optimise their alpha waste transport management, and more generally contribute to their D and D projects. After describing succinctly the design of the TN GEMINI, the paper will explain how the packaging is being operated. Teachings from experience will be shared. (orig.)

  4. Microbial analysis of the buffer/container experiment at AECL's underground research laboratory

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, S.

    1996-07-01

    The Buffer/Container Experiment (BCE) was carried out at AECL's Underground Research Laboratory (URL) for 2.5 years to examine the in situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived the conditions (i.e., compaction, heat and desiccation) in the BCE and to determine which group(s) of microorganisms would be dominant in such a simulated vault environment. Such knowledge will be very useful in assessing the potential effects of microbial activity on the concept for deep disposal of Canada's nuclear fuel waste, proposed by AECL. 46 refs., 31 tabs., 35 figs

  5. Design and operational experience with a portable tritium cleanup system

    International Nuclear Information System (INIS)

    Maienschein, J.L.; Wilson, S.W.; Garcia, F.

    1991-06-01

    We built a portable tritium cleanup system to scavenge tritium from contaminated gases in any tritium-containing system in the LLNL Tritium Facility. The cleanup system uses standard catalytic oxidation of tritium to water followed by water removal with a molecular sieve dryer. The cleanup unit, complete with instrumentation, is contained in a portable cart that is rolled into place and connected to the apparatus to be cleaned. The cleanup systems is effective, low-tech, simple, and reliable. The nominal flow rate of the system is 30 liters/minute, and the decontamination factor is > 1000. In this paper we will show design information on our portable cleanup system, and will discuss our operational experience with it over the past several years

  6. A self-contained, programmable microfluidic cell culture system with real-time microscopy access

    DEFF Research Database (Denmark)

    Skafte-Pedersen, Peder; Hemmingsen, Mette; Sabourin, David

    2011-01-01

    Utilizing microfluidics is a promising way for increasing the throughput and automation of cell biology research. We present a complete self-contained system for automated cell culture and experiments with real-time optical read-out. The system offers a high degree of user-friendliness, stability...... enables the system to perform parallel, programmable and multiconditional assays on a single chip. A modular approach provides system versatility and allows many different chips to be used dependent upon application. We validate the system's performance by demonstrating on-chip passive switching...... and mixing by peristaltically driven flows. Applicability for biological assays is demonstrated by on-chip cell culture including on-chip transfection and temporally programmable gene expression....

  7. Explaining the user experience of recommender systems

    NARCIS (Netherlands)

    Knijnenburg, B.P.; Willemsen, M.C.; Gantner, Z.; Soncu, H.; Newell, C.

    2012-01-01

    Research on recommender systems typically focuses on the accuracy of prediction algorithms. Because accuracy only partially constitutes the user experience of a recommender system, this paper proposes a framework that takes a user-centric approach to recommender system evaluation. The framework

  8. High temperature experiments on a 4 tons UF6 container TENERIFE program

    Energy Technology Data Exchange (ETDEWEB)

    Casselman, C.; Duret, B.; Seiler, J.M.; Ringot, C.; Warniez, P.

    1991-12-31

    The paper presents an experimental program (called TENERIFE) whose aim is to investigate the behaviour of a cylinder containing UF{sub 6} when exposed to a high temperature fire for model validation. Taking into account the experiments performed in the past, the modelization needs further information in order to be able to predict the behaviour of a real size cylinder when engulfed in a 800{degrees}C fire, as specified in the regulation. The main unknowns are related to (1) the UF{sub 6} behaviour beyond the critical point, (2) the relationship between temperature field and internal pressure and (3) the equivalent conductivity of the solid UF{sub 6}. In order to investigate these phenomena in a representative way it is foreseen to perform experiments with a cylinder of real diameter, but reduced length, containing 4 tons of UF{sub 6}. This cylinder will be placed in an electrically heated furnace. A confinement vessel prevents any dispersion of UF{sub 6}. The heat flux delivered by the furnace will be calibrated by specific tests. The cylinder will be changed for each test.

  9. An Undergraduate Experiment in Alarm System Design.

    Science.gov (United States)

    Martini, R. A.; And Others

    1988-01-01

    Describes an experiment involving data acquisition by a computer, digital signal transmission from the computer to a digital logic circuit and signal interpretation by this circuit. The system is being used at the Illinois Institute of Technology. Discusses the fundamental concepts involved. Demonstrates the alarm experiment as it is used in…

  10. Image processing in 60Co container inspection system

    International Nuclear Information System (INIS)

    Wu Zhifang; Zhou Liye; Wang Liqiang; Liu Ximing

    1999-01-01

    The authors analyzes the features of 60 Co container inspection image, the design of several special processing methods for container image and some normal processing methods for two-dimensional digital image, including gray enhancement, pseudo-enhancement, space filter, edge enhancement, geometry process, etc. It gives out the way to carry out the above mentioned process in Windows 95 or Win NT. It discusses some ways to improve the image processing speed on microcomputer and good results were obtained

  11. A description of the apparatus to be used in interaction experiments with the ABC laser system

    International Nuclear Information System (INIS)

    Caruso, A.; Strangio, M.; Andreoli, P.L.; Cerioni, I.; Di Paolo, A.; Di Virgilio, L.

    1988-01-01

    This report contains the part of the Frascati Laboratorio Fusione Laser activity related to the Apparatus (target chamber, position and alignement system, diagnostics) to be used in the interaction experiments with the ABC laser system

  12. Irradiation experiment on fast reactor metal fuels containing minor actinides up to 7 at.% burnup

    International Nuclear Information System (INIS)

    Ohta, H.; Yokoo, T.; Ogata, T.; Inoue, T.; Ougier, M.; Glatz, J.P.; Fontaine, B.; Breton, L.

    2007-01-01

    Fast reactor metal fuels containing minor actinides (MAs: Np, Am, Cm) and rare earths (REs) have been irradiated in the fast reactor PHENIX. In this experiment, four types of fuel alloys, U-19Pu-10Zr, U-19Pu-10Zr-2MA-2RE, U-19Pu-10Zr-5MA-5RE and U-19Pu-10Zr-5MA (wt.%), are loaded into part of standard metal fuel stacks. The postirradiation examinations will be conducted at ∼2.4, ∼7 and ∼11 at.% burnup. As for the low-burnup fuel pins, nondestructive postirradiation tests have already been performed and the fuel integrity was confirmed. Furthermore, the irradiation experiment for the intermediate burnup goal of ∼7 at.% was completed in July 2006. For the irradiation period of 356.63 equivalent full-power days, the neutron flux level remained in the range of 3.5-3.6 x 10 15 n/cm 2 /s at the axial peak position. On the other hand, the maximum linear power of fuel alloys decreased gradually from 305-315 W/cm (beginning of irradiation) to 250-260 W/cm (end of irradiation). The discharged peak burnup was estimated to be 6.59-7.23 at.%. The irradiation behavior of MA-containing metal fuels up to 7 at.% burnup was predicted using the ALFUS code, which was developed for U-Pu-Zr ternary fuel performance analysis. As a result, it was evaluated that the fuel temperature is distributed between ∼410 deg. C and ∼645 deg. C at the end of the irradiation experiment. From the stress-strain analysis based on the preliminarily employed cladding irradiation properties and the FCMI stress distribution history, it was predicted that a cladding strain of not more than 0.9% would appear. (authors)

  13. Full scale impact testing for environmental and safety control of energy material shipping container systems

    International Nuclear Information System (INIS)

    Seagren, R.D.

    1978-01-01

    Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety

  14. Evaluation of an air drilling cuttings containment system

    Energy Technology Data Exchange (ETDEWEB)

    Westmoreland, J.

    1994-04-01

    Drilling at hazardous waste sites for environmental remediation or monitoring requires containment of all drilling fluids and cuttings to protect personnel and the environment. At many sites, air drilling techniques have advantages over other drilling methods, requiring effective filtering and containment of the return air/cuttings stream. A study of. current containment methods indicated improvements could be made in the filtering of radionuclides and volatile organic compounds, and in equipment like alarms, instrumentation or pressure safety features. Sandia National Laboratories, Dept. 61 11 Environmental Drilling Projects Group, initiated this work to address these concerns. A look at the industry showed that asbestos abatement equipment could be adapted for containment and filtration of air drilling returns. An industry manufacturer was selected to build a prototype machine. The machine was leased and put through a six-month testing and evaluation period at Sandia National Laboratories. Various materials were vacuumed and filtered with the machine during this time. In addition, it was used in an actual air drive drilling operation. Results of these tests indicate that the vacuum/filter unit will meet or exceed our drilling requirements. This vacuum/filter unit could be employed at a hazardous waste site or any site where drilling operations require cuttings and air containment.

  15. 40 CFR 281.37 - Financial responsibility for UST systems containing petroleum.

    Science.gov (United States)

    2010-07-01

    ... systems containing petroleum. 281.37 Section 281.37 Protection of Environment ENVIRONMENTAL PROTECTION... for No-Less-Stringent § 281.37 Financial responsibility for UST systems containing petroleum. (a) In... UST systems containing petroleum, the state requirements for financial responsibility for petroleum...

  16. Materials performance in off-gas systems containing iodine

    International Nuclear Information System (INIS)

    Beavers, J.A.; Berry, W.E.; Griess, J.C.

    1981-11-01

    During the reprocessing of spent reactor fuel elements, iodine is released to gas streams from which it is ultimately removed by conversion to nonvolatile iodic acid. Under some conditions iodine can produce severe corrosion in off-gas lines; in this study these conditions were established. Iron- and nickel-based alloys containing more than 6% molybdenum, such as Hastelloy G (7%), Inconel 625 (9%), and Hastelloy C-276 (16%), as well as titanium and zirconium, remained free of attack under all conditions tested. When the other materials, notably the austenitic stainless steels, were exposed to gas streams containing even only low concentrations of iodine and water vapors at 25 and 40 0 C, a highly corrosive, brownish-green liquid formed on their surfaces. In the complete absence of water vapor, the iodine-containing liquid did not form and all materials remained unaffected. The liquid that formed had a low pH (usually 2 inhibited attack

  17. Experiment Management System for the SND Detector

    Science.gov (United States)

    Pugachev, K.

    2017-10-01

    We present a new experiment management system for the SND detector at the VEPP-2000 collider (Novosibirsk). An important part to report about is access to experimental databases (configuration, conditions and metadata). The system is designed in client-server architecture. User interaction comes true using web-interface. The server side includes several logical layers: user interface templates; template variables description and initialization; implementation details. The templates are meant to involve as less IT knowledge as possible. Experiment configuration, conditions and metadata are stored in a database. To implement the server side Node.js, a modern JavaScript framework, has been chosen. A new template engine having an interesting feature is designed. A part of the system is put into production. It includes templates dealing with showing and editing first level trigger configuration and equipment configuration and also showing experiment metadata and experiment conditions data index.

  18. The PACE-1450 experiment - Crack and leakage behavior of a pre-stressed concrete containment wall considering ageing

    International Nuclear Information System (INIS)

    Hermann, N.; Mueller, H.S.; Niklasch, C.; Michel-Ponnelle, S.; Bento, C.; Masson, B.

    2015-01-01

    As an intermediate sized experiment the PACE-1450 experiment aims to investigate the behavior of a curved specimen (length: 3.5 m, width: 1.8 m, height: 1.2 m) which is representative for a 1450 MWe nuclear power plant containment under accidental loading conditions. One focus of this experimental test campaign is the consideration of the ageing of the structure which among other effects leads to a pre-stressing loss. The crack behavior of the realistically reinforced specimen is of as much interest as it is the leakage behavior when an inner pressure occurs within the containment. The reinforcement layout of the specimen is very similar to the original geometry and consists mainly of reinforcement meshes of bars near the inner and outer surface and four pre-stressing cables in the circumferential direction. During the tests the specimen is loaded by pressure which simulates the internal accidental containment pressure of up to 6 bars (absolute pressure). The resulting ring tensile stress in the cylindrical part of the containment is externally applied by hydraulic jacks. An initial pre-stressing of the specimen of 12 MPa is realized in such a way that decreasing the pre-stressing force for the purpose of simulating the ageing of the structure is possible. The facility allows for the cracking of the pre-stressed specimen and for leakage measurements at different controlled crack widths. The specimen is equipped with embedded optical fiber strain and temperature sensors and a sound detection system to record the initiation of cracks. The paper explains the test set-up and presents results of the ongoing test series regarding the cracking and leakage behavior of the specimen

  19. Analytical studies related to Indian PHWR containment system performance

    International Nuclear Information System (INIS)

    Haware, S.K.; Markandeya, S.G.; Ghosh, A.K.; Kushwaha, H.S.; Venkat Raj, V.

    1998-01-01

    Build-up of pressure in a multi-compartment containment after a postulated accident, the growth, transportation and removal of aerosols in the containment are complex processes of vital importance in deciding the source term. The release of hydrogen and its combustion increases the overpressure. In order to analyze these complex processes and to enable proper estimation of the source term, well tested analytical tools are necessary. This paper gives a detailed account of the analytical tools developed/adapted for PSA level 2 studies. (author)

  20. Analysis of railcar-shipping container system response to impact conditions

    International Nuclear Information System (INIS)

    Bartholomew, R.J.; Butler, T.A.

    1980-01-01

    An existing mathematical model for simulating railcar-container system response to coupling impacts was revised to simulate configurations that were tested in full-scale experiments. The structural model is represented with the lumped-parameter technique. The resulting equations are linear except for those for the coupler forces experienced during the impact. Results from the mathematical model are compared with load and acceleration data obtained during the full-scale tests. The model predicts actual response accurately enough to make it useful as a design and safety analysis tool

  1. Lipid containing nanodrug delivery system for the treatment of Tuberculosis

    CSIR Research Space (South Africa)

    Lemmer, Yolandy

    2010-09-01

    Full Text Available of the antibiotics in the cells, hence reducing the dose frequency and simultaneously improve patient compliance. The cell wall envelope of Mycobacterium tuberculosis (M.tb) contains unique high molecular weight lipids. Of these, the most abundant are mycolic acids...

  2. Star camera aspect system suitable for use in balloon experiments

    International Nuclear Information System (INIS)

    Hunter, S.D.; Baker, R.G.

    1985-01-01

    A balloon-borne experiment containing a star camera aspect system was designed, built, and flown. This system was designed to provide offset corrections to the magnetometer and inclinometer readings used to control an azimuth and elevation pointed experiment. The camera is controlled by a microprocessor, including commendable exposure and noise rejection threshold, as well as formatting the data for telemetry to the ground. As a background program, the microprocessor runs the aspect program to analyze a fraction of the pictures taken so that aspect information and offset corrections are available to the experiment in near real time. The analysis consists of pattern recognition of the star field with a star catalog in ROM memory and a least squares calculation. The performance of this system in ground based tests is described. It is part of the NASA/GSFC High Energy Gamma-Ray Balloon Instrument (2)

  3. Determining the effect of turbulent shear on containment aerosol dynamics using microgravity experiments

    International Nuclear Information System (INIS)

    Scott, C.K.; Abdelbaky, M.

    1997-01-01

    Determining the characteristics of large aerosol aggregates 'clusters' under turbulent conditions is fundamental for predicting the behaviour of radioactive aerosols inside the reactor containment following a severe accident. Studying such rapidly settling clusters is extremely difficult in ground-based experiments due to the effect of the earth's gravity. In this study, the microgravity environment is exploited to investigate the effect of turbulent shear on the aggregation and breakage of clusters by examining their structure and measuring their strength parameters while suspended under weightlessness conditions. A parametric model is introduced to correlate the experimental results over into nuclear aerosol models. It was demonstrated that the cluster parameters depend mainly on the turbulent field intensity as well as initial powder conditions. (author)

  4. Investigations on the gas distribution phenomena inside the containment system of LWRs

    International Nuclear Information System (INIS)

    Manfredini, A.; Oriolo, F.; Villotti, A.

    1994-01-01

    The importance of mixing and distribution phenomena of hydrogen gas in the reactor safety is emphasised in the advanced reactor concepts, that heavily rely upon the passive cooling systems during a typical severe accident sequence. An advanced methodology for evaluating the temporal and spatial distribution of non condensable gases, including the simulation of buoyancy-driven flows and the effects of the various ESFs activation, in a multi-compartment containment system of a LWR is reviewed. The methodology employs an analogy technique with electrical networks to determine the convection flows among the containment compartments and evaluates, inside a single node, the profile of the vertical concentrations of steam and non condensable gases. The application of the proposed models to simulate the gas distribution phenomena occurring in the HDR E11.2, in the FIPLOC-F2 and in the NUPEC M-7-1 tests demonstrates the importance of these models providing information about local details and spatial distribution. The main results from the post-test analysis performed to simulate the thermal-hydraulic responses of the above mentioned experiments are presented and demonstrate the improvements and the reduction of the error band with respect to the experimental data. This methodology allows to perform a realistic prediction of severe accident sequence inside the containment system of the actual and advanced passive generation of LWRs. (author). 14 refs., 11 figs

  5. Faulted systems recovery experience. Final report, May 1992

    International Nuclear Information System (INIS)

    1992-05-01

    This report addresses the recovery (i.e., return to service from a faulted, or otherwise unavailable, condition) of important nuclear power plant front-line and support systems and equipment. It contains information based on operating experience relative to the times to recover from a variety of plant events. It also indicates the nature of the operator actions involved. This information is intended to provide useful insights to utilities who are undertaking Individual Plant Examinations (IPEs) per Generic Letter 88-20 of the Nuclear Regulatory Commission. The report provides a database of recovery experience primarily derived from Licensee Event Reports (LERs). The database contains recovery duration information for 205 demand events and 98 nondemand events. In particular, it contains recovery durations for 42 pump related and 143 valve related events that are representative of demand conditions. Experience shows that, overall, about one-half of all pumps and valves are recovered in 30 minutes or less. Specific recovery experience is dependent on the equipment type, the plant system involved, and the failure mode encountered. (author)

  6. Dynamic Stability Experiment of Maglev Systems,

    Science.gov (United States)

    1995-04-01

    This report summarizes the research performed on maglev vehicle dynamic stability at Argonne National Laboratory during the past few years. It also... maglev system, it is important to consider this phenomenon in the development of all maglev systems. This report presents dynamic stability experiments...on maglev systems and compares their numerical simulation with predictions calculated by a nonlinear dynamic computer code. Instabilities of an

  7. Impact system for ultrafast synchrotron experiments

    International Nuclear Information System (INIS)

    Jensen, B. J.; Owens, C. T.; Ramos, K. J.; Yeager, J. D.; Saavedra, R. A.; Luo, S. N.; Hooks, D. E.; Iverson, A. J.; Fezzaa, K.

    2013-01-01

    The impact system for ultrafast synchrotron experiments, or IMPULSE, is a 12.6-mm bore light-gas gun (<1 km/s projectile velocity) designed specifically for performing dynamic compression experiments using the advanced imaging and X-ray diffraction methods available at synchrotron sources. The gun system, capable of reaching projectile velocities up to 1 km/s, was designed to be portable for quick insertion/removal in the experimental hutch at Sector 32 ID-B of the Advanced Photon Source (Argonne, IL) while allowing the target chamber to rotate for sample alignment with the beam. A key challenge in using the gun system to acquire dynamic data on the nanosecond time scale was synchronization (or bracketing) of the impact event with the incident X-ray pulses (80 ps width). A description of the basic gun system used in previous work is provided along with details of an improved launch initiation system designed to significantly reduce the total system time from launch initiation to impact. Experiments were performed to directly measure the gun system time and to determine the gun performance curve for projectile velocities ranging from 0.3 to 0.9 km/s. All results show an average system time of 21.6 ± 4.5 ms, making it possible to better synchronize the gun system and detectors to the X-ray beam.

  8. Delayed phenomena analysis from French PWR containment instrumentation system

    International Nuclear Information System (INIS)

    Costaz, J.L.

    1987-01-01

    The analysis of the large amount of measurements which has been now gathered by EDF on its twenty two PWR 900 MW shows that the behaviour of concrete under creep and shrinkage effects is in good agreement with the values given as correct estimates by french regulations and taken into account for the design of nuclear prestressed structures. None of the containment buildings studied here showed significant differences with the regulations theoretical values and consequently all the measurements remain in the field of the allowable strain variations used for design. On the other hand, if the instant loading elastic modulus is clearly determined for each containment, and its effect on theoretical creep taken into account, it was not possible up till now to extract from measurements some particular effects such as type of concrete and agregates or climatic effects. (orig.)

  9. Depressurization-filtration system of the containment of French PWR's

    International Nuclear Information System (INIS)

    L'homme, A.; Schektman, N.

    1987-01-01

    In the hypothetical event of a core meltdown occurring in a pressurized water reactor, and in order to preserve the integrity of the containment threatened by a build-up in pressure, EDF has developed, with the CEA, a decompression device which filters the containment internal atmosphere by using an unused containment penetration, and a sand-box, as filtering mechanism. This device and its procedure for utilization, constitute the U5 procedure. Check-tests on a semi-industrial scale have been carried out at the Nuclear Research Centre at Cadarache, by using columns of sand 80 cm high, according to following varying criteria: the granulometry of the sand, that of the aerosols, the flow-through speed, and the percentage steam content of the fluid to be filtered. The filtering material chosen is sand of a median diameter of 0.6 mm. (log normal distribution). The purification factor is above 10. The device tested meets the chosen targets, and is applied today to French units on condition to simple modifications concerning specific aspects of different series. The first is expected to be put into service during 1987

  10. Validation of computer code TRAFIC used for estimation of charcoal heatup in containment ventilation systems

    International Nuclear Information System (INIS)

    Yadav, D.H.; Datta, D.; Malhotra, P.K.; Ghadge, S.G.; Bajaj, S.S.

    2005-01-01

    Full text of publication follows: Standard Indian PHWRs are provided with a Primary Containment Filtration and Pump-Back System (PCFPB) incorporating charcoal filters in the ventilation circuit to remove radioactive iodine that may be released from reactor core into the containment during LOCA+ECCS failure which is a Design Basis Accident for containment of radioactive release. This system is provided with two identical air circulation loops, each having 2 full capacity fans (1 operating and 1 standby) for a bank of four combined charcoal and High Efficiency Particulate Activity (HEPA) filters, in addition to other filters. While the filtration circuit is designed to operate under forced flow conditions, it is of interest to understand the performance of the charcoal filters, in the event of failure of the fans after operating for some time, i.e., when radio-iodine inventory is at its peak value. It is of interest to check whether the buoyancy driven natural circulation occurring in the filtration circuit is sufficient enough to keep the temperature in the charcoal under safe limits. A computer code TRAFIC (Transient Analysis of Filters in Containment) was developed using conservative one dimensional model to analyze the system. Suitable parametric studies were carried out to understand the problem and to identify the safety of existing system. TRAFIC Code has two important components. The first one estimates the heat generation in charcoal filter based on 'Source Term'; while the other one performs thermal-hydraulic computations. In an attempt validate the Code, experimental studies have been carried out. For this purpose, an experimental set up comprising of scaled down model of filtration circuit with heating coils embedded in charcoal for simulating the heating effect due to radio iodine has been constructed. The present work of validation consists of utilizing the results obtained from experiments conducted for different heat loads, elevations and adsorbent

  11. Microbial analysis of the buffer/container experiment at AECL's Underground Research Laboratory

    International Nuclear Information System (INIS)

    Stroes-Gascoyne, S.; Hamon, C.J.; Haveman, S.A.; Delaney, T.L.

    1996-05-01

    The Buffer/Container experiment was carried out for 2.5 years to examine the in-situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived to conditions and to determine which groups of microorganisms would be dominant in such a simulated vault environment. Microbial analyses were initiated within 24 hour of sampling for all types of samples taken. The culture results showed an almost universal disappearance of viable microorganisms in the samples taken from near the heater surface. The microbial activity measurements confirmed the lack of viable organisms with very weak or no activity measured in most of these samples. Generally, aerobic heterotrophic culture conditions gave the highest mean colony-forming units (CFU) values at both 25 and 50 C. Under anaerobic conditions, and especially at 50 C, lower mean CFU values were obtained. In all samples analyzed, numbers of sulfate reducing bacteria were less than 1000 CFU/g dry material. Methanogens were either not present or were found in very low numbers. Anaerobic sulfur oxidizing bacteria were found in higher numbers in most sample types with sufficient moisture. The statistical evaluation of the culture data demonstrated clearly that the water content was the variable limiting the viability of the bacteria present, and not the temperature. 68 refs, 35 figs, 37 tabs

  12. Reflooding experiments on a 49-rod cluster containing a long 90% blockage

    International Nuclear Information System (INIS)

    Pearson, K.G.; Cooper, C.A.; Jowitt, D.; Kinneir, J.H.

    1983-01-01

    A series of reflooding experiments was performed on a model fuel assembly, containing a very severe partial blockage, in the THETIS rig. The assembly comprised 49 full length, electrically heated fuel rod simulators and the blockage was created by attaching thin-walled, preformed swellings to a group of 16 rods. Results are presented for single phase and forced reflooding experiments. The most important findings relate to the improvements in heat transfer created by spacer grids and the nature of the heat transfer processes within the blockage. Spacer grids are shown to improve heat transfer by increasing turbulence and also, when wet, by cooling the steam flowing through them. Liquid penetration evidently deteriorates as the rewetting front approaches the blockage, allowing the steam through the blockage to superheat strongly and giving rise to a late peak in cladding temperature. At low reflooding rates there is a temperature penalty associated with the blockage which becomes increasingly larger as the reflooding rate is reduced. The adequacy of cooling in this very severe blockage becomes questionable when the reflooding rate falls to about 2cm/s. (U.K.)

  13. Microbial analysis of the buffer/container experiment at AECL`s Underground Research Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Stroes-Gascoyne, S; Hamon, C J; Haveman, S A; Delaney, T L [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs; Pedersen, K; Ekendahl, S; Jahromi, N; Arlinger, J; Hallbeck, L [Univ. of Goeteborg, (Sweden). Dept. of General and Marine Microbiology; Daumas, S; Dekeyser, K [Guiges Recherche Appliquee en Microbiologie, Aix-en-Provence, (France)

    1996-05-01

    The Buffer/Container experiment was carried out for 2.5 years to examine the in-situ performance of compacted buffer material in a single emplacement borehole under vault-relevant conditions. During decommissioning of this experiment, numerous samples were taken for microbial analysis to determine if the naturally present microbial population in buffer material survived to conditions and to determine which groups of microorganisms would be dominant in such a simulated vault environment. Microbial analyses were initiated within 24 hour of sampling for all types of samples taken. The culture results showed an almost universal disappearance of viable microorganisms in the samples taken from near the heater surface. The microbial activity measurements confirmed the lack of viable organisms with very weak or no activity measured in most of these samples. Generally, aerobic heterotrophic culture conditions gave the highest mean colony-forming units (CFU) values at both 25 and 50 C. Under anaerobic conditions, and especially at 50 C, lower mean CFU values were obtained. In all samples analyzed, numbers of sulfate reducing bacteria were less than 1000 CFU/g dry material. Methanogens were either not present or were found in very low numbers. Anaerobic sulfur oxidizing bacteria were found in higher numbers in most sample types with sufficient moisture. The statistical evaluation of the culture data demonstrated clearly that the water content was the variable limiting the viability of the bacteria present, and not the temperature. 68 refs, 35 figs, 37 tabs.

  14. Magnet system studies for the Zeus experiment

    International Nuclear Information System (INIS)

    Baynham, D.E.; Coombs, R.C.; Uden, C.N.

    1985-11-01

    The ZEUS experiment will be mounted at the HERA accelerator complex currently under construction at DESY, Hamburg. A large volume of magnetic field will be required for charge selection of particles and track fitting. Two superconducting magnet systems which meet the parameters of the ZEUS Experiment are described; a small solenoid with good radiation transparency and a large aperture Helmholtz coil configuration. Basic design concepts and parameters are presented. (author)

  15. Smart container UWB sensor system for situational awareness of intrusion alarms

    Science.gov (United States)

    Romero, Carlos E.; Haugen, Peter C.; Zumstein, James M.; Leach, Jr., Richard R.; Vigars, Mark L.

    2013-06-11

    An in-container monitoring sensor system is based on an UWB radar intrusion detector positioned in a container and having a range gate set to the farthest wall of the container from the detector. Multipath reflections within the container make every point on or in the container appear to be at the range gate, allowing intrusion detection anywhere in the container. The system also includes other sensors to provide false alarm discrimination, and may include other sensors to monitor other parameters, e.g. radiation. The sensor system also includes a control subsystem for controlling system operation. Communications and information extraction capability may also be included. A method of detecting intrusion into a container uses UWB radar, and may also include false alarm discrimination. A secure container has an UWB based monitoring system

  16. Development of the interactive model between Component Cooling Water System and Containment Cooling System using GOTHIC

    International Nuclear Information System (INIS)

    Byun, Choong Sup; Song, Dong Soo; Jun, Hwang Yong

    2006-01-01

    In a design point of view, component cooling water (CCW) system is not full-interactively designed with its heat loads. Heat loads are calculated from the CCW design flow and temperature condition which is determined with conservatism. Then the CCW heat exchanger is sized by using total maximized heat loads from above calculation. This approach does not give the optimized performance results and the exact trends of CCW system and the loads during transient. Therefore a combined model for performance analysis of containment and the component cooling water(CCW) system is developed by using GOTHIC software code. The model is verified by using the design parameters of component cooling water heat exchanger and the heat loads during the recirculation mode of loss of coolant accident scenario. This model may be used for calculating the realistic containment response and CCW performance, and increasing the ultimate heat sink temperature limits

  17. Binding in some few-body systems containing antimatter

    International Nuclear Information System (INIS)

    Armour, E.A.G.

    2009-01-01

    It is well known that the system made up of a fixed proton and antiproton and an electron (or a positron) has no bound states if the internuclear distance R 0 . In this paper, I consider the more complicated system in which the electron and the positron are both present and investigate the possibility of obtaining a lower bound on the value of R for which the system has no bound states. I also investigate the implications of the existence of bound states of the simpler, one light particle system regarding bound states of the more complicated system. This article is based on the presentation by E. A. G. Armour at the Fifth Workshop on Critical Stability, Erice, Sicily. (author)

  18. Aqueous-salt system containing ytterbium nitrate and pyridine nitrate

    International Nuclear Information System (INIS)

    Zhuravlev, E.F.; Khisaeva, D.A.; Izmajlova, L.V.

    1983-01-01

    Cross-section method has been used to study solubility in ternary aqueous-salt system Yb(NO 3 ) 3 -C 5 H 5 NxHNO 3 -H 2 0 at 25 and 50 deg C. It is established that the system is characterized by chemical interaction. Congruently soluble compound of Yb(NO 3 ) 3 x2[C 5 H 5 NxHNO 3 ] composition is discovered in the system. Composition of the compound is confirmed by chemical analysis; its infrared spectra are studied. Interplanar distances are determined; derivatogram of the compound is given. The results of the works are compared with analogous investigations of another rare earth nitrates

  19. Containment system of contamination in irradiated materials handling laboratories

    International Nuclear Information System (INIS)

    Lobao, A.S.T.; Araujo, J.A. de; Camilo, R.L.

    1988-01-01

    A study to prevent radiactivity release in lab scale is presented. As a basis for the design all the limits established by the IAEA for ventilation systems were observed. An evaluation of the different parameters involved in the design have been made, resulting in the specification of the working areas, ducts and filtering systems in order to get the best conditions for the safe handling of irradiated materials. (author) [pt

  20. Pathways, Networks, and Systems: Theory and Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Joseph H. Nadeau; John D. Lambris

    2004-10-30

    The international conference provided a unique opportunity for theoreticians and experimenters to exchange ideas, strategies, problems, challenges, language and opportunities in both formal and informal settings. This dialog is an important step towards developing a deep and effective integration of theory and experiments in studies of systems biology in humans and model organisms.

  1. COCOSYS: Status of development and validation of the German containment code system

    International Nuclear Information System (INIS)

    Allelein, H.-J.; Arndt, S.; Klein-Hessling, W.; Schwarz, S.; Spengler, C.; Weber, G.

    2006-01-01

    simulation of the chemistry inside the core melt to calculate the release of gaseous components and fission products. The overall concept of the COCOSYS system has turned out to be suitable for parallel calculation of different processes and including further detailed models. First attempt for the connection with the CFD code CFX4.1 have been made. External codes like ATHLET for reactor circuit thermal hydraulics, LAVA for melt spreading, DET3D for denotative hydrogen combustion and the industrial CFD code CFX are connected with COCOSYS. COCOSYS is subject to an ongoing internal and external validation process. At present this validation process is mainly based on tests being performed in the German ThAI facility. Experiments to be performed in ThAI dealing with hydrogen combustion, recombiner behaviour and aerosol and iodine issues are currently offered to the community as an OECD project. Examples given for the successful validation are the participation in the OECD/NEA ISP-47 and the benchmark for the CCI-2 test in the frame of the OECD-MCCI project. E. g. COCOSYS has been used in licensing procedure performed for the installation of catalytic recombiners in German nuclear power plants. Variation of the boundary conditions have underlined the need of detailed nodalization of the containment and the need of comprehensive simulation of system components (like doors, ventilation systems, rupture discs), having an influence on the overall gas distribution and on local effect. In the future further improvements and model extensions like pyrolysis processes, direct containment heating sand the combined use with CFD models will be performed. (author)

  2. Benchmark of the HDR E11.2 containment hydrogen mixing experiment using the MAAP4 code

    International Nuclear Information System (INIS)

    Lee, Sung, Jin; Paik, Chan Y.; Henry, R.E.

    1997-01-01

    The MAAP4 code was benchmarked against the hydrogen mixing experiment in a full-size nuclear reactor containment. This particular experiment, designated as E11.2, simulated a small loss-of-coolant-accident steam blowdown into the containment followed by the release of a hydrogen-helium gas mixture. It also incorporated external spray cooling of the steel dome near the end of the transient. Specifically, the objective of this bench-mark was to demonstrate that MAAP4, using subnodal physics, can predict an observed gas stratification in the containment

  3. The achievement and assessment of safety in systems containing software

    International Nuclear Information System (INIS)

    Ball, A.; Dale, C.J.; Butterfield, M.H.

    1986-01-01

    In order to establish confidence in the safe operation of a reactor protection system, there is a need to establish, as far as it is possible, that: (i) the algorithms used are correct; (ii) the system is a correct implementation of the algorithms; and (iii) the hardware is sufficiently reliable. This paper concentrates principally on the second of these, as it applies to the software aspect of the more accurate and complex trip functions to be performed by modern reactor protection systems. In order to engineer safety into software, there is a need to use a development strategy which will stand a high chance of achieving a correct implementation of the trip algorithms. This paper describes three broad methodologies by which it is possible to enhance the integrity of software: fault avoidance, fault tolerance and fault removal. Fault avoidance is concerned with making the software as fault free as possible by appropriate choice of specification, design and implementation methods. A fault tolerant strategy may be advisable in many safety critical applications, in order to guard against residual faults present in the software of the installed system. Fault detection and removal techniques are used to remove as many faults as possible of those introduced during software development. The paper also discusses safety and reliability assessment as it applies to software, outlining the various approaches available. Finally, there is an outline of a research project underway in the UKAEA which is intended to assess methods for developing and testing safety and protection systems involving software. (author)

  4. A self contained Linux based data acquisition system for 2D detectors with delay line readout

    International Nuclear Information System (INIS)

    Beltran, D.; Toledo, J.; Klora, A.C.; Ramos-Lerate, I.; Martinez, J.C.

    2007-01-01

    This article describes a fast and self-contained data acquisition system for 2D gas-filled detectors with delay line readout. It allows the realization of time resolved experiments in the millisecond scale. The acquisition system comprises of an industrial PC running Linux, a commercial time-to-digital converter and an in-house developed histogramming PCI card. The PC provides a mass storage for images and a graphical user interface for system monitoring and control. The histogramming card builds images with a maximum count rate of 5 MHz limited by the time-to-digital converter. Histograms are transferred to the PC at 85 MB/s. This card also includes a time frame generator, a calibration channel unit and eight digital outputs for experiment control. The control software was developed for easy integration into a beamline, including scans. The system is fully operational at the Spanish beamline BM16 at the ESRF in France, the neutron beamlines Adam and Eva at the ILL in France, the Max Plank Institute in Stuttgart in Germany, the University of Copenhagen in Denmark and at the future ALBA synchrotron in Spain. Some representative collected images from synchrotron and neutron beamlines are presented

  5. HECLA experiments on interaction between metallic melt and hematite-containing concrete

    Energy Technology Data Exchange (ETDEWEB)

    Sevon, Tuomo, E-mail: tuomo.sevon@vtt.f [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland); Kinnunen, Tuomo; Virta, Jouko; Holmstroem, Stefan; Kekki, Tommi; Lindholm, Ilona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT, Espoo (Finland)

    2010-10-15

    In a hypothetical severe accident in a nuclear power plant, molten materials may come into contact with concrete, causing concrete ablation. In five HECLA experiments the interaction between metallic melt and concrete was investigated by pouring molten stainless steel at almost 1800 {sup o}C into cylindrical concrete crucibles. The tests were transient, i.e. no decay heat simulation was used. The main objective was to test the behavior of the FeSi concrete, containing hematite (Fe{sub 2}O{sub 3}) and siliceous aggregates. This special concrete type is used as a sacrificial layer in the Olkiluoto 3 EPR reactor pit, and very scarce experimental data is available about its behavior at high temperatures. It is concluded that no clear differences between the ablation of FeSi concrete and ordinary siliceous concrete were observed. The ablation depths were small, 25 mm at maximum. No dramatic effects, such as cracking of large pieces of concrete due to the thermal shock, took place. An important side result of the test series was gaining knowledge of the properties of the special concrete type. Chemical analyses were conducted and mechanical properties were measured.

  6. Correction of chromatic abberation in electrostatic lense systems containing quadrupoles

    International Nuclear Information System (INIS)

    Baranova, L.A.; Ul'yanova, N.S.; Yavor, S.Ya.

    1991-01-01

    Possibility of chromatic abberation correction in immersion systems consisting of axysimmetric and quadrupole lenses is shown. Concrete examples are presented. A number of new directions in science and technique, using ion beams are intensively developed presently. When using them accute necessity arises in chromatic abberation correction, while large-scale energy scattering is observed as a rule in such cases

  7. Tracer verification and monitoring of containment systems (II)

    International Nuclear Information System (INIS)

    Williams, C.V.; Dunn, S.D.; Lowry, W.E.

    1997-01-01

    A tracer verification and monitoring system, SEAtrace trademark, has been designed and field tested which uses gas tracers to evaluate, verify, and monitor the integrity of subsurface barriers. This is accomplished using an automatic, rugged, autonomous monitoring system combined with an inverse optimization code. A gaseous tracer is injected inside the barrier and an array of wells outside the barrier are monitored. When the tracer gas is detected, a global optimization code is used to calculate the leak parameters, including leak size, location, and when the leak began. The multipoint monitoring system operates in real-time, can be used to measure both the tracer gas and soil vapor contaminants, and is capable of unattended operation for long periods of time (months). The global optimization code searches multi-dimensional open-quotes spaceclose quotes to find the best fit for all of the input parameters. These parameters include tracer gas concentration histories from multiple monitoring points, medium properties, barrier location, and the source concentration. SEAtrace trademark does not attempt to model all of the nuances associated with multi-phase, multi-component flow, but rather, the inverse code uses a simplistic forward model which can provide results which are reasonably accurate. The system has calculated leak locations to within 0.5 meters and leak radii to within 0.12 meters

  8. V1334 Cyg: A Triple System Containing a Classical Cepheid

    Science.gov (United States)

    Evans, N. R.

    2000-05-01

    HR 8157 = ADS 14859 = HD 203156 = V1334 Cyg was recognized a hundred years ago to be a marginally resolved visual binary. Millis (1969, Lowell Obs Bull, 7, 113) discovered that the brightest star in the system is a low amplitude classical Cepheid with a pulsation period of 3.3 days. Early radial velocity observations by Abt and Levy (1970, PASP, 82, 334) differed from scattered radial velocity observations in the first half of the century implying that in addition to the long period system, the Cepheid is also a member of a short period binary. We have observed Cepheid V1334 Cyg A for nearly 30 years. From this radial velocity data we have derived an orbit with a period of 5 years. The orbit provides limits on the mass of the companion (V1334 Cyg C) of 3.1 to 4.4 solar masses. We have used an IUE high resolution spectrum to conclude that the hottest star in the system (V1334 Cyg B) which dominates the spectrum in the ultraviolet is the wide companion since the velocity is very near the systemic velocity. Financial support was supplied through a Natural Sciences and Engineering Research Council, Canada (NSERC) grant and HST Grant GO-07478.01-96A, and from the Chandra Science Center NASA Contract NAS8-39073.

  9. Data processing system for neutron experiments

    Energy Technology Data Exchange (ETDEWEB)

    Emoto, T; Yamamuro, N [Tokyo Inst. of Tech. (Japan). Research Lab. of Nuclear Reactor

    1979-03-01

    A data processing system for neutron experiments has been equipped at the Pelletron Laboratory of the Research Laboratory for Nuclear Reactors. The system comprises a Hewlett Packard 21 MX computer and a CAMAC standard. It can control two ADCs and some CAMAC modules. CAMAC control programs as well as data acquisition programs with high-level language can be readily developed. Terminals are well designed for man-machine interactions and program developments. To demonstrate the usefulness of the system, it was applied for the on-line data processing of neutron spectrum measurement.

  10. Source term aspects associated with future PWR containment systems

    International Nuclear Information System (INIS)

    Kuczera, B.; Kebler, G.; Ehrhardt, J.; Scholtyssek, W.

    1994-01-01

    The overall objective of reactor safety is to protect the population against dangerous releases of radioactive materials from nuclear power plants. In context with a reinforcement of the defense-in-depth strategy the common safety requirements on future nuclear power plants converge in the objective that these plants should be so safe that even in case of a severe accident there will be no need of off-site emergency actions such as an evacuation or resettlement of the population from the vicinity of a nuclear power plant. It is shown by the example of a future 1400 MWe pressurized water reactor (PWR) plant that this goal can be attained in principle by providing a double containment with the annulus vented via an appropriate emergency standby filter. Within the framework of severe accident consequence mitigation a set of parameters for accident conditions and emergency filter efficiencies is elaborated under which the German lower levels of intervention for evacuation are not attained. (author). 10 refs., 3 tabs., 5 figs

  11. Design of an experiment to measure fire exposure of packages aboard container cargo ships

    International Nuclear Information System (INIS)

    Koski, J.A.

    1998-01-01

    The test described in this paper is intended to measure the typical accident environment for a radioactive materials package in a fire abroad a container cargo ship. A stack of nice used standard cargo containers will be variously loaded with empty packages, simulated packages and combustible cargo and placed over a large hydrocarbon pool fire of one hour duration. Fire environments, both inside and outside the containers, typical of on-deck stowage will be measured as well as the potential for container-to-container fire spread. With the use of the inverse heat conduction calculations, the local heat transfer to the simulated packages can be estimated from thermocouple data. Data recorded will also provide information on fire durations in each container, fire intensity and container-to-container fire spread characteristics. (authors)

  12. Introduction to Test Facility for Iodine Retention in Filtered Containment Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jaehoon; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In many countries the implementation of FCVS's is under discussion to mitigate fission product release not only in the short-term but also in the long-term view. To verify the performance of FCVS, the large-scaled tests have been performed such as advanced containment experiments (ACE), the iodine and aerosol retention rate test facility (JAVA), etc. The elemental and organic iodides are the main gaseous iodine species in the containment atmosphere. For the iodine retention, experimental programs have confirmed the existence of gaseous organic iodine in some cases in higher concentrations than for gaseous molecular iodine (I{sub 2}). The Reaction of Methyl iodide (CH{sub 3}I) with surfaces and the removal by containment filters and scrubbers is less efficient in comparison to molecular iodine. In the recent years, an experimental and analytical work has been conducted at the Paul Scherrer Institute (PSI) to develop a process leading to a fast, comprehensive and reliable retention of volatile iodine species in aqueous solutions. New FCVS test facility to verify the performance of FCVS is designed and under construction. The iodine retention tests are planned with elemental iodine or with organic iodide loaded carrier gas consisting of pure non-condensable gas, pure steam and of typical mixtures of non-condensable gas/steam. This paper introduces the iodine generation and measurement system for the iodine retention test of FCVS. In severe accidents elemental and organic iodides are the main gaseous iodine species in the containment atmosphere. Release of the gaseous species in sufficient quantities from containment to environment generates a risk for public health. The filtered containment venting systems (FCVS) can considerably reduce the leakage of radioactive materials to the environment. New integral test facility is prepared to verify a performance of the FCVS. The test facility consists of a test vessel, thermal-hydraulic, and aerosol/iodine generation and

  13. Surveillance and Monitoring Program Full-Scale Experiments to Evaluate the Potential for Corrosion in 3013 Containers

    Energy Technology Data Exchange (ETDEWEB)

    Narlesky, Joshua Edward [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Berg, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Duque, Juan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Harradine, David Martin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hill, Dallas Dwight [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kaczar, Gregory Michael [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lillard, R. Scott [Univ. of Akron, OH (United States); Lopez, Annabelle Sarita [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, Max Alfonso [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Peppers, Larry G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rios, Daniel [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Edward L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stroud, Mary Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Trujillo, Leonardo Alberto [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Veirs, Douglas Kirk [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wilson, Kennard Virden Jr. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Worl, Laura Ann [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-11-21

    A set of six long-term, full-scale experiments were initiated to determine the type and extent of corrosion that occurs in 3013 containers packaged with chloride-bearing plutonium oxide materials. The materials were exposed to a high relative humidity environment representative of actual packaging conditions for the materials in storage. The materials were sealed in instrumented, inner 3013 containers with corrosion specimens designed to test the corrosiveness of the environment inside the containers under various conditions. This report focuses on initial loading conditions that are used to establish a baseline to show how the conditions change throughout the storage lifetime of the containers.

  14. A system for designing and simulating particle physics experiments

    International Nuclear Information System (INIS)

    Zelazny, R.; Strzalkowski, P.

    1987-01-01

    In view of the rapid development of experimental facilities and their costs, the systematic design and preparation of particle physics experiments have become crucial. A software system is proposed as an aid for the experimental designer, mainly for experimental geometry analysis and experimental simulation. The following model is adopted: the description of an experiment is formulated in a language (here called XL) and put by its processor in a data base. The language is based on the entity-relationship-attribute approach. The information contained in the data base can be reported and analysed by an analyser (called XA) and modifications can be made at any time. In particular, the Monte Carlo methods can be used in experiment simulation for both physical phenomena in experimental set-up and detection analysis. The general idea of the system is based on the design concept of ISDOS project information systems. The characteristics of the simulation module are similar to those of the CERN Geant system, but some extensions are proposed. The system could be treated as a component of greater, integrated software environment for the design of particle physics experiments, their monitoring and data processing. (orig.)

  15. Design of Drug Delivery Systems Containing Artemisinin and Its Derivatives

    Directory of Open Access Journals (Sweden)

    Blessing Atim Aderibigbe

    2017-02-01

    Full Text Available Artemisinin and its derivatives have been reported to be experimentally effective for the treatment of highly aggressive cancers without developing drug resistance, they are useful for the treatment of malaria, other protozoal infections and they exhibit antiviral activity. However, they are limited pharmacologically by their poor bioavailability, short half-life in vivo, poor water solubility and long term usage results in toxicity. They are also expensive for the treatment of malaria when compared to other antimalarials. In order to enhance their therapeutic efficacy, they are incorporated onto different drug delivery systems, thus yielding improved biological outcomes. This review article is focused on the currently synthesized derivatives of artemisinin and different delivery systems used for the incorporation of artemisinin and its derivatives.

  16. Nuclear containment systems and in-service inspection status of Korea nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jihong, Park; Jaekeun, Hong; Banuk, Park [Korea Institute of Machinery and Materials, Dept. of Authorized Test and Evaluation, Kyungnam (Korea, Republic of)

    2007-07-01

    20 unit nuclear power plants in Korea have been operated and maintained since the first unit started in commercial service in 1978. Most recently 4 units were under construction and several units were planned to be constructed. by industries. 4 types of nuclear containment systems have been constructed until now: first, metal containments, then pre-stressed concrete containments with grouted tendon systems, followed by pre-stressed concrete containments with un-grouted tendon systems, and Korea standard nuclear containments. All the nuclear containments should be inspected periodically. Therefore for periodic in-service inspection, several appropriate technical requirements should be applied differently depending on the specific nuclear containment types. With the changes of times, nuclear containment systems have undergone a remarkable change, and finally nuclear containment system of Korea standard nuclear power plant was settled down, and as a matter of course it dominates the trend of present and future nuclear containment systems. Overall in-service inspection results of most Korea nuclear containments have not showed any serious evidence of degradation.

  17. Process and closure system for a radioactive waste container

    International Nuclear Information System (INIS)

    Meyer, Andre.

    1974-01-01

    The closure process described is for a cylindrical radioactive waste drum. It makes use of a closure system for the drum comprising two lids separated by a twin flange seal. It consists essentially in placing a double flange 'O' ring inside the upper lip of the drum, and after filling has been completed, fitting the first lid on the twin flange 'O' ring and pushing down this lid whilst squashing the upper flange and then putting on the second lid in the usual prescribed manner. A description is also given of the drum sealing apparatus [fr

  18. System and process for aluminization of metal-containing substrates

    Energy Technology Data Exchange (ETDEWEB)

    Chou, Yeong-Shyung; Stevenson, Jeffry W.

    2017-12-12

    A system and method are detailed for aluminizing surfaces of metallic substrates, parts, and components with a protective alumina layer in-situ. Aluminum (Al) foil sandwiched between the metallic components and a refractory material when heated in an oxidizing gas under a compression load at a selected temperature forms the protective alumina coating on the surface of the metallic components. The alumina coating minimizes evaporation of volatile metals from the metallic substrates, parts, and components in assembled devices that can degrade performance during operation at high temperature.

  19. System and process for aluminization of metal-containing substrates

    Science.gov (United States)

    Chou, Yeong-Shyung; Stevenson, Jeffry W

    2015-11-03

    A system and method are detailed for aluminizing surfaces of metallic substrates, parts, and components with a protective alumina layer in-situ. Aluminum (Al) foil sandwiched between the metallic components and a refractory material when heated in an oxidizing gas under a compression load at a selected temperature forms the protective alumina coating on the surface of the metallic components. The alumina coating minimizes evaporation of volatile metals from the metallic substrates, parts, and components in assembled devices during operation at high temperature that can degrade performance.

  20. Role of unsaturated soil in a waste containment system

    Energy Technology Data Exchange (ETDEWEB)

    Lim, P.C.; Tay, J.H. [Nanyang Technological Univ. (Singapore)

    1996-12-31

    The role of the unsaturated properties of sand as a drainage layer in a composite liner system for landfills is investigated. The effect of the unsaturated properties of coarse-grained soil on contaminant migration was evaluated by means of a series of simulations using a one-dimensional model of a two- and a three-layer soil liner system for advection and diffusion, respectively. The results showed that under seepage conditions, the effect of an unsaturated sand layer on the advancement of the concentration front was quite insignificant. The arrival time of the C/C{sub o} = 0.5 concentration front increased from 651 days for the case with no sand layer to approximately 951 days for the case with a 1.0-m sand layer. A steady-state flow condition was ultimately established in the sand, and this fact suggests that the capillary action might not be effective. For diffusion, the arrival time of the concentration front increased nonlinearly with a decrease in the degree of saturation and linearly with increasing depths of the sand layer. At a residual degree of saturation, the arrival times of the C/C{sub o} = 0.01 and 0.5 concentration front at the base of the 1-m sand layer were 26.9 and 877.4 years as compared to 1.52 and 2.62 years by advection, respectively. 17 refs., 11 figs.

  1. Methodology to analysis of aging processes of containment spray system

    International Nuclear Information System (INIS)

    Borges, D. da Silva; Lava, D.D.; Moreira, M. de L.; Ferreira Guimarães, A.C.; Fernandes da Silva, L.

    2015-01-01

    This paper presents a contribution to the study of aging process of components in commercial plants of Pressurized Water Reactors (PWRs). The motivation for write this work emerged from the current perspective nuclear. Numerous nuclear power plants worldwide have an advanced operating time. This problem requires a process to ensure the confiability of the operative systems of these plants, because of this, it is necessary a methodologies capable of estimate the failure probability of the components and systems. In addition to the safety factors involved, such methodologies can to be used to search ways to ensure the extension of the life cycle of nuclear plants, which inevitably will pass by the decommissioning process after the operating time of 40 years. This process negatively affects the power generation, besides demanding an enormous investment for such. Thus, this paper aims to present modeling techniques and sensitivity analysis, which together can generate an estimate of how components, which are more sensitive to the aging process, will behave during the normal operation cycle of a nuclear power plant. (authors)

  2. (Liquid + liquid) phase behavior for systems containing (aromatic + TBA + methylcyclohexane)

    International Nuclear Information System (INIS)

    Ghanadzadeh, H.; Ghanadzadeh, A.

    2004-01-01

    The determination region of solubility of TBA (tert-butanol) with representative compounds of the gasoline was investigated experimentally at temperature of 298.2 K. Type 1 (liquid + liquid) phase diagrams were obtained for (methylcyclohexane + TBA + aromatic compounds). These results were correlated simultaneously by the UNIQUAC model. The values of the interaction parameters between each pair of components in the systems were obtained for the UNIQUAC model using the experimental result. The root mean square deviation (RMSD) between the observed and calculated mole percents was 1.88 for (methylcyclohexane + TBA + benzene), 2.45 for (methylcyclohexane + TBA + toluene) and 2.86 for (methylcyclohexane + TBA + ethylbenzene). The mutual solubility of methylcyclohexane and aromatic compounds (e.g., benzene toluene and ethylbenzene (BTE)) was also investigated by the addition of TBA at temperature of 298.2 K

  3. Thermodynamic Modeling of Natural Gas Systems Containing Water

    DEFF Research Database (Denmark)

    Karakatsani, Eirini K.; Kontogeorgis, Georgios M.

    2013-01-01

    As the need for dew point specifications remains very urgent in the natural gas industry, the development of accurate thermodynamic models, which will match experimental data and will allow reliable extrapolations, is needed. Accurate predictions of the gas phase water content in equilibrium...... with a heavy phase were previously obtained using cubic plus association (CPA) coupled with a solid phase model in the case of hydrates, for the binary systems of water–methane and water–nitrogen and a few natural gas mixtures. In this work, CPA is being validated against new experimental data, both water...... content and phase equilibrium data, and solid model parameters are being estimated for four natural gas main components (methane, ethane, propane, and carbon dioxide). Different tests for the solid model parameters are reported, including vapor-hydrate-equilibria (VHE) and liquid-hydrate-equilibria (LHE...

  4. Influence of local capillary trapping on containment system effectiveness

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Steven [University Of Texas At Austin, Austin, TX (United States). Center for Petroleum and Geosystems Engineering

    2014-03-31

    Immobilization of CO2 injected into deep subsurface storage reservoirs is a critical component of risk assessment for geologic CO2 storage (GCS). Local capillary trapping (LCT) is a recently established mode of immobilization that arises when CO2 migrates due to buoyancy through heterogeneous storage reservoirs. This project sought to assess the amount and extent of LCT expected in storage formations under a range of injection conditions, and to confirm the persistence of LCT if the seal overlying the reservoir were to lose its integrity. Numerical simulation using commercial reservoir simulation software was conducted to assess the influence of injection. Laboratory experiments, modeling and numerical simulation were conducted to assess the effect of compromised seal integrity. Bench-scale (0.6 m by 0.6 m by 0.03 m) experiments with surrogate fluids provided the first empirical confirmation of the key concepts underlying LCT: accumulation of buoyant nonwetting phase at above residual saturations beneath capillary barriers in a variety of structures, which remains immobile under normal capillary pressure gradients. Immobilization of above-residual saturations is a critical distinction between LCT and the more familiar “residual saturation trapping.” To estimate the possible extent of LCT in a storage reservoir an algorithm was developed to identify all potential local traps, given the spatial distribution of capillary entry pressure in the reservoir. The algorithm assumes that the driving force for CO2 migration can be represented as a single value of “critical capillary entry pressure” Pc,entrycrit, such that cells with capillary entry pressure greater/less than Pc,entrycrit act as barriers/potential traps during CO2 migration. At intermediate values of Pc,entrycrit, the barrier regions become more laterally extensive in the reservoir

  5. Analysis of the exoplanet containing system Kepler-13

    Science.gov (United States)

    Budding, E.; Püsküllü, Ç.; Rhodes, M. D.

    2018-03-01

    We have applied the close binary system analysis program WinFitter, with its physically detailed fitting function, to an intensive study of the complex multiple system Kepler-13 using photometry data from all 13 short cadence quarters downloaded from the NASA Exoplanet Archive (NEA) (http://exoplanetarchive.ipac.caltech.edu). The data-point error of our normalized, phase-sequenced and binned (380 points per bin: 0.00025 phase interval) flux values, at 14 ppm, allows the model's specification for the mean reference flux level of the system to a precision better than 1 ppm. Our photometrically derived values for the mass and radius of KOI13.01 are 6.8±0.6 MJ and 1.44±0.04 RJ. The star has a radius of 1.67±0.05 R_{⊙}. Our modelling sets the mean of the orbital inclination i at 94.35±0.14°, with the star's mean precession angle φp—49.1±5.0° and obliquity θo 67.9 ± 3.0°, though there are known ambiguities about the sense in which such angles are measured. Our findings did not confirm secular variation in the transit modelling parameters greater than their full correlated errors, as argued by previous authors, when each quarter's data was best-fitted with a determinable parameter set without prejudice. However, if we accept that most of the parameters remain the same for each transit, then we could confirm a small but steady diminution in the cosine of the orbital inclination over the 17 quarter timespan. This is accompanied by a slight increase of the star's precession angle (less negative), but with no significant change in the obliquity of its spin axis. There are suggestions of a history of strong dynamical interaction with a highly distorted planet rotating in a 3:2 resonance with its revolution, together with a tidal lag of ˜30 deg. The mean precessional period is derived to be about 1000 y, but at the present time the motion of the star's rotation axis appears to be supporting the gravitational torque, rather than providing the balance against it

  6. System of and method for transparent management of data objects in containers across distributed heterogenous resources

    Science.gov (United States)

    Moore, Reagan W.; Rajasekar, Arcot; Wan, Michael Y.

    2007-09-11

    A system of and method for maintaining data objects in containers across a network of distributed heterogeneous resources in a manner which is transparent to a client. A client request pertaining to containers is resolved by querying meta data for the container, processing the request through one or more copies of the container maintained on the system, updating the meta data for the container to reflect any changes made to the container as a result processing the re quest, and, if a copy of the container has changed, changing the status of the copy to indicate dirty status or synchronizing the copy to one or more other copies that may be present on the system.

  7. The Trigger System of the CMS Experiment

    OpenAIRE

    Felcini, Marta

    2008-01-01

    We give an overview of the main features of the CMS trigger and data acquisition (DAQ) system. Then, we illustrate the strategies and trigger configurations (trigger tables) developed for the detector calibration and physics program of the CMS experiment, at start-up of LHC operations, as well as their possible evolution with increasing luminosity. Finally, we discuss the expected CPU time performance of the trigger algorithms and the CPU requirements for the event filter farm at start-up.

  8. Experiences from the Roadrunner petascale hybrid systems

    Energy Technology Data Exchange (ETDEWEB)

    Kerbyson, Darren J [Los Alamos National Laboratory; Pakin, Scott [Los Alamos National Laboratory; Lang, Mike [Los Alamos National Laboratory; Sancho Pitarch, Jose C [Los Alamos National Laboratory; Davis, Kei [Los Alamos National Laboratory; Barker, Kevin J [Los Alamos National Laboratory; Peraza, Josh [Los Alamos National Laboratory

    2010-01-01

    The combination of flexible microprocessors (AMD Opterons) with high-performing accelerators (IBM PowerXCell 8i) resulted in the extremely powerful Roadrunner system. Many challenges in both hardware and software were overcome to achieve its goals. In this talk we detail some of the experiences in achieving performance on the Roadrunner system. In particular we examine several implementations of the kernel application, Sweep3D, using a work-queue approach, a more portable Thread-building-blocks approach, and an MPI on the accelerator approach.

  9. Thermal Hydraulic Analysis on Containment Filtered Venting System

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Lee, Byung Chul [FNC Technology Co. Ltd., Yongin (Korea, Republic of); Lee, Sang Won; Kim, Hyeong Taek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In this study, the thermal hydraulic conditions (e. g. pressure and flow rate) at each component have been examined and the sensitivity analysis on CFVS design parameters (e. g. water inventory, volumetric flow rate). The purpose is to know the possible range of flow conditions at each component to determine the optimum size of filtration system. GOTHIC code has been used to simulate the thermal-hydraulic behavior inside of CFVS. The behavior of flows in the CFVS has been investigated. The vessel water level and the flow rates during the CFVS operation are examined. It was observed that the vessel water level would be changed significantly due to steam condensation/thermal expansion and steam evaporation. Therefore, the vessel size and the initial water inventory should be carefully determined to keep the minimum water level required for filtration components and not to flood the components in the upper side of the vessel. It has been also observed that the volumetric flow rate is maintained during the CFVS operation, which is beneficial for pool scrubbing units. However, regarding the significant variations at the orifice downstream, careful design would be necessary.

  10. TENERIFE program: high temperature experiments on A 4 tons UF6 container

    International Nuclear Information System (INIS)

    Casselman, C.; Duret, B.; Seiler, J.M.; Ringot, C.; Warniez, P.; Wataru, M.; Shiomi, S.; Ozaki, S.; Yamakawa, H.

    1993-01-01

    To know the input of the future thermo-mechanical code, we have to get a better understanding of the thermo-physical evolution of the UF 6 which pressurizes the container. This evolution is function of: a) the heat transfer rate from the fire to the container b) the UF 6 behaviour in the container. These tests are essentially analytical at simulated fire temperatures of between 800 and 1000degC. They use a representative mass of UF 6 (around 4 tons). The tests will not seek to rupture the test container which has a diameter equal to the 48Y container, but shorter length. These tests carried out in realistic conditions (typical thermal gradient at the wall, characteristic period for UF 6 internal mass transfer) should make possible to improve knowledge of two fundamental phenomena: 1) vaporization of UF 6 on contact with the heated wall (around 400degC), a phenomenon which controls the container internal pressurization kinetic, 2) the equivalent conductivity of solid UF 6 , a phenomenon which is linked to the heat transfer by UF 6 vaporization-condensation through the solid's porosities and which depends on the diameter of the container. In addition, they will allow the influence of other parameters to be studied, such as UF 6 container filling mode or the mechanical characteristics of the container material. A UF 6 container fitted with instruments (wall temperature, UF 6 temperature, pressure) is heated by a rapid heat transient in a radiating furnace where the temperature and thermal power supplied can be measured. The test continues until pre-established thresholds have been reached: 1) strain threshold measured on the container surface (strain gauges positioned on the outside), 2) maximum temperature threshold of UF 6 , 3) container internal pressure threshold. (J.P.N.)

  11. Classification of structural component and degradation mechanisms for containment systems

    International Nuclear Information System (INIS)

    Judge, R.C.B.

    1994-01-01

    UK licence requirements for operation of nuclear power plants is dependent, inter alia, upon the licensee making and implementing adequate arrangements for the regular and systematic examination, inspection, maintenance and testing of all plant which may affect safety (Licence Condition 28). Similarly, the US NRC's Maintenance Rule (published in 10CFR50.65) specifies that a maintenance programme should be developed for plant systems, structures and components determined to be sensitive to ageing which will be used for the balance of the current (and, if relevant, extended) operating licence period. Against this background, the plant operators are seeking to minimise operating and maintenance costs and to enhance plant availability. This leads to a need to optimise the plant inspection and monitoring regimes whilst meeting regulatory requirements. In this paper, a conceptual framework for classifying civil structures and significant ageing mechanisms is described. This provides a systematic approach to making quantitative assessments of the likelihood and of potential degradation mechanisms and forms a consistent framework and a logical basis for prioritising inspection and maintenance schedules. The proposed method is analogous to a fault tree assessment, in which the likelihood of degradation due to a specific mechanism is considered as an event. The structures are considered in terms of their subcomponents. For each subcomponent, the value assigned to the likelihood of degradation is progressively reduced by a sequence of factors which make allowance for the structural and safety significance of any degradation and for the potential for timely detection of any degradation. Illustrative values for these factors are quoted in the text; it is recommended that these values are reviewed following a trial application of the method. (author)

  12. Interfacing system isolation experience review. Final report, August 1991

    International Nuclear Information System (INIS)

    1991-08-01

    A light water reactor power plant has auxiliary systems interconnected with the reactor coolant system that are not designed for reactor operating pressure. These principally include the shutdown heat removal systems and various emergency core cooling injection systems. There are multiple isolation valves that prevent rector vessel pressure from causing overpressurization in low pressure interfacing systems. Combinations of hardware failures or operational errors are necessary to expose these systems to overpressurization. This experience review provides insights regarding the risk that an auxiliary system might become over pressurized from the reactor system. While analyses show that for the pressures involved the probability of auxiliary system failure is low, the auxiliary system conceivably might fail outside of containment while the plant is at power. Such a potential event has come to be called an interfacing system loss of coolant accident (ISLOCA). This report provides a compilation of occurrences where valve leakage, valve failure, or valve mispositioning played a role in the ability to maintain interfacing system isolation. Seventeen U.S. BWR events, twenty three U.S. PWR events and one foreign event are discussed in the report. Eleven of the U.S. BWR events and ten U.S. PWR events are judged to relate directly to the so-called ISLOCA event in that they fulfilled one or more of the failures necessary to cause an ISLOCA. (author)

  13. 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

  14. 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.

  15. 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.)

  16. Experimental investigation of a two-phase closed thermosyphon assembly for passive containment cooling system

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kim, Sang Nyung [Kyunghee Univ., Gyeonggi-do (Korea, Republic of)

    2017-06-15

    After the Fukushima accident, increasing interest has been raised in passive safety systems that maintain the integrity of the containment building. To improve the reliability and safety of nuclear power plants, long-term passive cooling concepts have been developed for advanced reactors. In a previous study, the proposed design was based on an ordinary cylindrical Two-Phase Closed Thermosyphon (TPCT). The exact assembly size and number of TPCTs should be elaborated upon through accurate calculations based on experiments. While the ultimate goal is to propose an effective MPHP design for the PCCS and experimentally verify its performance, a TPCT assembly that was manufactured based on the conceptual design in this paper was tested.

  17. Experiences in development, qualification, and use of concrete high-integrity containers in commercial disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Reno, H.W.

    1985-01-01

    Disposal of EPICOR prefilters as commercial radioactive wastes is being accomplished by using a first-of-a-kind, reinforced concrete, high-integrity container in lieu of prior in situ solidification of resins before disposal of prefilters. Experiences in developing, testing, certifying, and using high-integrity containers are an untold story worthy of review for the benefit of the nuclear industry at large. The lessons learned in gaining regulatory acceptance of the concrete HIC are discussed

  18. Experiences in development, qualification, and use of concrete high-integrity containers in commercial disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Schmitt, R.C.; Reno, H.W.

    1985-01-01

    Disposal of EPICOR prefilters as commercial radioactive wastes is being accomplished by using a first-of-a-kind, reinforced concrete, high-integrity container (HIC) in lieu of prior in situ solidification of resins before disposal of prefilters. Experiences in developing, testing, certifying, and using high-integrity containers are an untold story worthy of review for the benefit of the nuclear industry at large. The lessons learned in gaining regulatory acceptance of the concrete HIC are discussed. 6 refs., 1 tab

  19. Process and container system for transferring or transporting fuel elements from a nuclear power station to a store

    International Nuclear Information System (INIS)

    Vox, A.J.

    1984-01-01

    A system of containers with three types of containers (an inside container, a transport container and a storage container) is used. One either sets the inside container open on the lid side into the transport container first in the water pond of the nuclear power station, and one then sets the fuel elements into the inside container, or one places the inside container, loaded with fuel elements away from the transport container, into the transport container. Both containers are then closed and are transported to the store as a unit. The storage container open on the lid side is prepared there, the floor of the transport container is opened and this, together with the inside container, is lifted above the storage container or set above the storage container. The inside container is then lowered onto the storage container, the transport container is removed and the lid of the storage container is closed. (orig./HP) [de

  20. The convertible client/server technology in large container inspection system

    International Nuclear Information System (INIS)

    Chen Zhiqiang; Zhang Li; Gao Wenhuan; Kang Kejun

    2001-01-01

    The author presents a new convertible client/server technology in distributed networking environment of a large container inspection system. The characteristic and advantage of this technology is introduced. The authors illustrate the policy of the technology to develop the networking program, and provide one example about how to program the software in large container inspection system using the new technology

  1. Women's Experience in the Workers' Compensation System.

    Science.gov (United States)

    Guthrie, Robert; Jansz, Janis

    2006-09-01

    Gender differences is a question of major importance within workers' compensation given the increased role of women in the workforce over the past several decades. This article reviews literature relating to women's experiences following work injury. An Australian study is used as background to exploring the broad issue of the question of gender equity in workers' compensation. In doing so it takes account of historical, legal and medical issues. Women's experience in the workers' compensation system is different to that of men due to a range of factors. It is heavily influenced by the industrial environment in which they work. Women are paid less than men in many instances and work in gender-segregated circumstances, which often reduces their industrial bargaining power. Women also suffer different forms of injury and disease to men because of the different nature of their work. The Australian experience suggests that as a consequence of the combination of lesser industrial bargaining power, lower wages and differing forms of injury and disease women often receive less than men in compensation payments, struggle to obtain equity in the dispute resolution process and experience greater difficulties in returning to work following injury or disease.

  2. 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)

  3. Immediate data access system for LHD experiments

    International Nuclear Information System (INIS)

    Emoto, M.; Iwadare, Y.; Nagayama, Y.

    2004-01-01

    Several kinds of computer systems are used to perform large helical device (LHD) experiments, and each produces its own data format. Therefore, it has been difficult to deal with these data simultaneously. In order to solve this problem, the Kaiseki server was developed; it has been facilitating the unified retrieval of LHD data. The data acquired or analyzed by various computer systems are converted into the unified ASCII format, or Kaiseki format, and transferred to the Kaiseki server. With this method, the researchers can visualize and analyze the data produced by various kinds of computers in the same way. Because validations are needed before registering on the Kaiseki server, it takes time to make the validated data available. However, some researchers need data as soon as it is gathered in order to adjust their instruments during the experiments. To satisfy this requirement, a new visualization system has been under development. The new system has two ways to visualize the data as physical values from the raw data. If the conversion task is not complex, the NIFSscope, a visualization tool, converts the raw data into physics data by itself. If the task is too complex to handle, it asks the ANACalc server to make physics data. When the ANACalc server receives a request, it delegates calculation programs to convert the acquired data into physics data. Because the interfaces between the server and the calculation processes are independent of programming languages and operating systems, the calculation processes can be placed on different computers and the server load can be reduced. Therefore, the system can respond to changes in requirements by replacing the calculation programs, and can easily be expanded by increasing the number of calculation servers

  4. Decision Support for Optimal Repositioning of Containers in a Feeder System

    Directory of Open Access Journals (Sweden)

    Danijela Tuljak-Suban

    2008-03-01

    Full Text Available The transport of empty containers represents a seriousproblem in the fast growing sphere of maritime container transport.The most widespread type of container transport organizationin maritime transport is the hub and spoke mode, whichenables the transport of a great number of containers via largevessels between hub ports, from where feeder ships transport tosmaller ports that thus gravitate to the central hub port. The articlecontains a detailed analysis of the northern Adriatic portsand the feeder connections with the hub ports of the Mediterranean.A two-level VRPPD (Vehicle Routing Problem withPickup and Delivery problem is modelled on a graph, wherethe transport of full containers is privileged over the transport ofempty containers. This enables the simulation of the feeder systemin the n01them Adriatic, meaning that it shows the ship'soperator the movement programme with minimal transportcosts for the superfluous empty containers in the complex of theregular transports of full containers in the feeder system.

  5. Physical experiments on the utility of non-iodine-containing contrast media

    International Nuclear Information System (INIS)

    Kirschner, H.; Burmester, U.; Stringaris, K.; Jentsch, F.

    1979-01-01

    Because of the excellent ability of the CT scanner in analyzing absorption, the use of non-iodine-containing contrast media will be discussed. Experimental studies of the chemical elements with the atomic number Z = 1-83 with a scan tension of 120 kV were made to determine how many atoms of a given element are necessary to replace one iodine atom in an iodine-containing contrast medium, whereby the same contrast enhancement in the scan picture as with the iodine-containing contrast medium is produced. (orig.) 891 ORU/orig. 892 MB [de

  6. Experimental study on iodine chemistry (EXSI) - Containment experiments with elemental iodine

    International Nuclear Information System (INIS)

    Kaerkelae, T.; Auvinen, A.; Holm, J.; Ekberg, C.; Glaenneskog, H.

    2009-10-01

    The behaviour of iodine during a severe accident has been studied in several experimental programs, ranging from the large-scale PHEBUS FP tests and intermediate-scale ThAI tests to numerous separate effect studies. Oxidation of iodine in gas phase has been one of the greatest remaining uncertainties in iodine behaviour during a severe accident. In this study the possible formation of iodine oxide aerosol due to radiolytic oxidation of gaseous iodine is experimentally tested and the reaction products are analysed. The experimental facility applied in this study is based on the sampling system built at VTT for ISTP program project CHIP conducted IRSN. The experimental facility and the measuring technology are sophisticated and unique in the area of nuclear research as well as in the field of aerosol science. The results from the experiments show an extensive particle formation when ozone and gaseous iodine react with each other. The formed particles were collected on filters, while gaseous iodine was trapped into bubbles. The particles were iodine oxides and the size of particles was approximately 100 nm. The transport of gaseous iodine through the facility decreased when both gaseous iodine and ozone were fed together into facility. Experimental study on radiolytic oxidation of iodine was conducted in co-operation between VTT and Chalmers University of Technology as a part of the NKS-R programs. (author)

  7. Experimental study on iodine chemistry (EXSI) - Containment experiments with elemental iodine

    Energy Technology Data Exchange (ETDEWEB)

    Kaerkelae, T.; Auvinen, A. (VTT Technical Research Centre of Finland (Finland)); Holm, J.; Ekberg, C. (Chalmers Univ. of Technology (Sweden)); Glaenneskog, H. (Vattenfall Power Consultant (Sweden))

    2009-10-15

    The behaviour of iodine during a severe accident has been studied in several experimental programs, ranging from the large-scale PHEBUS FP tests and intermediate-scale ThAI tests to numerous separate effect studies. Oxidation of iodine in gas phase has been one of the greatest remaining uncertainties in iodine behaviour during a severe accident. In this study the possible formation of iodine oxide aerosol due to radiolytic oxidation of gaseous iodine is experimentally tested and the reaction products are analysed. The experimental facility applied in this study is based on the sampling system built at VTT for ISTP program project CHIP conducted IRSN. The experimental facility and the measuring technology are sophisticated and unique in the area of nuclear research as well as in the field of aerosol science. The results from the experiments show an extensive particle formation when ozone and gaseous iodine react with each other. The formed particles were collected on filters, while gaseous iodine was trapped into bubbles. The particles were iodine oxides and the size of particles was approximately 100 nm. The transport of gaseous iodine through the facility decreased when both gaseous iodine and ozone were fed together into facility. Experimental study on radiolytic oxidation of iodine was conducted in co-operation between VTT and Chalmers University of Technology as a part of the NKS-R programs. (author)

  8. Experimental study on iodine chemistry (EXSI) - Containment experiments with methyl iodide

    Energy Technology Data Exchange (ETDEWEB)

    Holm, J.; Glaenneskog, H.; Ekberg, C. (Chalmers Univ. of Technology (Sweden)); Kaerkelae, T.; Auvinen, A. (VTT Technical Research Centre of Finland (Finland))

    2010-05-15

    An experimental study on radiolytic decomposition of methyl iodide was conducted in co-operation between VTT and Chalmers University of Technology as a part of the NKS-R programs. The behaviour of iodine during a severe accident has been studied in several experimental programs, ranging from the large-scale PHEBUS FP tests and intermediate-scale ThAI tests to numerous separate effect studies. In year 2008 the NROI project, a Nordic collaboration studying iodine chemistry in the containment was started. During 2009, oxidation of iodine, especially organic iodine, was studied within the NROI project. The chemistry of organic iodine in the gas phase is still one of the greatest remaining uncertainties concerning iodine behaviour during a severe accident. During the first year of the NROI project the oxidation of elemental iodine, I2, with ozone and UV-light was investigated. In this study organic iodide, in this case methyl iodide, was investigated in similar conditions as in the NROI-1 project. The experimental facility applied in this study is based on the sampling system built at VTT for the ISTP project CHIP conducted by IRSN. The experimental facility and the measuring technology are sophisticated and unique in the area of nuclear research as well as in the field of aerosol science. Experimental results showed that the methyl iodide concentration in the facility was reduced with increasing temperature and increasing UVC intensity. Similar behaviour occurred when ozone was present in the system. Formed organic gas species during the decomposition of methyl iodide was mainly formaldehyde and methanol. Instant and extensive particle formation occurred when methyl iodide was transported through a UVC radiation field and/or when ozone was present. The size of the formed primary particles was about 10 nm and the size of secondary particles was between 50-150 nm. From the SEM-EDX analyses of the particles, the conclusion was drawn that these were some kind of iodine

  9. Experimental study on iodine chemistry (EXSI) - Containment experiments with methyl iodide

    International Nuclear Information System (INIS)

    Holm, J.; Glaenneskog, H.; Ekberg, C.; Kaerkelae, T.; Auvinen, A.

    2010-05-01

    An experimental study on radiolytic decomposition of methyl iodide was conducted in co-operation between VTT and Chalmers University of Technology as a part of the NKS-R programs. The behaviour of iodine during a severe accident has been studied in several experimental programs, ranging from the large-scale PHEBUS FP tests and intermediate-scale ThAI tests to numerous separate effect studies. In year 2008 the NROI project, a Nordic collaboration studying iodine chemistry in the containment was started. During 2009, oxidation of iodine, especially organic iodine, was studied within the NROI project. The chemistry of organic iodine in the gas phase is still one of the greatest remaining uncertainties concerning iodine behaviour during a severe accident. During the first year of the NROI project the oxidation of elemental iodine, I2, with ozone and UV-light was investigated. In this study organic iodide, in this case methyl iodide, was investigated in similar conditions as in the NROI-1 project. The experimental facility applied in this study is based on the sampling system built at VTT for the ISTP project CHIP conducted by IRSN. The experimental facility and the measuring technology are sophisticated and unique in the area of nuclear research as well as in the field of aerosol science. Experimental results showed that the methyl iodide concentration in the facility was reduced with increasing temperature and increasing UVC intensity. Similar behaviour occurred when ozone was present in the system. Formed organic gas species during the decomposition of methyl iodide was mainly formaldehyde and methanol. Instant and extensive particle formation occurred when methyl iodide was transported through a UVC radiation field and/or when ozone was present. The size of the formed primary particles was about 10 nm and the size of secondary particles was between 50-150 nm. From the SEM-EDX analyses of the particles, the conclusion was drawn that these were some kind of iodine

  10. 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

  11. Containment Domains: A Scalable, Efficient and Flexible Resilience Scheme for Exascale Systems

    Directory of Open Access Journals (Sweden)

    Jinsuk Chung

    2013-01-01

    Full Text Available This paper describes and evaluates a scalable and efficient resilience scheme based on the concept of containment domains. Containment domains are a programming construct that enable applications to express resilience needs and to interact with the system to tune and specialize error detection, state preservation and restoration, and recovery schemes. Containment domains have weak transactional semantics and are nested to take advantage of the machine and application hierarchies and to enable hierarchical state preservation, restoration and recovery. We evaluate the scalability and efficiency of containment domains using generalized trace-driven simulation and analytical analysis and show that containment domains are superior to both checkpoint restart and redundant execution approaches.

  12. The systems psychodynamic experiences of organisational transformation amongst support staff

    Directory of Open Access Journals (Sweden)

    Martin Steyn

    2016-10-01

    Full Text Available Orientation: The unconscious impact of organisational transformation is often neglected and even denied. This research revealed the manifestation and impact of high levels and different forms of anxiety experienced by employees during transformation. Research objective: The objective was to study and describe the manifesting systems psychodynamic behaviour amongst support staff during organisational transformation. Motivation for the study: Organisational transformation is mostly researched from a leadership viewpoint. Little research data are available on the experiences of support staff on the receiving end of decisions about and implementation of transformation. Research design, approach and method: A qualitative approach within the phenomenological hermeneutic interpretive stance was used. The research was set in a government organisation. A semi-structured interview with four conveniently and purposefully chosen support staff members was thematically analysed using systems psychodynamics as theoretical paradigm. Main findings: Four themes manifested, namely de-authorisation and detachment, being bullied and seduced by leadership, the organisation in the mind as incompetent, and a dangerous and persecutory system. In the discussion, the basic assumptions and relevant constructs are interpreted. Practical implications: Understanding the transformation experiences of support staff could assist the industrial psychologist to facilitate appropriate support in coaching more junior staff towards increasing wellness and work performance. Contribution: Organisational transformation is highlighted as an anxiety provoking experience especially on the lower levels of the organisation. Its potentially deep and complex psychological impact could possibly derail parts of the system if not managed in a psychologically contained manner.

  13. Evaluation of Container Closure System Integrity for Frozen Storage Drug Products.

    Science.gov (United States)

    Nieto, Alejandra; Roehl, Holger; Brown, Helen; Nikoloff, Jonas; Adler, Michael; Mahler, Hanns-Christian

    2016-01-01

    Sometimes, drug product for parenteral administration is stored in a frozen state (e.g., -20 °C or -80 °C), particularly during early stages of development of some biotech molecules in order to provide sufficient stability. Shipment of frozen product could potentially be performed in the frozen state, yet possibly at different temperatures, for example, using dry ice (-80 °C). Container closure systems of drug products usually consist of a glass vial, rubber stopper, and an aluminum crimped cap. In the frozen state, the glass transition temperature (Tg) of commonly used rubber stoppers is between -55 and -65 °C. Below their Tg, rubber stoppers are known to lose their elastic properties and become brittle, and thus potentially fail to maintain container closure integrity in the frozen state. Leaks during frozen temperature storage and transportation are likely to be transient, yet, can possibly risk container closure integrity and lead to microbial contamination. After thawing, the rubber stopper is supposed to re-seal the container closure system. Given the transient nature of the possible impact on container closure integrity in the frozen state, typical container closure integrity testing methods (used at room temperature conditions) are unable to evaluate and thus confirm container closure integrity in the frozen state. Here we present the development of a novel method (thermal physical container closure integrity) for direct assessment of container closure integrity by a physical method (physical container closure integrity) at frozen conditions, using a modified He leakage test. In this study, different container closure systems were evaluated with regard to physical container closure integrity in the frozen state to assess the suitability of vial/stopper combinations and were compared to a gas headspace method. In summary, the thermal physical container closure integrity He leakage method was more sensitive in detecting physical container closure

  14. Application and study of advanced network technology in large container inspection system

    International Nuclear Information System (INIS)

    Li Zheng; Kang Kejun; Gao Wenhuan; Wang Jingjin

    1996-01-01

    Large Container Inspection System (LCIS) based on radiation imaging technology is a powerful tool for the customs to check the contents inside a large container without opening it. An image distributed network system is composed of center manager station, image acquisition station, environment control station, inspection processing station, check-in station, check-out station, database station by using advanced network technology. Mass data, such as container image data, container general information, manifest scanning data, commands and status, must be on-line transferred between different stations. Advanced network technology and software programming technique are presented

  15. FAUST/CONTAIN; FAUST/CONTAIN

    Energy Technology Data Exchange (ETDEWEB)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-08-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  16. Process systems of PHWR - Indian experience

    Energy Technology Data Exchange (ETDEWEB)

    Ramandan, T S.V. [Madras Atomic Power Station (MAPS), Madras (India)

    1991-04-01

    Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

  17. Process systems of PHWR - Indian experience

    International Nuclear Information System (INIS)

    Ramandan, T.S.V.

    1991-01-01

    Three operational problems are discussed in this paper. The reactors in Madras Atomic Power Station (MAPS), India are Pressurised Heavy Water Reactors PHWR), similar to Douglas Point PGS. The moderator heavy water is pumped into the bottom half of the calandria (horizontal reactor vessel) through one inlet manifold plenum chamber and horizontal louvers which help to distribute the moderator evenly at a very low velocity. The outlet from the calandria is through a smaller manifold structure at a higher elevation. The moderator is held on the shell side of the calandria by means of helium gas pressure differential between top of calandria and dump tank located below. The primary coolant system consists of 306 coolant channels containing the fuel and steam generators (SGs) and pumps on either side of the reactor. Each SC consists of 11 Nos. inverted U tube vertical heat exchangers where heat is transferred from primary coolant heavy water to secondary light water to produce steam. (author)

  18. A magnet system for HEP experiments

    CERN Document Server

    Gaddi, A

    2012-01-01

    This chapter describes the sequence of steps that lead to the design of a magnet system for modern HEP detectors. We start looking to the main types of magnets used in HEP experiments, along with some basic formulae to set the main parameters, such as ampere-turns, impedance and stored energy. A section is dedicated to the description of the iron yoke, with emphasis on magnet-detector integration and assembly, steel characteristics, stray field issues and alternative design. In the second part of the chapter we start looking at a brief history of superconducting magnets and a comparison between warm and superconducting ones. Following that, we describe the commonly used superconducting cables, the conductor design and technology and the winding techniques. A section of the chapter is dedicated to the cryogenic design, vacuum insulation and other ancillary systems. We also describe the power circuit, with the power supply unit, the current leads, the current measurement devices and other instruments and safety...

  19. Plasma focus system: Design, construction and experiments

    International Nuclear Information System (INIS)

    Alacakir, A.; Akguen, Y.; Boeluekdemir, A. S.

    2007-01-01

    The aim of this work is to construct a compact experimental system for fusion research. The design, construction and experiments of the 3 kJ Mather type plasma focus machine is described. This machine is established for neutron yield and fast neutron radiography by D-D reaction which is given by D + D→ 3 He (0.82 MeV) + n (2.45 MeV) . Investigation of the geometry of plasma focus machine in the presence of high voltage drive and vacuum system setup is shown. 108 neutron per pulse and 200 kA peak current is obtained for many shots. Scintillator screen for fast neutron imaging, sensitive to 2.45 MeV neutrons, is also manufactured in our labs. Structural neutron shielding computations for safety is also completed

  20. Qinshan plant display system: experience to date

    International Nuclear Information System (INIS)

    Bin, L.; Jiangdong, Y.; Weili, C.; Haidong, W.; Wangtian, L.; Lockwood, R.; Doucet, R.; Trask, D.; Judd, R.

    2004-01-01

    The two CANDU 6 units operated by the Third Qinshan Nuclear Power Corporation (TQNPC) include, as part of a control centre upgrade, a new plant display system (PDS). The PDS provides plant operators with new display and monitoring functionality designed to compliment the DCC capability. It includes new overview and trend displays (e.g., critical safety parameter monitor and user-defined trends), and enhanced annunciation based on AECL's Computerized Alarm Message List System (CAMLS) including an alarm interrogation capability. This paper presents a review of operating experience gained since the PDS was commissioned more than three years ago. It includes feedback provided by control room operators and trainers, PDS maintainers, and AECL development and support staff. It also includes an overview of improvements implemented since the PDS and suggestions for the future enhancements. (author)

  1. Experiments to investigate direct containment heating phenomena with scaled models of the Calvert Cliffs Nuclear Power Plant

    International Nuclear Information System (INIS)

    Blanchat, T.K.; Pilch, M.M.; Allen, M.D.

    1997-02-01

    The Surtsey Test Facility is used to perform scaled experiments simulating High Pressure Melt Ejection accidents in a nuclear power plant (NPP). The experiments investigate the effects of direct containment heating (DCH) on the containment load. The results from Zion and Surry experiments can be extrapolated to other Westinghouse plants, but predicted containment loads cannot be generalized to all Combustion Engineering (CE) plants. Five CE plants have melt dispersal flow paths which circumvent the main mitigation of containment compartmentalization in most Westinghouse PWRs. Calvert Cliff-like plant geometries and the impact of codispersed water were addressed as part of the DCH issue resolution. Integral effects tests were performed with a scale model of the Calvert Cliffs NPP inside the Surtsey test vessel. The experiments investigated the effects of codispersal of water, steam, and molten core stimulant materials on DCH loads under prototypic accident conditions and plant configurations. The results indicated that large amounts of coejected water reduced the DCH load by a small amount. Large amounts of debris were dispersed from the cavity to the upper dome (via the annular gap). 22 refs., 84 figs., 30 tabs

  2. Headspace concentrations of explosive vapors in containers designed for canine testing and training: theory, experiment, and canine trials.

    Science.gov (United States)

    Lotspeich, Erica; Kitts, Kelley; Goodpaster, John

    2012-07-10

    It is a common misconception that the amount of explosive is the chief contributor to the quantity of vapor that is available to trained canines. In fact, this quantity (known as odor availability) depends not only on the amount of explosive material, but also the container volume, explosive vapor pressure and temperature. In order to better understand odor availability, headspace experiments were conducted and the results were compared to theory. The vapor-phase concentrations of three liquid explosives (nitromethane, nitroethane and nitropropane) were predicted using the Ideal Gas Law for containers of various volumes that are in use for canine testing. These predictions were verified through experiments that varied the amount of sample, the container size, and the temperature. These results demonstrated that the amount of sample that is needed to saturate different sized containers is small, predictable and agrees well with theory. In general, and as expected, once the headspace of a container is saturated, any subsequent increase in sample volume will not result in the release of more vapors. The ability of canines to recognize and alert to differing amounts of nitromethane has also been studied. In particular, it was found that the response of trained canines is independent of the amount of nitromethane present, provided it is a sufficient quantity to saturate the container in which it is held. Copyright © 2012 Elsevier Ireland Ltd. All rights reserved.

  3. Investigations into the design of a filter system for PWR containment venting

    International Nuclear Information System (INIS)

    Dillmann, H.G.; Wilhelm, J.G.

    1991-01-01

    The reactors of power stations in the Federal Republic of Germany are being or have already been equipped with systems for containment venting under severe accident conditions. Two different offgas cleaning systems are available. One system, realizing a complete passive filtering concept, consists of a multistage metal fiber filter for coarse particulates and aerosol removal and an additional molecular sieve filter for gaseous iodine retention connected in series. The requirements made with respect to aerosol filtration includes among others the capability of retaining 60 kg of a recondensing aerosol with a 0.5 μm mean geometric mass diameter. BaSO 4 and SnO 2 were used as tracer aerosols in the experiments. All the decontamination factors were > 1,000. Vaporous iodine is removed on molecular sieve filters (zeolite filters) subsequent to airborne particulate filtration. As Ag-zeolites act as catalysts in the H 2 O 2 reaction and thus might give rise to a violent exothermal reaction, the catalytic effect was suppressed by substituting mixed doping for doping solely with silver. The removal efficiencies achieved with Ag-zeolites and zeolites with mixed doping in air-steam mixtures are indicated, and investigations of the catalytic behavior in air-steam-H 2 mixtures are described

  4. Experiments on interactions between zirconium-containing melt and water (ZREX). Hydrogen generation and chemical augmentation of energetics

    Energy Technology Data Exchange (ETDEWEB)

    Cho, D.H.; Armstrong, D.R.; Gunther, W.H. [Argonne National Lab., IL (United States); Basu, S.

    1998-01-01

    The results of the first data series of experiments on interactions between zirconium-containing melt and water are described. These experiments involved dropping 1-kg batches of pure zirconium or zirconium-zirconium dioxide mixture melt into a column of water. A total of nine tests were conducted, including four with pure zirconium melt and five with Zr-ZrO{sub 2} mixture melt. Explosions took place only in those tests which were externally triggered. While the extent of zirconium oxidation in the triggered experiments was quite extensive, the estimated explosion energetics were found to be very small compared to the combined thermal and chemical energy available. (author)

  5. Advanced domestic digital satellite communications systems experiments

    Science.gov (United States)

    Iso, A.; Izumisawa, T.; Ishida, N.

    1984-02-01

    The characteristics of advanced digital transmission systems were measured, using newly developed small earth stations and a K-band and C-band communication satellite. Satellite link performance for data, facsimile, video and packet switching information transmission at bit rates ranging from 6.4 kbit/s to 6.3 Mbit/s have been confirmed, using a small K-band earth station and a demand-assignment time division multiple access system. A low-capacity omni-use C-band terminal experiment has verified a telephone channel transmission performance by spread-spectrum multiple access. Single point to multipoint transmission characteristics of the 64 kbit/s data signals from the computer center were tested, using a receive-only 4 GHz earth terminal. Basic satellite link performance was confirmed under clear-sky conditions. Precise satellite orbit and attitude keeping experiments were carried out to obtain precise satellite antenna pointing accuracy for development of K-band earth stations that do not require satellite tracking equipment. Precise station keeping accuracy of 0.02 degrees was obtained.

  6. Satellite power system (SPS) public outreach experiment

    Energy Technology Data Exchange (ETDEWEB)

    McNeal, S.R.

    1980-12-01

    To improve the results of the Satellite Power System (SPS) Concept Development and Evaluation Program, an outreach experiment was conducted. Three public interest groups participated: the L-5 Society (L-5), Citizen's Energy Project (CEP), and the Forum for the Advancement of Students in Science and Technology (FASST). Each group disseminated summary information about SPS to approximately 3000 constituents with a request for feedback on the SPS concept. The objectives of the outreach were to (1) determine the areas of major concern relative to the SPS concept, and (2) gain experience with an outreach process for use in future public involvement. Due to the combined efforts of all three groups, 9200 individuals/organizations received information about the SPS concept. Over 1500 receipients of this information provided feedback. The response to the outreach effort was positive for all three groups, suggesting that the effort extended by the SPS Project Division to encourage an information exchange with the public was well received. The general response to the SPS differed with each group. The L-5 position is very much in favor of SPS; CEP is very much opposed and FASST is relatively neutral. The responses are analyzed, and from the responses some questions and answers about the satellite power system are presented in the appendix. (WHK)

  7. Modeling bubble condenser containment with computer code COCOSYS: post-test calculations of the main steam line break experiment at ELECTROGORSK BC V-213 test facility

    International Nuclear Information System (INIS)

    Lola, I.; Gromov, G.; Gumenyuk, D.; Pustovit, V.; Sholomitsky, S.; Wolff, H.; Arndt, S.; Blinkov, V.; Osokin, G.; Melikhov, O.; Melikhov, V.; Sokoline, A.

    2005-01-01

    Containment of the WWER-440 Model 213 nuclear power plant features a Bubble Condenser, a complex passive pressure suppression system, intended to limit pressure rise in the containment during accidents. Due to lack of experimental evidence of its successful operation in the original design documentation, the performance of this system under accidents with ruptures of large high-energy pipes of the primary and secondary sides remains a known safety concern for this containment type. Therefore, a number of research and analytical studies have been conducted by the countries operating WWER-440 reactors and their Western partners in the recent years to verify Bubble Condenser operation under accident conditions. Comprehensive experimental research studies at the Electrogorsk BC V-213 test facility, commissioned in 1999 in Electrogorsk Research and Engineering Centre (EREC), constitute essential part of these efforts. Nowadays this is the only operating large-scale facility enabling integral tests on investigation of the Bubble Condenser performance. Several large international research projects, conducted at this facility in 1999-2003, have covered a spectrum of pipe break accidents. These experiments have substantially improved understanding of the overall system performance and thermal hydraulic phenomena in the Bubble Condenser Containment, and provided valuable information for validating containment codes against experimental results. One of the recent experiments, denoted as SLB-G02, has simulated steam line break. The results of this experiment are of especial value for the engineers working in the area of computer code application for WWER-440 containment analyses, giving an opportunity to verify validity of the code predictions and identify possibilities for model improvement. This paper describes the results of the post-test calculations of the SLB-G02 experiment, conducted as a joint effort of GRS, Germany and Ukrainian technical support organizations for

  8. The uranium liquid argon calorimeter of the D0 experiment: Experience in realizing a large system

    International Nuclear Information System (INIS)

    Guryn, W.

    1991-01-01

    The major aspects in realizing the calorimeter system of the D OE experiment are discussed. They include: technologies developed for calorimeter production, schedule, and experience with module production

  9. Heat Transfer Reactor Experiment (HTRE)-3 Container Storage Unit Resource Conservation Recovery Act closure plan

    International Nuclear Information System (INIS)

    Spry, M.J.

    1992-11-01

    This document describes the closure of the HTRE-3 Container Storage Unit under the requirements of the Resource Conservation and Recovery Act. The unit's location, size, history, and current status are described. The document also summarizes the decontamination and decommissioning efforts performed in 1983 and provides an estimate of,waste residues remaining in the HTRE-3 assembly. A risk evaluation was performed that demonstrates that the residue does not pose a hazard to public health or the environment. Based on the risk evaluation, it is proposed that the HTRE-3 Container Storage Unit be closed in its present condition, without further decontamination or removal activities

  10. Management of empty pesticide containers – An experience from Santa Cruz, Bolivia 2014-16

    DEFF Research Database (Denmark)

    Huici, Omar; Jensen, Olaf Chresten; Jørs, Erik

    The mismanagement of empty containers of pesticides, posing a risk to the environment and the health of people, has motivated the promotion of international policies and guidelines to mitigate the problems. Despite these guidelines the attention to this problem is inadequate in Bolivia. The objec......The mismanagement of empty containers of pesticides, posing a risk to the environment and the health of people, has motivated the promotion of international policies and guidelines to mitigate the problems. Despite these guidelines the attention to this problem is inadequate in Bolivia...

  11. Operational Experience with the Frontier System in CMS

    International Nuclear Information System (INIS)

    Blumenfeld, Barry; Dykstra, Dave; Kreuzer, Peter; Du Ran; Wang Weizhen

    2012-01-01

    The Frontier framework is used in the CMS experiment at the LHC to deliver conditions data to processing clients worldwide, including calibration, alignment, and configuration information. Each central server at CERN, called a Frontier Launchpad, uses tomcat as a servlet container to establish the communication between clients and the central Oracle database. HTTP-proxy Squid servers, located close to clients, cache the responses to queries in order to provide high performance data access and to reduce the load on the central Oracle database. Each Frontier Launchpad also has its own reverse-proxy Squid for caching. The three central servers have been delivering about 5 million responses every day since the LHC startup, containing about 40 GB data in total, to more than one hundred Squid servers located worldwide, with an average response time on the order of 10 milliseconds. The Squid caches deployed worldwide process many more requests per day, over 700 million, and deliver over 40 TB of data. Several monitoring tools of the tomcat log files, the accesses of the Squids on the central Launchpad servers, and the availability of remote Squids have been developed to guarantee the performance of the service and make the system easily maintainable. Following a brief introduction of the Frontier framework, we describe the performance of this highly reliable and stable system, detail monitoring concerns and their deployment, and discuss the overall operational experience from the first two years of LHC data-taking.

  12. Operational Experience with the Frontier System in CMS

    Science.gov (United States)

    Blumenfeld, Barry; Dykstra, Dave; Kreuzer, Peter; Du, Ran; Wang, Weizhen

    2012-12-01

    The Frontier framework is used in the CMS experiment at the LHC to deliver conditions data to processing clients worldwide, including calibration, alignment, and configuration information. Each central server at CERN, called a Frontier Launchpad, uses tomcat as a servlet container to establish the communication between clients and the central Oracle database. HTTP-proxy Squid servers, located close to clients, cache the responses to queries in order to provide high performance data access and to reduce the load on the central Oracle database. Each Frontier Launchpad also has its own reverse-proxy Squid for caching. The three central servers have been delivering about 5 million responses every day since the LHC startup, containing about 40 GB data in total, to more than one hundred Squid servers located worldwide, with an average response time on the order of 10 milliseconds. The Squid caches deployed worldwide process many more requests per day, over 700 million, and deliver over 40 TB of data. Several monitoring tools of the tomcat log files, the accesses of the Squids on the central Launchpad servers, and the availability of remote Squids have been developed to guarantee the performance of the service and make the system easily maintainable. Following a brief introduction of the Frontier framework, we describe the performance of this highly reliable and stable system, detail monitoring concerns and their deployment, and discuss the overall operational experience from the first two years of LHC data-taking.

  13. Experiences from occupational exposure limits set on aerosols containing allergenic proteins

    DEFF Research Database (Denmark)

    Nielsen, Gunnar; Larsen, Søren; Hansen, Jitka S

    2012-01-01

    Occupational exposure limits (OELs) together with determined airborne exposures are used in risk assessment based managements of occupational exposures to prevent occupational diseases. In most countries, OELs have only been set for few protein-containing aerosols causing IgE-mediated allergies. ...... is available for setting OELs for proteins and protein-containing aerosols where the critical effect is IgE sensitization and IgE-mediated airway diseases.......Occupational exposure limits (OELs) together with determined airborne exposures are used in risk assessment based managements of occupational exposures to prevent occupational diseases. In most countries, OELs have only been set for few protein-containing aerosols causing IgE-mediated allergies...... for setting OELs. Our aim is to analyse prerequisites for setting OELs for the allergenic protein-containing aerosols. Opposite to the key effect of toxicological reactions, two thresholds, one for the sensitization phase and one for elicitation of IgE-mediated symptoms in sensitized individuals, are used...

  14. Remote-Reading Safety and Safeguards Surveillance System for 3013 Containers

    International Nuclear Information System (INIS)

    Lechelt, W. M.; Skorpik, J. R.; Silvers, K. L.; Szempruch, R. W.; Douglas, D. G.; Fein, K. O.

    2002-01-01

    At Hanford's Plutonium Finishing Plant (PFP), plutonium oxide is being loaded into stainless steel containers for long-term storage on the Hanford Site. These containers consist of two weld-sealed stainless steel cylinders nested one within the other. A third container holds the plutonium within the inner cylinder. This design meets the U.S. Department of Energy (DOE) storage standard, DOE-STD- 3013-2000, which anticipates a 50-year storage lifetime. The 3013 standard also requires a container surveillance program to continuously monitor pressure and to assure safeguards are adequate. However, the configuration of the container system makes using conventional measurement and monitoring methods difficult. To better meet the 3013 monitoring requirements, a team from Fluor Hanford (who manages the PFP), Pacific Northwest National Laboratory (PNNL), and Vista Engineering Technologies, LLC, developed a safer, cost-efficient, remote PFP 3013 container surveillance system. This new surveillance system is a combination of two successfully deployed technologies: (1) a magnetically coupled pressure gauge developed by Vista Engineering and (2) a radio frequency (RF) tagging device developed by PNNL. This system provides continuous, 100% monitoring of critical parameters with the containers in place, as well as inventory controls. The 3013 container surveillance system consists of three main elements: (1) an internal magnetic pressure sensor package, (2) an instrument pod (external electronics package), and (3) a data acquisition storage and display computer. The surveillance system described in this paper has many benefits for PFP and DOE in terms of cost savings and reduced personnel exposure. In addition, continuous safety monitoring (i.e., internal container pressure and temperature) of every container is responsible nuclear material stewardship and fully meets and exceeds DOE's Integrated Surveillance Program requirements

  15. Experiments on container materials for Swiss high-level waste disposal projects. Part IV

    International Nuclear Information System (INIS)

    Simpson, J.P.

    1989-12-01

    One concept for final disposal of high-level waste in switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of Northern Switzerland. The waste will be placed in a container which will be required to function as a high integrity barrier for at least 100 years. This report is the fourth and last in the current series dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports: cast steel, nodular cast iron, copper and Ti-Code 12. This report deals with the problem of demonstrating that cast steel containers will not fail by stress corrosion cracking and with the problem of hydrogen produced by the reduction of water. The experimental results on pre-cracked specimens revealed no susceptibility of cast steel to stress corrosion cracking under model repository conditions. No crack growth was detected on compact DCB specimens exposed in aerobic and anaerobic groundwaters at 80 and 140 o C for 16-24 months. Cast steel remains a candidate material for high-level waste containers. As expected from thermodynamic considerations no hydrogen could be detected from copper immersed in model groundwaters at 50 o C. Hydrogen is evolved from corroding steel under anaerobic conditions. Hydrogen evolution due to corrosion of iron or steel in waste repositories has to be considered in any safety analysis; the amounts produced can be significant. Evidence todate suggests that both cast steel and copper are suitable container materials. Because the corrosion behaviour of both materials is sensitive to service conditions, in particular length of the aerobic phase, groundwater chemistry and temperature, further testing should be undertaken when a specific site has been identified. (author) 9 tabs., 11 figs., 25 refs

  16. Experiments on container materials for Swiss high-level waste disposal projects. Part 2

    International Nuclear Information System (INIS)

    Simpson, J.P.

    1984-12-01

    The present concept for final disposal of high-level waste in Switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of northern Switzerland. The waste will be placed in a container which is required to function as a high integrity barrier for at least 1000 years. This report is the second of a set of two dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports; they were cast steel, nodular cast iron, copper and Ti-Code 12. It was concluded that some testing was needed, in particular with respect to corrosion, in order to confirm these materials as candidate container materials. The experimental programme included: 1) corrosion tests on copper under gamma radiation; 2) immersion corrosion tests on the four candidate materials including welded specimens; 3) corrosion testing of the four materials in saturated bentonite; 4) constant strain rate testing of Ti-Code 12 and copper at 80 degrees C; 5) the behaviour of copper, Ti-Code 12 and Zircaloy-2 when immersed in liquid lead; 6) corrosion potential and galvanic current measurements on several material pairs. The standard test medium was natural mineral water from the Bad Saeckingen source. This water has a total dissolved solids content of approx. 3200 mg/l, about 1600 mg/l as chloride. The oxygen level was defined as 0.1 μg/g. In certain cases this medium was modified in order to test under more severe conditions. The results of the corrosion tests confirm in general the evaluation in the first part of the report. All of the materials are suitable for high-level waste containers: cast steel, nodular cast iron and copper as single layer containers, and Ti-Code 12 as an outer corrosion resistant layer. Copper could also be used under an outer steel layer, where it could arrest local penetration

  17. Design and implementation of check out station for large container inspection system

    International Nuclear Information System (INIS)

    Yao Dongsheng; Gao Wenhuan; Kang Kejun

    1997-01-01

    In Large Container Inspection System (LCIS), Check Out Station (COS) is in charge of deciding whether a container is allowed to pass or has to be opened for checking. Several different top level architecture designs for COS are discussed and analyzed according to the practical requirements of COS

  18. The Information Science Experiment System - The computer for science experiments in space

    Science.gov (United States)

    Foudriat, Edwin C.; Husson, Charles

    1989-01-01

    The concept of the Information Science Experiment System (ISES), potential experiments, and system requirements are reviewed. The ISES is conceived as a computer resource in space whose aim is to assist computer, earth, and space science experiments, to develop and demonstrate new information processing concepts, and to provide an experiment base for developing new information technology for use in space systems. The discussion covers system hardware and architecture, operating system software, the user interface, and the ground communication link.

  19. Study on transient hydrogen behavior and effect on passive containment cooling system of the advanced PWR

    International Nuclear Information System (INIS)

    Wang Yan

    2014-01-01

    A certain amount of hydrogen will be generated due to zirconium-steam reaction or molten corium concrete interaction during severe accidents in the pressurized water reactor (PWR). The generated hydrogen releases into the containment, and the formed flammable mixture might cause deflagration or detonation to produce high thermal and pressure loads on the containment, which may threaten the integrity of the containment. The non-condensable hydrogen in containment may also reduce the steam condensation on the containment surface to affect the performance of the passive containment cooling system (PCCS). To study the transient hydrogen behavior in containment with the PCCS performance during the accidents is significant for the further study on the PCCS design and the hydrogen risk mitigation. In this paper, a new developed PCCS analysis code with self-reliance intellectual property rights, which had been validated by comparison on the transients in the containment during the design basis accidents with other developed PCCS analysis code, is brief introduced and used for the transient simulation in the containment under a postulated small break LOCA of cold-leg. The results show that the hydrogen will flow upwards with the coolant released from the break and spread in the containment by convection and diffusion, and it results in the increase of the pressure in the containment due to reducing the heat removal capacity of the PCCS. (author)

  20. Disposal containers for radioactive waste materials and separation systems for radioactive waste materials

    International Nuclear Information System (INIS)

    Rubin, L.S.

    1986-01-01

    A separation system for dewatering radioactive waste materials includes a disposal container, drive structure for receiving the container, and means for releasably attaching the container to the drive structure. The separation structure disposed in the container adjacent the inner surface of the side wall structure retains solids while allowing passage of liquids. The inlet port structure in the container top wall is normally closed by first valve structure that is centrifugally actuated to open the inlet port and the discharge port structure at the container periphery receives liquid that passes through the separation structure and is normally closed by a second valve structure that is centrifugally actuated to open the discharge ports. The container also includes a coupling structure for releasable engagement with the centrifugal drive structure. The centrifugal force produced when the container is driven in rotation by the drive structure opens the valve structures, and radioactive waste material introduced into the container through the open inlet port is dewatered, and the waste is compacted. The ports are automatically closed by the valves when the container drum is not subjected to centrifugal force such that containment effectiveness is enhanced and exposure of personnel to radioactive materials is minimized. (author)

  1. The Design of Cooling System Model on The AP1000 Containment

    International Nuclear Information System (INIS)

    Daddy Setyawan; Yerri Noer Kartiko; Aryadi Suwono; Ari Darmawan Pasek; Nathanael P Tandian; Efrizon Umar

    2009-01-01

    The policy of national energy leads to the utilization of new energy as nuclear energy, and also contains some efforts to increase reactor safety and optimizing in the design of safety system component such as passive cooling system on reactor containment tank. Because of this, the assessment of safety level to passive safety system needs to be made. To increase the understanding it, the design of cooling system model on containment tank should be done to get safety level on cooling system in the AP1000 containment. To reach the similar model with reality and inexpensive cost, we should make assessment about similarity and dimensionless number. While the heat transfer of air natural circulation and water spray cooling system are a result of gravity approach, we can calculate Grashof modification number and Reynolds number respectively. By this approach, we have a factor of forty for laboratory model. From this model, we hope that we get characteristic correlation to heat transfer on the containment of AP1000 for both air natural circulation and water spray result from gravity. Finally, we can assess the safety level of passive cooling system on the AP1000 containment. (author)

  2. TECHNICAL GUIDANCE DOCUMENT: CONSTRUCTION QUALITY MANAGEMENT FOR REMEDIAL ACTION AND REMEDIAL DESIGN WASTE CONTAINMENT SYSTEMS

    Science.gov (United States)

    This Technical Guidance Document is intended to augment the numerous construction quality control and construction quality assurance (CQC and CQA) documents that are available far materials associated with waste containment systems developed for Superfund site remediation. In ge...

  3. A PROBABILISTIC METHOD FOR ESTIMATING MONITORING POINT DENSITY FOR CONTAINMENT SYSTEM LEAK DETECTION

    Science.gov (United States)

    The use of physical and hydraulic containment systems for the isolation of contaminated ground water and aquifer materials ssociated with hazardous waste sites has increased during the last decade. The existing methodologies for monitoring and evaluating leakage from hazardous w...

  4. Advanced oxidation process-biological system for wastewater containing a recalcitrant pollutant.

    Science.gov (United States)

    Oller, I; Malato, S; Sánchez-Pérez, J A; Maldonado, M I; Gernjak, W; Pérez-Estrada, L A

    2007-01-01

    Two advanced oxidation processes (AOPs), ozonation and photo-Fenton, combined with a pilot aerobic biological reactor at field scale were employed for the treatment of industrial non-biodegradable saline wastewater (TOC around 200 mgL(-1)) containing a biorecalcitrant compound, alpha-methylphenylglycine (MPG), at a concentration of 500 mgL(-1). Ozonation experiments were performed in a 50-L reactor with constant inlet ozone of 21.9 g m(-3). Solar photo-Fenton tests were carried out in a 75-L pilot plant made up of four compound parabolic collector (CPC) units. The catalyst concentration employed in this system was 20 mgL(-1) of Fe2+ and the H2O2 concentration was kept in the range of 200-500mgL(-1). Complete degradation of MPG was attained after 1,020 min of ozone treatment, while only 195 min were required for photo-Fenton. Samples from different stages of both AOPs were taken for Zahn-Wellens biocompatibility tests. Biodegradability enhancement of the industrial saline wastewater was confirmed (>70% biodegradability). Biodegradable compounds generated during the preliminary oxidative processes were biologically mineralised in a 170-L aerobic immobilised biomass reactor (IBR). The global efficiency of both AOP/biological combined systems was 90% removal of an initial TOC of over 500 mgL(-1).

  5. The ALTA cosmic ray experiment electronics system

    International Nuclear Information System (INIS)

    Brouwer, W.; Burris, W.J.; Caron, B.; Hewlett, J.; Holm, L.; Hamilton, A.; McDonald, W.J.; Pinfold, J.L.; Price, P.; Schaapman, J.R.; Sibley, L.; Soluk, R.A.; Wampler, L.J.

    2005-01-01

    Understanding the origin and propagation of high-energy cosmic rays is a fundamental area of astroparticle physics with major unanswered questions. The study of cosmic rays with energy more than 10 14 eV, probed only by ground-based experiments, has been restricted by the low particle flux. The Alberta Large-area Time-coincidence Array (ALTA) uses a sparse array of cosmic ray detection stations located in high schools across a large geographical area to search for non-random high-energy cosmic ray phenomena. Custom-built ALTA electronics is based on a modular board design. Its function is to control the detectors at each ALTA site allowing precise measurements of event timing and energy in the local detectors as well as time synchronization of all of the sites in the array using the global positioning system

  6. Detector control system for an LHC experiment

    CERN Document Server

    Mato, P

    1998-01-01

    The purpose of this document is to provide the user requirements for a detector control system kernel for the LHC experiments following the ESA standard PSS-05 [1]. The first issue will be used to provide the basis for an evaluation of possible development philosophies for a kernel DCS. As such it will cover all the major functionality but only to a level of detail sufficient for such an evaluation to be performed. Many of the requirements are therefore intentionally high level and generic, and are meant to outline the functionality that would be required of the kernel DCS, but not yet to the level of the detail required for implementation. The document is also written in a generic fashion in order not to rule out any implementation technology.1

  7. Thermal-Hydraulic Analysis of the Nuclear Power Engineering Corporation Containment Experiments with GOTHIC

    International Nuclear Information System (INIS)

    Wiles, Lawrence E.; George, Thomas L.

    2003-01-01

    GOTHIC version 7.0 was used to model five tests that were conducted in the Nuclear Power Engineering Corporation facility in Japan. The tests involved steam and helium injection into a preheated, spray-moderated, 1/4-scale model of a pressurized water reactor dry containment. Comparison of GOTHIC predictions to measured data for pressure, vapor temperatures, structure surface temperatures, and helium concentrations provided the opportunity to evaluate methods for modeling gas dispersion, drop heat and mass transfer, and surface heat transfer.The test facility includes three floors. The lower two floors are partitioned into a variety of rooms that simulate the lower regions of the modeled containment. On the upper floor, rooms that simulate the steam generator enclosures and the pressurizer enclosure extend into the dome, which represents about two-thirds of the total volume of the containment.The GOTHIC model was defined with 30 control volumes using a mix of lumped parameter volumes and subdivided volumes that employ a three-dimensional mesh. Each volume included several thermal conductors to model the various structures. More than 100 flow paths were used to model the hydraulic connections.Comparison of predictions to data showed that enhanced grid resolution in the vicinity of the steam-helium release point served to limit dispersion of the steam-helium plume. The data comparisons also suggested that spray effectiveness was reduced by drop impact with the containment wall and by the high drop concentration. The data comparisons further suggested that the presence of condensation, sprays, splashing, and other wetting mechanisms should be considered to obtain a reasonable estimate of the effect of liquid films on the structure surfaces

  8. In vitro irradiation system for radiobiological experiments

    International Nuclear Information System (INIS)

    Tesei, Anna; Zoli, Wainer; D’Errico, Vincenzo; Romeo, Antonino; Parisi, Elisabetta; Polico, Rolando; Sarnelli, Anna; Arienti, Chiara; Menghi, Enrico; Medri, Laura; Gabucci, Elisa; Pignatta, Sara; Falconi, Mirella; Silvestrini, Rosella

    2013-01-01

    Although two-dimensional (2-D) monolayer cell cultures provide important information on basic tumor biology and radiobiology, they are not representative of the complexity of three-dimensional (3-D) solid tumors. In particular, new models reproducing clinical conditions as closely as possible are needed for radiobiological studies to provide information that can be translated from bench to bedside. We developed a novel system for the irradiation, under sterile conditions, of 3-D tumor spheroids, the in vitro model considered as a bridge between the complex architectural organization of in vivo tumors and the very simple one of in vitro monolayer cell cultures. The system exploits the same equipment as that used for patient treatments, without the need for dedicated and highly expensive instruments. To mimic the passage of radiation beams through human tissues before they reach the target tumor mass, 96-multiwell plates containing the multicellular tumor spheroids (MCTS) are inserted into a custom-built phantom made of plexiglass, the material most similar to water, the main component of human tissue. The system was used to irradiate CAEP- and A549-derived MCTS, pre-treated or not with 20 μM cisplatin, with a dose of 20 Gy delivered in one session. We also tested the same treatment schemes on monolayer CAEP and A549 cells. Our preliminary results indicated a significant increment in radiotoxicity 20 days after the end of irradiation in the CAEP spheroids pre-treated with cisplatin compared to those treated with cisplatin or irradiation alone. Conversely, the effect of the radio- chemotherapy combination in A549-derived MCTS was similar to that induced by cisplatin or irradiation alone. Finally, the 20 Gy dose did not affect cell survival in monolayer CAEP and A549 cells, whereas cisplatin or cisplatin plus radiation caused 100% cell death, regardless of the type of cell line used. We set up a system for the irradiation, under sterile conditions, of tumor cells

  9. Sludge Treatment Project Engineered Container Retrieval And Transfer System Preliminary Design Hazard Analysis Supplement 1

    International Nuclear Information System (INIS)

    Franz, G.R.; Meichle, R.H.

    2011-01-01

    This 'What/If' Hazards Analysis addresses hazards affecting the Sludge Treatment Project Engineered Container Retrieval and Transfer System (ECRTS) NPH and external events at the preliminary design stage. In addition, the hazards of the operation sequence steps for the mechanical handling operations in preparation of Sludge Transport and Storage Container (STSC), disconnect STSC and prepare STSC and Sludge Transport System (STS) for shipping are addressed.

  10. Validation of the containment code Sirius: interpretation of an explosion experiment on a scale model

    International Nuclear Information System (INIS)

    Blanchet, Y.; Obry, P.; Louvet, J.; Deshayes, M.; Phalip, C.

    1979-01-01

    The explicit 2-D axisymmetric Langrangian code SIRIUS, developed at the CEA/DRNR, Cadarache, deals with transient compressive flows in deformable primary tanks with more or less complex internal component geometries. This code has been subjected to a two-year intensive validation program on scale model experiments and a number of improvements have been incorporated. This paper presents a recent calculation of one of these experiments using the SIRIUS code, and the comparison with experimental results shows the encouraging possibilities of this Lagrangian code

  11. Double container system for the transport and storage of radioactive materials

    International Nuclear Information System (INIS)

    Popp, F.W.; Pontani, B.; Ernst, E.

    1987-01-01

    The double container system consists of an inner storage container made of steel for the gastight inclusion of the radioactive material to be stored and an outer shielding container which ensures the necessary shielding and mechanical safety in handling and transport. A neutron moderator layer of material containing hydrogen, preferably polyethylene, is present in the annular gap between the outer shielding container and the inner storage container. In order to achieve good shielding with simultaneous very good heat conduction from the inside to the outside, the moderator layer consists of individual polyethylene rings stacked above one another. There is an H profile ring made of heat conducting metal material between each two polyethylene rings. The legs of the H profile ring surround the sides of the two polyethylene rings for fixing it. (orig.) [de

  12. 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

  13. Severe Accident Mitigation through Improvements in Filtered Containment Vent Systems and Containment Cooling Strategies for Water Cooled Reactors. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    2017-05-01

    One of the most important lessons from the accident at the Fukushima Daiichi nuclear power plant is that a reliable containment venting system can be crucial for effective accident management during severe accidents, especially for smaller volume containments in relation to the rated nuclear power. Containment venting can enhance the capability to maintain core cooling and containment integrity as well as reduce uncontrolled radioactive releases to the environment if the venting system has a filtration capacity. In general, a filtered containment vent system increases the flexibility of plant personnel in coping with unforeseen events. This publication provides the overview of the current status of related activities with the goal to share information between Member States on actions, upgrades, and new technologies pertaining to containment cooling and venting.

  14. Computation programs for the thermofluidodynamic transient analysis in the containment system following a LOCA

    International Nuclear Information System (INIS)

    Gorlandi, A.; Mazzini, M.; Oriolo, F.

    1979-01-01

    This works briefly describes the features of the computation codes available at the Istituto di Impianti Nucleari of the Pisa University for the analysis of the thermofluidodynamic transient in the containment system of a nuclear power plant following a LOCA (RELAP 4/MOD.S, COMPARE, FUMO and CONTEMPT-LT/026). More details are contained in the Annex. Particular attention has been devoted to the opportunity to study, through the computation codes, the effects of the sub division of a full pressure containment system

  15. Study on the Behaviors of a Conceptual Passive Containment Cooling System

    Directory of Open Access Journals (Sweden)

    Jianjun Wang

    2014-01-01

    Full Text Available The containment is an ultimate and important barrier to mitigate the consequences after the release of mass and energy during such scenarios as loss of coolant accident (LOCA or main steam line break (MSLB. In this investigation, a passive containment cooling system (PCCS concept is proposed for a large dry concrete containment. The system is composed of series of heat exchangers, long connecting pipes with relatively large diameter, valves, and a water tank, which is located at the top of the system and serves as the final heat sink. The performance of the system is numerically studied in detail under different conditions. In addition, the influences of condensation heat transfer conditions and containment environment temperature conditions are also studied on the behaviors of the system. The results reveal that four distinct operating stages could be experienced as follows: startup stage, single phase quasisteady stage, flashing speed-up transient stage, and flashing dominated quasisteady operating stage. Furthermore, the mechanisms of system behaviors are thus analyzed. Moreover, the feasibility of the system is also discussed to meet the design purpose for the containment integrity requirement. Considering the passive feature and the compactness of the system, the proposed PCCS is promising for the advanced integral type reactor.

  16. A hydroponic system for microgravity plant experiments

    Science.gov (United States)

    Wright, B. D.; Bausch, W. C.; Knott, W. M.

    1988-01-01

    The construction of a permanently manned space station will provide the opportunity to grow plants for weeks or months in orbit for experiments or food production. With this opportunity comes the need for a method to provide plants with a continuous supply of water and nutrients in microgravity. The Capillary Effect Root Environment System (CERES) uses capillary forces to maintain control of circulating plant nutrient solution in the weightless environment of an orbiting spacecraft. The nutrient solution is maintained at a pressure slightly less than the ambient air pressure while it flows on one side of a porous membrane. The root, on the other side of the membrane, is surrounded by a thin film of nutrient solution where it contacts the moist surface of the membrane. The root is provided with water, nutrients and air simultaneously. Air bubbles in the nutrient solution are removed using a hydrophobic/hydrophilic membrane system. A model scaled to the size necessary for flight hardware to test CERES in the space shuttle was constructed.

  17. The detector safety system for LHC experiments

    CERN Document Server

    Schmeling, Sascha; Lüders, S; Morpurgo, Giulio

    2004-01-01

    The Detector Safety System (DSS), currently being developed at CERN under the auspices of the Joint Controls Project (JCOP), will be responsible for assuring the protection of equipment for the four Large Hadron Collider (LHC)**1 experiments. Thus, the DSS will require a high degree of both availability and reliability. After evaluation of various possible solutions, a prototype is being built based on a redundant Siemens PLC**2 front-end, to which the safety- critical part of the DSS task is delegated. This is then supervised by a PVSS**3 SCADA**4 system via an OPC**5 server. The PLC front-end is capable of running autonomously and of automatically taking predefined protective actions whenever required. The supervisory layer provides the operator with a status display and with limited online reconfiguration capabilities. Configuration of the code running in the PLCs will be completely data driven via the contents of a "configuration database." Thus, the DSS can easily adapt to the different and constantly ev...

  18. Synthetic vision systems: operational considerations simulation experiment

    Science.gov (United States)

    Kramer, Lynda J.; Williams, Steven P.; Bailey, Randall E.; Glaab, Louis J.

    2007-04-01

    Synthetic vision is a computer-generated image of the external scene topography that is generated from aircraft attitude, high-precision navigation information, and data of the terrain, obstacles, cultural features, and other required flight information. A synthetic vision system (SVS) enhances this basic functionality with real-time integrity to ensure the validity of the databases, perform obstacle detection and independent navigation accuracy verification, and provide traffic surveillance. Over the last five years, NASA and its industry partners have developed and deployed SVS technologies for commercial, business, and general aviation aircraft which have been shown to provide significant improvements in terrain awareness and reductions in the potential for Controlled-Flight-Into-Terrain incidents / accidents compared to current generation cockpit technologies. It has been hypothesized that SVS displays can greatly improve the safety and operational flexibility of flight in Instrument Meteorological Conditions (IMC) to a level comparable to clear-day Visual Meteorological Conditions (VMC), regardless of actual weather conditions or time of day. An experiment was conducted to evaluate SVS and SVS-related technologies as well as the influence of where the information is provided to the pilot (e.g., on a Head-Up or Head-Down Display) for consideration in defining landing minima based upon aircraft and airport equipage. The "operational considerations" evaluated under this effort included reduced visibility, decision altitudes, and airport equipage requirements, such as approach lighting systems, for SVS-equipped aircraft. Subjective results from the present study suggest that synthetic vision imagery on both head-up and head-down displays may offer benefits in situation awareness; workload; and approach and landing performance in the visibility levels, approach lighting systems, and decision altitudes tested.

  19. Synthetic Vision Systems - Operational Considerations Simulation Experiment

    Science.gov (United States)

    Kramer, Lynda J.; Williams, Steven P.; Bailey, Randall E.; Glaab, Louis J.

    2007-01-01

    Synthetic vision is a computer-generated image of the external scene topography that is generated from aircraft attitude, high-precision navigation information, and data of the terrain, obstacles, cultural features, and other required flight information. A synthetic vision system (SVS) enhances this basic functionality with real-time integrity to ensure the validity of the databases, perform obstacle detection and independent navigation accuracy verification, and provide traffic surveillance. Over the last five years, NASA and its industry partners have developed and deployed SVS technologies for commercial, business, and general aviation aircraft which have been shown to provide significant improvements in terrain awareness and reductions in the potential for Controlled-Flight-Into-Terrain incidents/accidents compared to current generation cockpit technologies. It has been hypothesized that SVS displays can greatly improve the safety and operational flexibility of flight in Instrument Meteorological Conditions (IMC) to a level comparable to clear-day Visual Meteorological Conditions (VMC), regardless of actual weather conditions or time of day. An experiment was conducted to evaluate SVS and SVS-related technologies as well as the influence of where the information is provided to the pilot (e.g., on a Head-Up or Head-Down Display) for consideration in defining landing minima based upon aircraft and airport equipage. The "operational considerations" evaluated under this effort included reduced visibility, decision altitudes, and airport equipage requirements, such as approach lighting systems, for SVS-equipped aircraft. Subjective results from the present study suggest that synthetic vision imagery on both head-up and head-down displays may offer benefits in situation awareness; workload; and approach and landing performance in the visibility levels, approach lighting systems, and decision altitudes tested.

  20. Multilevel Workflow System in the ATLAS Experiment

    International Nuclear Information System (INIS)

    Borodin, M; De, K; Navarro, J Garcia; Golubkov, D; Klimentov, A; Maeno, T; Vaniachine, A

    2015-01-01

    The ATLAS experiment is scaling up Big Data processing for the next LHC run using a multilevel workflow system comprised of many layers. In Big Data processing ATLAS deals with datasets, not individual files. Similarly a task (comprised of many jobs) has become a unit of the ATLAS workflow in distributed computing, with about 0.8M tasks processed per year. In order to manage the diversity of LHC physics (exceeding 35K physics samples per year), the individual data processing tasks are organized into workflows. For example, the Monte Carlo workflow is composed of many steps: generate or configure hard-processes, hadronize signal and minimum-bias (pileup) events, simulate energy deposition in the ATLAS detector, digitize electronics response, simulate triggers, reconstruct data, convert the reconstructed data into ROOT ntuples for physics analysis, etc. Outputs are merged and/or filtered as necessary to optimize the chain. The bi-level workflow manager - ProdSys2 - generates actual workflow tasks and their jobs are executed across more than a hundred distributed computing sites by PanDA - the ATLAS job-level workload management system. On the outer level, the Database Engine for Tasks (DEfT) empowers production managers with templated workflow definitions. On the next level, the Job Execution and Definition Interface (JEDI) is integrated with PanDA to provide dynamic job definition tailored to the sites capabilities. We report on scaling up the production system to accommodate a growing number of requirements from main ATLAS areas: Trigger, Physics and Data Preparation. (paper)

  1. Regulatory Concerns on the In-Containment Water Storage System of the Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Ahn, Hyung-Joon; Lee, Jae-Hun; Bang, Young-Seok; Kim, Hho-Jung

    2002-01-01

    The in-containment water storage system (IWSS) is a newly adopted system in the design of the Korean Next Generation Reactor (KNGR). It consists of the in-containment refueling water storage tank, holdup volume tank, and cavity flooding system (CFS). The IWSS has the function of steam condensation and heat sink for the steam release from the pressurizer and provides cooling water to the safety injection system and containment spray system in an accident condition and to the CFS in a severe accident condition. With the progress of the KNGR design, the Korea Institute of Nuclear Safety has been developing Safety and Regulatory Requirements and Guidances for safety review of the KNGR. In this paper, regarding the IWSS of the KNGR, the major contents of the General Safety Criteria, Specific Safety Requirements, Safety Regulatory Guides, and Safety Review Procedures were introduced, and the safety review items that have to be reviewed in-depth from the regulatory viewpoint were also identified

  2. Applications of containment technologies in Australia for contamination remediation/control: Status and experiences

    International Nuclear Information System (INIS)

    Bouazza, A.; Parker, R.J.

    1997-01-01

    In discussing containment technologies in Australia it is important to understand the factors which influence environmental control of wastes. Although Australia is considered to be an and country, there is only limited reliance on use of groundwater for domestic purposes and this is mainly in rural areas. In many areas, the groundwater is brackish to saline, thus limiting the use of the water. The limited use of groundwater for domestic purposes and the sparse population of Australia combine to produce an environmental regulatory framework very different to North America and Europe. Up until very recently, the approach to disposal of industrial, mining and domestic waste has been based on the principle of open-quotes dilute and disperseclose quotes. However, this attitude has changed, new regulations have been put forward imposing much greater control over the disposal of all forms of waste. This paper provides an overview of the containment technology in Australia as used in certain states with a discussion on the regulatory aspect. It presents examples of some of the innovative techniques that can be considered in the limited Australian regulatory environment

  3. Design and construction of thermal desorption measurement system for tritium contained materials

    International Nuclear Information System (INIS)

    Hara, M.; Hatano, Y.; Calderoni, P.; Shimada, M.

    2014-01-01

    The dual-mode thermal desorption analysis system was designed and built in Idaho National Laboratory (INL) to examine the evolution of the hydrogen isotope gas from materials. The system is equipped with a mass spectrometer for stable hydrogen isotopes and an ionization chamber for tritium components. The performance of the system built was tested with using tritium contained materials. The evolution of tritiated gas species from contaminated materials was measured successfully by using the system. (author)

  4. A study on hydrogen burn due to the operation of containment spray system

    International Nuclear Information System (INIS)

    Park, S.Y.; Kim, D.H.; Jin, Y.; Park, C.K.

    1995-01-01

    The bounding calculation for inflammable gas combustion due to the steam condensation by the operation of the containment spray system was performed. Sensitivity study was performed for two initiating events, station blackout and loss of coolant accident. The parameters for sensitivity study are the condition of cavity, wet or dry, and the timing of operation of the containment spray system. It is shown, based on MAAP4 analyses, that: for dry cavity, auto-ignition burn and hydrogen laden jet burn due to the high temperature in the reactor cavity consumes large amount of burnable gas in the containment and reduces the peak pressure at the global burn by flammability criteria; for wet cavity, large amount of hydrogen and carbon monoxide are generated after dryout of the reactor cavity, but burn is prohibited due to the low gas temperature in the high concentration of the steam. The late operation of the containment spray system condenses the steam rapidly, which results in the global burn at high concentration of burnable gas in the containment. The containment peak pressure from this burn is determined to be high enough to threaten the containment integrity significantly. (author). 3 refs., 3 tabs

  5. Nuclear Power Plant Operating Experience from the IAEA/NEA International Reporting System for Operating Experience 2012-2014

    International Nuclear Information System (INIS)

    2018-03-01

    The International Reporting System for Operating Experience (IRS) is an essential element of the international operating experience feedback system for nuclear power plants. Its fundamental objective is to contribute to improving safety of commercial nuclear power plants which are operated worldwide. IRS reports contain information on events of safety significance with important lessons learned which assist in reducing recurrence of events at other plants. This sixth publication, covering the period 2012 - 2014, follows the structure of the previous editions. It highlights important lessons based on a review of the approximately 240 event reports received from the participating countries over this period.

  6. APPARENT DIGESTIBILTY EXPERIMENT WITH NILE TILAPIA (OREOCHROMIS NILOTICUS FED DIETS CONTAINING CITRULLUS LANATUS SEEDMEAL

    Directory of Open Access Journals (Sweden)

    Wasiu Adeyemi JIMOH

    2015-12-01

    Full Text Available Apparent digestibility coefficients of nutrients in Citrullus lanatus based diets were determined for Nile tilapia (Oreochromis niloticus using AIA as marker or indicator. 150 tilapia fingerlings of average weight 6.12±0.05g were acclimatized for a week, weighed and allotted into five dietary treatments; CTR, DT2, DT3, DT4 and DT5 containing 0, 15, 30, 45 and 60% Citrullus lanatus respectively. The diets were isonitrogenous, isocaloric and isolipidic. Each treatment was replicated three times with ten fish per replicate. Fish were fed 5% body weight on two equal proportions per day. The results from the study indicated that there was no significant variation (p>0.05 in the apparent organic matter and gross energy digestibility coefficients of the diets; that there was significant (p0.05 in the apparent digestibility coefficients of nutrients (protein, energy, lipid and carbohydrates between the diets up to 30% replacement levels for tilapia.

  7. Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

    International Nuclear Information System (INIS)

    Hilliard, R.K.; Postma, A.K.

    1975-01-01

    Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

  8. The development of the thermohydraulic analysis code for the passive containment cooling system: PCCSAC

    International Nuclear Information System (INIS)

    Wang Jianyu; Zhang Shenru; Min Yuanyou

    1994-01-01

    To estimate the performance of the passive containment cooling system (PCCS) of the AC-600 nuclear power plant, the PCCSAC code is developed currently by the jointed efforts between Tsinghua University and NPIC. Different features on the passive behavior of the system and the main components of the containment are considered in the code which is needed by the further AC-600 R and D Program. With a brief description of the AC-600 passive containment cooling system and components, the main thermohydraulic models and numerical scheme used in the PCCSAC code are introduced and the selected results of the verification and the prediction for the performance of the AC-600 passive containment cooling system under LOCA and a steam line break accident are presented to preliminarily demonstrate the applicability and reliability of the PCCSAC model. The current PCCSAC model is conservative and a further 2-D PCCSAC version is under consideration in addition to provide the database for models from some tests associated with the components and systems unique to AC-600 nuclear power plant to meet the requirement of the more realistic modelization for the AC-600 passive containment cooling system. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  10. Arsenic removal using steel manufacturing byproducts as permeable reactive materials in mine tailing containment systems.

    Science.gov (United States)

    Ahn, Joo Sung; Chon, Chul-Min; Moon, Hi-Soo; Kim, Kyoung-Woong

    2003-05-01

    Steel manufacturing byproducts were tested as a means of treating mine tailing leachate with a high As concentration. Byproduct materials can be placed in situ as permeable reactive barriers to control the subsurface release of leachate from tailing containment systems. The tested materials had various compositions of elemental Fe, Fe oxides, Ca-Fe oxides and Ca hydroxides typical of different steel manufacturing processes. Among these materials, evaporation cooler dust (ECD), oxygen gas sludge (OGS), basic oxygen furnace slag (BOFS) and to a lesser degree, electrostatic precipitator dust (EPD) effectively removed both As(V) and As(III) during batch experiments. ECD, OGS and BOFS reduced As concentrations to <0.5mg/l from 25mg/l As(V) or As(III) solution in 72 h, exhibiting higher removal capacities than zero-valent iron. High Ca concentrations and alkaline conditions (pH ca. 12) provided by the dissolution of Ca hydroxides may promote the formation of stable, sparingly soluble Ca-As compounds. When initial pH conditions were adjusted to 4, As reduction was enhanced, probably by adsorption onto iron oxides. The elution rate of retained As from OGS and ECD decreased with treatment time, and increasing the residence time in a permeable barrier strategy would be beneficial for the immobilization of As. When applied to real tailing leachate, ECD was found to be the most efficient barrier material to increase pH and to remove As and dissolved metals.

  11. The GR-value deviation from the additivity rule for irradiated systems containing heterocyclic compounds

    International Nuclear Information System (INIS)

    Nanobashvili, H.M.; Shanidze, G.V.; Khidesheli, G.I.; Panchvidze, M.V.

    1988-01-01

    The investigation of the low temperature radiolysis of binary systems containing heterocyclic compounds has been carried out. In the systems under study the G R -value deviation from the additivity rule is observed due to the energy transfer processes from matrix molecules. It is shown that heterocyclic compounds are good radioprotectors. (author)

  12. Introduction of filtration systems in container nurseries for nonchemical elimination of Phytophthora spp. from irrigation water

    Science.gov (United States)

    Thorsten Ufer; Heinrich Beltz; Thomas Brand; Katrin Kaminski; Ralf Lüttmann; Martin Posner; Stefan Wagner; Sabine Werres; Hans-Peter Wessels

    2006-01-01

    In a 3-year project the elimination of Phytophthora spp. from the recirculation water with different kinds of filtration systems will be tested under commercial conditions in container nurseries. First results indicate that the filtration systems eliminate Phytophthora spp. from the water.

  13. System for recovery of CO2 from flue gases containing SO2

    International Nuclear Information System (INIS)

    Sears, J. T.; Anada, H. R.

    1985-01-01

    An improved system for recovering CO 2 from flue gases containing SO 2 at low CO 2 partial pressure. The system includes the use of K 2 CO 3 as the solvent, regeneration of the solvent, and removal of SO 2 and SO 4

  14. Multilevel Workflow System in the ATLAS Experiment

    CERN Document Server

    Borodin, M; The ATLAS collaboration; Golubkov, D; Klimentov, A; Maeno, T; Vaniachine, A

    2015-01-01

    The ATLAS experiment is scaling up Big Data processing for the next LHC run using a multilevel workflow system comprised of many layers. In Big Data processing ATLAS deals with datasets, not individual files. Similarly a task (comprised of many jobs) has become a unit of the ATLAS workflow in distributed computing, with about 0.8M tasks processed per year. In order to manage the diversity of LHC physics (exceeding 35K physics samples per year), the individual data processing tasks are organized into workflows. For example, the Monte Carlo workflow is composed of many steps: generate or configure hard-processes, hadronize signal and minimum-bias (pileup) events, simulate energy deposition in the ATLAS detector, digitize electronics response, simulate triggers, reconstruct data, convert the reconstructed data into ROOT ntuples for physics analysis, etc. Outputs are merged and/or filtered as necessary to optimize the chain. The bi-level workflow manager - ProdSys2 - generates actual workflow tasks and their jobs...

  15. Proposed Reactor Operating Experience Feedback System Development

    International Nuclear Information System (INIS)

    Ahn, Seung Hoon; Kim, Min Chul; Huh, Chang Wook; Lee, Durk Hun; Bae, Koo Hyun

    2006-01-01

    Most events occurring in nuclear power plants are not individually significant, and prevented from progressing to accident conditions by a series of barriers against core damage and radioactive releases. Significant events, if occur, are almost always a breach of these multiple barriers. As illustrated in the 'Swiss cheese' model, the individual layers of defense or 'cheese slices' have weakness or 'holes.' These weaknesses are inconstant, i.e., the holes are open or close at random. When by chance all the holes are aligned, a hazard causes the significant event of concern. Elements of low significant events, inattention to detail, time or economic pressure, uncorrected poor practices/habits, marginal maintenance and equipment care, etc., make holes in the layers of defense; some elements may make more holes in different layers, incurring more chances to be aligned. An effective reduction of the holes, therefore, is gained through better knowledge or awareness of increasing trends of the event elements, followed by appropriate actions. According to the Swiss cheese metaphor, attention to the Operating Experience (OE) feedback system, as opposed to the individual and to randomness, is drawn from a viewpoint of reactor safety

  16. Wireless sensing system for non-invasive monitoring of attributes of contents in a container

    Science.gov (United States)

    Woodard, Stanley E. (Inventor)

    2010-01-01

    A wireless sensing system monitors the level, temperature, magnetic permeability and electrical dielectric constant of a non-gaseous material in a container. An open-circuit electrical conductor is shaped to form a two-dimensional geometric pattern that can store and transfer electrical and magnetic energy. The conductor resonates in the presence of a time-varying magnetic field to generate a harmonic response. The conductor is mounted in an environmentally-sealed housing. A magnetic field response recorder wirelessly transmits the time-varying magnetic field to power the conductor, and wirelessly detects the harmonic response that is an indication of at least one of level of the material in the container, temperature of the material in the container, magnetic permeability of the material in the container, and dielectric constant of the material in the container.

  17. Experience on Primary System Decommissioning in Jose Cabrera NPP

    Energy Technology Data Exchange (ETDEWEB)

    Paloma Molleda; Leandro Sanchez; David Rodriguez [ENSA, Cantabria (Spain)

    2015-10-15

    Primary System Decommissioning belongs to DCP(Decommissioning and Closure Plan) works and its scope includes: Steam Generator, Pressurizer, Refrigerant Circuit Pump and Primary Circuit Piping. All these dismantling activities were carried out on site, including preliminary steps before their removal (SAS installations, pre decontaminations, cutting and segmentations, segregations, etc.) and delivery to media/low activity nuclear waste disposal site. There are many cutting techniques available in market (most of them proved with positive results) as well as there are many different approaches about how to manage radioactive wastes in decommissioning projects (containers or great components disposal, containers burial, re fusion, etc.). Both issues are linked and, before starting a new project, it might be positive and quite useful to compare and study previous dismantling experiences, especially the lesson learned chapter. Primary System cut with diamond saw has been a challenge target, not only due to the methodology innovation (since until nowadays, the common use of this technology was performed in cutting concrete walls) because it has a huge range of positive aspects that, in our opinion, are attractive (apart from its mentioned versatility, in terms of cutting on site and every type of material)

  18. Experience on Primary System Decommissioning in Jose Cabrera NPP

    International Nuclear Information System (INIS)

    Paloma Molleda; Leandro Sanchez; David Rodriguez

    2015-01-01

    Primary System Decommissioning belongs to DCP(Decommissioning and Closure Plan) works and its scope includes: Steam Generator, Pressurizer, Refrigerant Circuit Pump and Primary Circuit Piping. All these dismantling activities were carried out on site, including preliminary steps before their removal (SAS installations, pre decontaminations, cutting and segmentations, segregations, etc.) and delivery to media/low activity nuclear waste disposal site. There are many cutting techniques available in market (most of them proved with positive results) as well as there are many different approaches about how to manage radioactive wastes in decommissioning projects (containers or great components disposal, containers burial, re fusion, etc.). Both issues are linked and, before starting a new project, it might be positive and quite useful to compare and study previous dismantling experiences, especially the lesson learned chapter. Primary System cut with diamond saw has been a challenge target, not only due to the methodology innovation (since until nowadays, the common use of this technology was performed in cutting concrete walls) because it has a huge range of positive aspects that, in our opinion, are attractive (apart from its mentioned versatility, in terms of cutting on site and every type of material)

  19. The principle and data analysis of online monitoring system of containment leak rate

    International Nuclear Information System (INIS)

    Zhang Chunwei; Yang Yongdeng; Qiao Yu; Liang Bo

    2014-01-01

    The use of online monitoring system of containment leak rate (EPP) in Qinshan 2nd nuclear power plant is introduced. When the containment leak rate reaches the operational limit, the system will automatically alarm and inform the unit operator to take the necessary action. But it is found that the EPP will give a mendacious alarm of 'Containment leak rate abnormity' once in a while during use. The mendacious alarm has an effect on the normal operation of the unit. The reason of the mendacious alarm is analyzed. The data monitored by the EPP are relative hysteretic and the veracity of the flow of compressed air into the containment has a significant influence on the data monitored by the EPP. (authors)

  20. Reliability analysis of the containment spray system of Angra-1 : the injection phase

    International Nuclear Information System (INIS)

    Gibelli, S.M.O.; Oliveira, L.F.S. de.

    1981-12-01

    The system studied is projected to perform two basic functions : to reduce the pressure and temperature in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment after a LOCA (loss of coolant accident), to break the main steam line or the main feed line in the containment and to remove the fission products, mainly the iodine of the containment atmosphere. The spray system was analyzed concerning the probability of non-acomplishment of both functions at the same time; therefore the failure of the components of the chemical aditions subsystem are included in the failure tree shown here. (E.G.) [pt

  1. Systems and methods for enhancing isolation of high-temperature reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per F.

    2017-09-26

    A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.

  2. Identification of Jets Containing b-Hadrons with Recurrent Neural Networks at the ATLAS Experiment

    CERN Multimedia

    CERN. Geneva

    2017-01-01

    A novel b-jet identification algorithm is constructed with a Recurrent Neural Network (RNN) at the ATLAS Experiment. This talk presents the expected performance of the RNN based b-tagging in simulated $t \\bar t$ events. The RNN based b-tagging processes properties of tracks associated to jets which are represented in sequences. In contrast to traditional impact-parameter-based b-tagging algorithms which assume the tracks of jets are independent from each other, RNN based b-tagging can exploit the spatial and kinematic correlations of tracks which are initiated from the same b-hadrons. The neural network nature of the tagging algorithm also allows the flexibility of extending input features to include more track properties than can be effectively used in traditional algorithms.

  3. Identification of Jets Containing $b$-Hadrons with Recurrent Neural Networks at the ATLAS Experiment

    CERN Document Server

    The ATLAS collaboration

    2017-01-01

    A novel $b$-jet identification algorithm is constructed with a Recurrent Neural Network (RNN) at the ATLAS experiment at the CERN Large Hadron Collider. The RNN based $b$-tagging algorithm processes charged particle tracks associated to jets without reliance on secondary vertex finding, and can augment existing secondary-vertex based taggers. In contrast to traditional impact-parameter-based $b$-tagging algorithms which assume that tracks associated to jets are independent from each other, the RNN based $b$-tagging algorithm can exploit the spatial and kinematic correlations between tracks which are initiated from the same $b$-hadrons. This new approach also accommodates an extended set of input variables. This note presents the expected performance of the RNN based $b$-tagging algorithm in simulated $t \\bar t$ events at $\\sqrt{s}=13$ TeV.

  4. On the possible role of thermal radiation in containment thermal–hydraulics experiments by the example of CFD analysis of TOSQAN T114 air–He test

    Energy Technology Data Exchange (ETDEWEB)

    Filippov, A.S.; Grigoryev, S.Yu. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Moscow Institute of Physics and Technology (Russian Federation); Tarasov, O.V. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation)

    2016-12-15

    Highlights: • Neglecting by heat radiation in simulation of containment tests may cause discrepancies. • To show that, heat exchange in T114 air-helium test was analyzed in different ways. • Effect of thermal radiation on local temperature was numerically obtained in air with ∼1% steam content. • Model of gas-structure heat exchange in containment should include heat radiation. - Abstract: One of the experiments of ERCOSAM–SAMARA (E–S) projects (TOSQAN T114) is examined from the viewpoint of the radiative heat transfer (RHT) contribution to the overall heat exchange. E–S projects and T114 test were focused on investigation of light gas stratification in severe accident containment atmosphere and stratification break-up after the activation of mitigation systems. The first from two phases of T114 test is considered during which helium is quasistatically injected into the upper part of the TOSQAN vessel having isothermal walls and initially filled by air. The developing free convection removes most of the heat acquired, but not all. Thus stable local deviations in calculated temperatures were obtained in simulations that were interpreted as the deficiencies of the physical heat-transfer model. The modeling of RHT was included in full CFD simulation that resulted in a better agreement in local temperatures. The results of comparative calculations performed without/with RHT modeling are described in the paper. The RHT model implemented in the used CFD code (ANSYS FLUENT) was tested on known analytical solutions. The RHT contribution in T114 test was also estimated analytically to demonstrate independently that it may be noticeable in this experiment. The same estimations may be valid for stagnant zones of severe accident containment. All that shows the need in further detailing of the role of RHT in gas-structure heat exchange: as for interpretation of some containment tests performed in pressure vessel as for containment modeling.

  5. Proposed development programme for a temporary containment system for alpha active decommissioning

    International Nuclear Information System (INIS)

    Pengelly, M.G.A.; Burnett, R.C.

    1983-06-01

    This report makes a proposal to design, develop and test a containment of modular construction under plutonium active conditions. While this proposal contemplates work with plutonium, the system, when fully developed, has obvious applications wherever a temporary containment of radioactive or toxic materials is required. The fundamental feature of the proposal is that strippable coatings are used to prevent the inner surfaces of the working area from becoming contaminated. It is envisaged that this method of protecting the surfaces will enable the modular containment structure to be disassembled and re-used. (author)

  6. Design and construction of reactor containment systems of the prototype fast breeder reactor MONJU

    International Nuclear Information System (INIS)

    Ikeda, Makinori; Kawata, Koji; Sato, Masaki; Ito, Masashi; Hayashi, Kazutoshi; Kunishima, Shigeru.

    1991-01-01

    The MONJU reactor containment systems consist of a reactor containment vessel, reactor cavity walls and cell liners. The reactor containment vessel is strengthened by ring stiffeners for earthquake stresses. To verify its earthquake-resistant strength, vibration and buckling tests were carried out by using 1/19 scale models. The reactor cavity walls, which form biological shield and support the reactor vessel, are constructed of steel plate frames filled with concrete. The cell liner consists of liner plates and thermal insulation to moderate the effects of sodium spills, and forms a gastight cell to maintain a nitrogen atmosphere. (author)

  7. The effects of age on nuclear power plant containment cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A. [Brookhaven National Lab., Upton, NY (United States); Davis, J. [Science Applications International Corp., New York, NY (United States)

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated.

  8. The effects of age on nuclear power plant containment cooling systems

    International Nuclear Information System (INIS)

    Lofaro, R.; Subudhi, M.; Travis, R.; DiBiasio, A.; Azarm, A.; Davis, J.

    1994-04-01

    A study was performed to assess the effects of aging on the performance and availability of containment cooling systems in US commercial nuclear power plants. This study is part of the Nuclear Plant Aging Research (NPAR) program sponsored by the US Nuclear Regulatory Commission. The objectives of this program are to provide an understanding of the aging process and how it affects plant safety so that it can be properly managed. This is one of a number of studies performed under the NPAR program which provide a technical basis for the identification and evaluation of degradation caused by age. The effects of age were characterized for the containment cooling system by reviewing and analyzing failure data from national databases, as well as plant-specific data. The predominant failure causes and aging mechanisms were identified, along with the components that failed most frequently. Current inspection, surveillance, and monitoring practices were also examined. A containment cooling system unavailability analysis was performed to examine the potential effects of aging by increasing failure rates for selected components. A commonly found containment spray system design and a commonly found fan cooler system design were modeled. Parametric failure rates for those components in each system that could be subject to aging were accounted for in the model to simulate the time-dependent effects of aging degradation, assuming no provisions are made to properly manage it. System unavailability as a function of increasing component failure rates was then calculated

  9. FAUST/CONTAIN

    International Nuclear Information System (INIS)

    Cherdron, W.; Minges, J.; Sauter, H.; Schuetz, W.

    1995-01-01

    The FAUNA facility has been restructured after completion of the sodium fire experiments. It is now serving LWR research, cf. report II on program no. 32.21.02 concerning steam explosions. The CONTAIN code system for computing the thermodynamic, aerosol and radiological phenomena in a containment under severe accident conditions is being developed with a new to fission product release and transport. (orig.)

  10. VEDS-Automated system for inspection of vehicles and containers for explosives and other threats

    International Nuclear Information System (INIS)

    Gozani, T.; Liu, F.; Sivakumar, M.

    2004-01-01

    Many parts of national infrastructures around the world are very vulnerable to terrorist threats in the form of large vehicle bombs. The larger bomb, the larger is the damage and its extent. The number of containers and vehicles crossing land or sea ports of entry is huge. Tough the probability is low, any vehicle may contain a threat. Any system addressing these enormous security tasks should obviously be based on excellent human intelligence to focus the attention on a much smaller number of high-risk containers and vehicles. These containers must then be subjected to a thorough and reliable inspection for the threats.Viable security system must incorporate a credible and effective inspection to achieve its purposes. It should have high performance and be operationally acceptable. This means the system must possess high detection capabilities, low false positive rate, fast response and provide automatic decision eliminating the need for human interpretation. Ancore has developed a range of new inspection devices, which are highly suitable for the above tasks. All the systems are automatic, material specific, high performance for a wide range and type of threats. Some of them are also highly modular, and compact. Some of the systems are fixed, other are relocatable, or fully mobile. The presentation will discuss Ancore's VEDS (Vehicle Explosive Detection System) which detects bulk explosives (expandable also to radiological and nuclear threats)) in marine containers, trucks and cars. The compact and rugged nature of the VEDS sensor makes it suitable for many forms of conveyance: mobile (van mounted), portal, forklift mounted, or mounted on container unloading rig. The physics principles of the system and some recent applications and results will be presented

  11. Leak rates and structural integrity tests for Laguna Verde Nuclear Power Plant primary containment. Regulatory experience

    International Nuclear Information System (INIS)

    Mamani Alegria, Yuri Raul; Salgado Gonzalez, Julio Ricardo

    1996-01-01

    In the Appendix A General Design Criteria for Nuclear Power Plants of the US Code of Federal Regulations title 10 part 50 (10CFR50) is established the Criterion 1 Quality standards and records which requires that structures, systems and components important to safety should be tested to quality standards according with the importance of the safety function to be performed. This regulation has been adopted by the Mexican Regulatory Body (CNSNS) for their nuclear power plants. (author)

  12. Hyaluron Filler Containing Lidocaine on a CPM Basis for Lip Augmentation: Reports from Practical Experience.

    Science.gov (United States)

    Fischer, Tanja C; Sattler, Gerhard; Gauglitz, Gerd G

    2016-06-01

    Lip augmentation with hyaluronic acid fillers is established. As monophasic polydensified hyaluronic acid products with variable density, CPM-HAL1 (Belotero Balance Lidocaine, Merz Aesthetics, Raleigh, NC) and CPM-HAL2 (Belotero Intense Lidocaine, Merz Aesthetics, Raleigh, NC) are qualified for beautification and particularly natural-looking rejuvenation, respectively. The aim of this article was to assess the handling and outcome of lip augmentation using the lidocaine-containing hyaluronic acid fillers, CPM-HAL1 and CPM-HAL2. Data were documented from patients who received lip augmentation by means of beautification and/or rejuvenation using CPM-HAL1 and/or CPM-HAL2. Observation period was 4 months, with assessment of natural outcome, evenness, distribution, fluidity, handling, malleability, tolerability, as well as patient satisfaction and pain. A total of 146 patients from 21 German centers participated. Physicians rated natural outcome and evenness as good or very good for more than 95% of patients. Distribution, fluidity, handling, and malleability were assessed for both fillers as good or very good in more than 91% of patients. At every evaluation point, more than 93% of patients were very or very much satisfied with the product. A total of 125 patients (85.6%) experienced transient injection-related side effects. Pain intensity during the procedure was mild (2.72 ± 1.72 on the 0-10 pain assessment scale) and abated markedly within 30 minutes (0.42 ± 0.57). Lip augmentation with hyaluronic acid fillers produced a long-term cosmetic result. Due to the lidocaine content, procedural pain was low and transient. Accordingly, a high degree of patient satisfaction was achieved that was maintained throughout the observation period. Thieme Medical Publishers 333 Seventh Avenue, New York, NY 10001, USA.

  13. Do entheogen-induced mystical experiences boost the immune system? Psychedelics, peak experiences, and wellness.

    Science.gov (United States)

    Roberts, T B

    1999-01-01

    Daily events that boost the immune system (as indicated by levels of salivary immunoglobulin A), some instances of spontaneous remission, and mystical experiences seem to share a similar cluster of thoughts, feelings, moods, perceptions, and behaviors. Entheogens--psychedelic drugs used in a religious context--can also produce mystical experiences (peak experiences, states of unitive consciousness, intense primary religious experiences) with the same cluster of effects. When this happens, is it also possible that such entheogen-induced mystical experiences strengthen the immune system? Might spontaneous remissions occur more frequently under such conditions? This article advances the so called "Emxis hypothesis"--that entheogen-induced mystical experiences influence the immune system.

  14. Liquid-Liquid Extraction in Systems Containing Butanol and Ionic Liquids – A Review

    Directory of Open Access Journals (Sweden)

    Kubiczek Artur

    2017-03-01

    Full Text Available Room-temperature ionic liquids (RTILs are a moderately new class of liquid substances that are characterized by a great variety of possible anion-cation combinations giving each of them different properties. For this reason, they have been termed as designer solvents and, as such, they are particularly promising for liquid-liquid extraction, which has been quite intensely studied over the last decade. This paper concentrates on the recent liquid-liquid extraction studies involving ionic liquids, yet focusing strictly on the separation of n-butanol from model aqueous solutions. Such research is undertaken mainly with the intention of facilitating biological butanol production, which is usually carried out through the ABE fermentation process. So far, various sorts of RTILs have been tested for this purpose while mostly ternary liquid-liquid systems have been investigated. The industrial design of liquid-liquid extraction requires prior knowledge of the state of thermodynamic equilibrium and its relation to the process parameters. Such knowledge can be obtained by performing a series of extraction experiments and employing a certain mathematical model to approximate the equilibrium. There are at least a few models available but this paper concentrates primarily on the NRTL equation, which has proven to be one of the most accurate tools for correlating experimental equilibrium data. Thus, all the presented studies have been selected based on the accepted modeling method. The reader is also shown how the NRTL equation can be used to model liquid-liquid systems containing more than three components as it has been the authors’ recent area of expertise.

  15. Image acquisition, transmission and assignment in 60Co container inspection system

    International Nuclear Information System (INIS)

    Wu Zhifang; Zhou Liye; Liu Ximing; Wang Liqiang

    1999-01-01

    The author describes the data acquisition mode and image reconstruction method in 60 Co container inspection system, analyzes the relationship between line pick period and geometry distortion, makes clear the demand to data transmitting rate. It discusses several data communication methods, draws up a plan for network, realizes automatic direction and reasonable assignment of data in the system, cooperation of multi-computer and parallel processing, thus greatly improves the systems inspection efficiency

  16. Hippo Experiment Data Access and Subseting System

    Science.gov (United States)

    Krassovski, M.; Hook, L.; Boden, T.

    2014-12-01

    HIAPER Pole-to-Pole Observations (HIPPO) was an NSF- and NOAA-funded, multi-year global airborne research project to survey the latitudinal and vertical distribution of greenhouse and related gases, and aerosols. Project scientists and support staff flew five month-long missions over the Pacific Basin on the NSF/NCAR Gulfstream V, High-performance Instrumented Airborne Platform for Environmental Research (HIAPER) aircraft between January 2009 and September 2011, spread throughout the annual cycle, from the surface to 14 km in altitude, and from 87°N to 67°S. Data from the HIPPO study of greenhouse gases and aerosols are now available to the atmospheric research community and the public. This comprehensive dataset provides the first high-resolution vertically resolved measurements of over 90 unique atmospheric species from nearly pole-to-pole over the Pacific Ocean across all seasons. The suite of atmospheric trace gases and aerosols is pertinent to understanding the carbon cycle and challenging global climate models. This dataset will provide opportunities for research across a broad spectrum of Earth sciences, including those analyzing the evolution in time and space of the greenhouse gases that affect global climate. The Carbon Dioxide Information Analysis Center (CDIAC) at Oak Ridge National Laboratory (ORNL) provides data management support for the HIPPO experiment including long-term data storage and dissemination. CDIAC has developed a relational database to house HIPPO merged 10-second meteorology, atmospheric chemistry, and aerosol data. This data set provides measurements from all Missions, 1 through 5, that took place from January of 2009 to September 2011. This presentation introduces newly build database and web interface, reflects the present state and functionality of the HIPPO Database and Exploration System as well as future plans for expansion and inclusion of combined discrete flask and GC sample GHG, Halocarbon, and hydrocarbon data.

  17. Experiments with a double solenoid system

    Energy Technology Data Exchange (ETDEWEB)

    Pampa Condori, R.; Lichtenthaeler Filho, R.; Faria, P.N. de; Lepine-Szily, A.; Mendes Junior, D.R.; Pires, K.C.C.; Assuncao, M.; Scarduelli, V.B.; Leistenschneider, E.; Morais, M.C.; Shorto, J.M.B.; Gasques, L. [Universidade de Sao Paulo (IF/USP), SP (Brazil). Inst. de Fisica

    2012-07-01

    Full text: RIBRAS [1] is presently the only experimental equipment in South America capable of producing secondary beams of rare isotopes. It consists of two superconducting solenoids, installed in one of the beam lines of the 8 MV Pelletron Tandem accelerator of the University of Sao Paulo. The exotic nuclei are produced in the collision between the primary beam of the Pelletron Accelerator and the primary target. The secondary beam is selected by the in-flight technique and is usually contaminated with particles coming from scattering and reactions in the primary target such as {sup 7}Li, alpha and other light particles as protons, deuterons and tritons. Solenoids are selectors with large acceptance and the double solenoid system provides ways to improve the quality of the secondary beam by using a degrador in the midst of the two solenoids. The main contamination of the secondary beam comes from {sup 7}Li{sup 2+} particles coming from the primary beam. A degrador placed between the two solenoids is able to separate those particles from the {sup 6}He beam providing an additional charge exchange {sup 7}Li{sup 2+-→}3{sup +}. In addition, the differential energy loss in the degrador provides further selection of the light particles as protons, deuterons, tritons and and alpha particles by the second solenoid. Here we present the results of the first experiment performed at RIBRAS using both solenoids. A pure {sup 6}He beam was produced and the reaction {sup 6}He+p was measured using a thick CH{sub 2} target. 1. R. Lichtenthaeler et al., Eur. Phys. J. A 25,s01,733 (2005) and Nucl. Phys. News 15, 25 (2005). (author)

  18. The development and testing of a modular containment system under plutonium active conditions

    International Nuclear Information System (INIS)

    Sanders, M.J.; Pengelly, M.G.A.

    1984-05-01

    A Modular Containment System has been designed, constructed and tested under plutonium active conditions at AEE Winfrith. The unit consists of a portable self-contained pressurised suit area, complete with shower entry tunnel and ventilation plant which can be assembled to enclose active plant to enable active operations to be carried out safely by operators dressed in standard pressurised suits. A fundamental feature of the system is the use of strippable coatings which are used to treat the interior surfaces prior to active operations to prevent permanent contamination of the structure. Details of construction are given together with results of trials. Whilst this report describes work with plutonium, the system has clear applications wherever temporary containment of radioactive or toxic materials is needed. (U.K.)

  19. Large-scale tests of aqueous scrubber systems for LMFBR vented containment

    International Nuclear Information System (INIS)

    McCormack, J.D.; Hilliard, R.K.; Postma, A.K.

    1980-01-01

    Six large-scale air cleaning tests performed in the Containment Systems Test Facility (CSTF) are described. The test conditions simulated those postulated for hypothetical accidents in an LMFBR involving containment venting to control hydrogen concentration and containment overpressure. Sodium aerosols were generated by continously spraying sodium into air and adding steam and/or carbon dioxide to create the desired Na 2 O 2 , Na 2 CO 3 or NaOH aerosol. Two air cleaning systems were tested: (a) spray quench chamber, educator venturi scrubber and high efficiency fibrous scrubber in series; and (b) the same except with the spray quench chamber eliminated. The gas flow rates ranged up to 0.8 m 3 /s (1700 acfm) at temperatures to 313 0 C (600 0 F). Quantities of aerosol removed from the gas stream ranged up to 700 kg per test. The systems performed very satisfactorily with overall aerosol mass removal efficiencies exceeding 99.9% in each test

  20. Analysis of large scale tests for AP-600 passive containment cooling system

    International Nuclear Information System (INIS)

    Sha, W.T.; Chien, T.H.; Sun, J.G.; Chao, B.T.

    1997-01-01

    One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 hours without action by the reactor operator. During a design-basis accident, i.e., either a loss-of-coolant or a main steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annual space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-ID code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single phase flow, transport equations for the κ-ε two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-ID results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized

  1. Mechanical properties and biocompatibility in alloy Ti-Ta system containing oxygen

    International Nuclear Information System (INIS)

    Ruiz, S.L.M.; Grandini, C.R.; Claro, A.P.R.A.

    2010-01-01

    Due to the excellent properties such as corrosion resistance, good mechanical strength/density, good performance at high temperatures, Ti is very useful in the chemical industry and aerospace. Currently, their use has expanded to the field of biomaterials, due to its excellent biocompatibility and reduced elasticity modulus, favouring the production of orthopaedic and dental prostheses. Promising alloys are the Ti-Ta system and researches have been directed to describe and understand the behavior of this system. In this paper, samples of Ti-Ta alloys containing 8 and 16% (wt%) containing interstitial oxygen were prepared and characterized by density, xray diffraction, hardness, elasticity modulus measurements and in vitro cytotoxicity tests. (author)

  2. 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)

  3. Analysis of L test series of ACE (Advanced Containment Experiments) project with modified corcon UW code

    International Nuclear Information System (INIS)

    Laguna Velasco, H.

    1994-01-01

    A series of experimental tests (so call L, Large scale) have been performance under sponsored of many research institutions around the world and management by Electric Power Research Institute at U.S.A. The goal of these tests is to analyze the phenomena of core-concrete interaction at the same conditions as severe accident in light water nuclear reactor. Results of these tests provides experimental data about thermohydraulic phenomenon and aerosol and fission products release. With these results, improves many codes that already have been developed to simulate core-concrete interaction during severe accident ; in case of CORCON.UW code is a improved version developed in University of Wisconsin at CORCON MOD 2. Scope of this work is shown results obtained from CORCON.UW improved. The improves consist of add data about BaSiO 3 , Ba 2 SiO 4 , BaZrO 3 , SrSiO 4 and SrZrO 3 , append Kutateladze's heat transfer correlation, and finally make more efficient the resolution of energy equations system through use a better algorithm. The results obtained by this improved code to the downward power and H 2 , H 2 O, CO and CO 2 release are agree with experimental results, and also it saved 40% of C.P.U. consumption during execution, due improve of energy equation system. Conclusions are, the increase of thermodynamics data in CORCON.UW produce a well results comparative with experimental results and update heat transfer correlations and algorithm brings a versatile code and reliable results. (Author)

  4. Development of aerosol decontamination factor evaluation method for filtered containment venting system

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae bong; Kim, Sung Il; Jung, Jaehoon; Ha, Kwang Soon; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Fission products would be released from molten corium pools which are relocated into the lower plenum of reactor pressure vessel, on the concrete pit and in the core catcher. In addition, steam, hydrogen and noncondensable gases such as CO and CO2 are generated during the core damage progression due to loss of coolant and the molten core-concrete interaction. Consequently, the pressure inside the containment could be increased continuously. Filtered containment venting is one action to prevent an uncontrolled release of radioactive fission products caused by an overpressure failure of the containment. After the Fukushima-Daiichi accident which was demonstrated the containment failure, many countries to consider the implementation of filtered containment venting system(FCVS) on nuclear power plant where these are not currently applied. In general evaluation for FCVS is conducted to determine decontamination factor on several conditions (aerosol diameter, submergence depth, water temperature, gas flow, steam flow rate, pressure, operating time,...). It is essential to quantify the mass concentration before and after FCVS for decontamination factor. This paper presents the development of the evaluation facility for filtered containment venting system at KAERI and an experimental investigation for aerosol removal performance. Decontamination factor for the FCVS is determined by filter measurement. The result of the aerosol size distribution measurement shows the aerosol removal performance by an aerosol size.

  5. Distributed Containment Control for Multiple Unknown Second-Order Nonlinear Systems With Application to Networked Lagrangian Systems.

    Science.gov (United States)

    Mei, Jie; Ren, Wei; Li, Bing; Ma, Guangfu

    2015-09-01

    In this paper, we consider the distributed containment control problem for multiagent systems with unknown nonlinear dynamics. More specifically, we focus on multiple second-order nonlinear systems and networked Lagrangian systems. We first study the distributed containment control problem for multiple second-order nonlinear systems with multiple dynamic leaders in the presence of unknown nonlinearities and external disturbances under a general directed graph that characterizes the interaction among the leaders and the followers. A distributed adaptive control algorithm with an adaptive gain design based on the approximation capability of neural networks is proposed. We present a necessary and sufficient condition on the directed graph such that the containment error can be reduced as small as desired. As a byproduct, the leaderless consensus problem is solved with asymptotical convergence. Because relative velocity measurements between neighbors are generally more difficult to obtain than relative position measurements, we then propose a distributed containment control algorithm without using neighbors' velocity information. A two-step Lyapunov-based method is used to study the convergence of the closed-loop system. Next, we apply the ideas to deal with the containment control problem for networked unknown Lagrangian systems under a general directed graph. All the proposed algorithms are distributed and can be implemented using only local measurements in the absence of communication. Finally, simulation examples are provided to show the effectiveness of the proposed control algorithms.

  6. Hydrodynamic calculation of a filter washing in liquids type used in containment venting systems

    International Nuclear Information System (INIS)

    Reyes G, A. A.; Sainz M, E.; Ortiz V, J.

    2015-09-01

    From the nuclear accident of Chernobyl, the European nuclear power plants have chosen to install filters on the venting pipes of the containment, whose function is to help to mitigate the consequences of a severe accident, by controlled depressurization of the containment passively through a filtered venting of the containment system. These systems are designed to relieve the internal pressure of the containment by means of the deliberate opening of pressure relief devices, either a valve or rupture disc during a severe accident and be channeled to the filter unit. In this paper the hydraulic response of a filter system of gases washing by liquid is evaluated, due to this information is necessary to estimate the effect that has the pressure increase of the contention on the discharge capacity of the venting pipes. By simulation of computational of fluid dynamics with the programs: CAELINUX-2014 and OpenFOAM, the hydrodynamic characteristics of the Multi Venturi System for gases washing from the containment, which could be included in the general model of the venting pipe, were obtained. Representative models of the Venturi tubes of each concentric area that forming the washing system were generated; and using parametric calculations the average mass flow rate established through each venturi, depending on its size and depth in which it is located inside the tank was estimated. Also, the pressure and mass flow rate required to activate each concentric area depending on the pressure and mass load from the containment were calculated, to estimate the maximum flow that is established through the filter. Finally, the velocity profiles and the characteristic pressure at which each area operates as well as the pressure drop of local and global discharge also were calculated. (Author)

  7. An assessment of Class-9 (core-melt) accidents for PWR dry-containment systems

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Saito, M.

    1981-01-01

    The phenomenology of core-melt accidents in dry containments was examined for the purpose of identifying the margins of safety in such Class-9 situations. The scale (geometry) effects appear to crucially limit the extent (severity) of steam explosions. This together with the established reduced explosivity of the corium-A/water system, and the inherently high capability of dry containments (redinforced concrete, and shields in some cases, seismic design etc.) lead to the conclusion that failure due to steam explosions may be considered essentially incredible. These premixture scaling considerations also impact ultimate debris disposition and coolability and need additional development. A water-flooded reactor cavity would have beneficial effects in limiting (but not necessarily eliminating) melt-concrete interactions. Independently of the initial degree of quenching and/or scale of fragmentation, mechanisms exist that drive the system towards ultimate stability (coolability). Additional studies, with intermediate-scale prototypic materials are recommended to better explore these mechanisms. Containment heat removal systems must provide the crucial capability of mitigating such accidents. Passive systems should be explored and assessed against currently available and/or improved active systems taking into account the rather loose time constraints required for activation. It appears that containment margins for accommodating the hydrogen problem are limited. This problem appears to stand out not only in terms of potential consequences but also in terms of lack of any readily available and clear cut solutions at this time. (orig.)

  8. Containers and systems for the measurement of radioactive gases and related methods

    Science.gov (United States)

    Mann, Nicholas R; Watrous, Matthew G; Oertel, Christopher P; McGrath, Christopher A

    2017-06-20

    Containers for a fluid sample containing a radionuclide for measurement of radiation from the radionuclide include an outer shell having one or more ports between an interior and an exterior of the outer shell, and an inner shell secured to the outer shell. The inner shell includes a detector receptacle sized for at least partial insertion into the outer shell. The inner shell and outer shell together at least partially define a fluid sample space. The outer shell and inner shell are configured for maintaining an operating pressure within the fluid sample space of at least about 1000 psi. Systems for measuring radioactivity in a fluid include such a container and a radiation detector received at least partially within the detector receptacle. Methods of measuring radioactivity in a fluid sample include maintaining a pressure of a fluid sample within a Marinelli-type container at least at about 1000 psi.

  9. Influence of geometrical and thermal hydraulic parameters on the short term containment system response

    International Nuclear Information System (INIS)

    Krishna Chandran, R.; Ali, Seik Mansoor; Balasubramaniyan, V.

    2014-01-01

    This paper discusses the effect of a number of geometrical and thermal hydraulic parameters on the containment peak pressure following a simulated LOCA. The numerical studies are carried out using an inhouse containment thermal hydraulics program called 'THYCON' with focus only on the short term transient response. In order to highlight the effect of above variables, a geometrically scaled (1:270) model of a typical 220 MWe Indian PHWR containment is considered. The discussions in this paper are limited to explaining the influence of individual parameters by comparing with a base case value. It is essential to mention that the results presented here are not general and should be taken as indicative only. Nevertheless, these numerical studies give insight into short term containment response that would be useful to both the system designer as well as the regulator. (author)

  10. Chapter 8: Exponential experiments on graphite moderated lattices fuelled by natural uranium tubes containing cylindrical graphite cores

    International Nuclear Information System (INIS)

    McCulloch, D.B.; Hoskins, T.A.

    1963-01-01

    Experiments have been carried out using a fuel element comprising a 2.75 in. o.d./2.40 in. i.d. natural uranium tube containing a graphite core of diameter 2.0 in. Values of material buckling and migration area asymmetry for lattices at 7 in., 8 in. and 8/2 in. pitch have been obtained, and correlated with the theory of Syrett (1961) to derive an effective resonance integral for the cored element. By comparison with the resonance integral for the same fuel tube without a core, a value for the constant 'γ' of the theory of Stace (1959) is obtained. (author)

  11. Simulation of KAEVER experiments on aerosol behavior in a nuclear power plant containment at accident conditions with the ASTEC code

    International Nuclear Information System (INIS)

    Kljenak, I.; Mavko, B.

    2006-01-01

    Experiments on aerosol behaviour in saturated and non-saturated atmosphere, which were performed in the KAEVER experimental facility, were simulated with the severe accident computer code ASTEC CPA V1.2. The specific purpose of the work was to assess the capability of the code to model aerosol condensation and deposition in the containment of a light-water-reactor nuclear power plant at severe accident conditions, if the atmosphere saturation conditions are simulated adequately. Five different tests were first simulated with boundary conditions, obtained from the experiments. In all five tests, a non-saturated atmosphere was simulated, although, in four tests, the atmosphere was allegedly saturated. The simulations were repeated with modified boundary conditions, to obtain a saturated atmosphere in all tests. Results of dry and wet aerosol concentrations in the test vessel atmosphere for both sets of simulations are compared to experimental results. (author)

  12. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2008-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  13. Design of reactor containment systems for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. It is a revision of the Safety Guide on Design of the Reactor Containment Systems in Nuclear Power Plants (Safety Series No. 50-Sg-D1) issued in 1985 and supplements the Safety Requirements publication on Safety of Nuclear Power Plants: Design. The present Safety Guide was prepared on the basis of a systematic review of the relevant publications, including the Safety of Nuclear Power Plants: Design, the Safety fundamentals publication on The Safety of Nuclear Installations, Safety Guides, INSAG Reports, a Technical Report and other publications covering the safety of nuclear power plants. 1.2. The confinement of radioactive material in a nuclear plant, including the control of discharges and the minimization of releases, is a fundamental safety function to be ensured in normal operational modes, for anticipated operational occurrences, in design basis accidents and, to the extent practicable, in selected beyond design basis accidents. In accordance with the concept of defence in depth, this fundamental safety function is achieved by means of several barriers and levels of defence. In most designs, the third and fourth levels of defence are achieved mainly by means of a strong structure enveloping the nuclear reactor. This structure is called the 'containment structure' or simply the 'containment'. This definition also applies to double wall containments. 1.3. The containment structure also protects the reactor against external events and provides radiation shielding in operational states and accident conditions. The containment structure and its associated systems with the functions of isolation, energy management, and control of radionuclides and combustible gases are referred to as the containment systems

  14. Effect for Recovery of the Containment Spray System to the Release of Cesium

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    In the perspective of the amount of Cs-137, the mass of Cs-137 correspondent with the 100TBq is calculated as 32g. However, during the severe accident, if the containment has been failed, it is generally expected that the mass of Cs-137 released to the environment is more than 1kg for most accident sequences So, the review and improvement of the PSA model in order to reduce containment failure frequency should be needed. Actually, the current PSA model is known to be constructed by the conservative assumptions, especially in the view point of Level 2 PSA model. Therefore, it is necessary to find this conservatism and to improve the Model using the reasonable assumptions. All of the domestic operating nuclear power plants are required to prepare the Accident Management Plan within 3 years and this Accident Management Plan should have to meet the New Safety Goal including the requirement that the sum of the accident frequency that the release of the radioactive nuclide Cs-137 to the environment exceeds the 100TBq should be less than 1.0E-6/RY. The containment spray system is the only facility that mitigates the containment over-pressurization in the operating nuclear power plants, such as Westinghouse type or OPR1000 type. In this study, the effects of the containment spray system recovery on the amount of Cesium released to the environment were analyzed. If the recovery of the containment spray system can be applied to the PSA model, it is expected that the containment failure frequency and also the amount of cesium released to the environment can be greatly reduced.

  15. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    International Nuclear Information System (INIS)

    Ha, Huiun; Suh, Jungsoo

    2016-01-01

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR

  16. Assessment of the effect of nitrogen gas on passive containment cooling system performance

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Huiun; Suh, Jungsoo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    As a part of the passive containment cooling system (PCCS) of Innovative PWR development project, we have been investigating the effect of the nitrogen gas released from safety injection tank (SIT) on PCCS performance. With the design characteristics of APR1400 and conceptual design of PCCS, we developed a GOTHIC model of the APR1400 containment with PCCS. The calculation model is described herein, and representative results from the calculation are presented as well. The results of the present work will be used for the design of PCCS. APR1400 GOTHIC model was developed for assessment on the effect of SIT nitrogen gas on passive containment cooling system performance. Calculation results confirmed that influence of nitrogen gas release is negligible; however, further studies should be performed to confirm effect of non-condensable gas on the final performance of PCCS. These insights are important for developing the PCCS of Innovative PWR.

  17. A Wireless Monitoring System for Cracks on the Surface of Reactor Containment Buildings.

    Science.gov (United States)

    Zhou, Jianguo; Xu, Yaming; Zhang, Tao

    2016-06-14

    Structural health monitoring with wireless sensor networks has been increasingly popular in recent years because of the convenience. In this paper, a real-time monitoring system for cracks on the surface of reactor containment buildings is presented. Customized wireless sensor networks platforms are designed and implemented with sensors especially for crack monitoring, which include crackmeters and temperature detectors. Software protocols like route discovery, time synchronization and data transfer are developed to satisfy the requirements of the monitoring system and stay simple at the same time. Simulation tests have been made to evaluate the performance of the system before full scale deployment. The real-life deployment of the crack monitoring system is carried out on the surface of reactor containment building in Daya Bay Nuclear Power Station during the in-service pressure test with 30 wireless sensor nodes.

  18. System for routine testing of self-contained and airline breathing equipment

    Energy Technology Data Exchange (ETDEWEB)

    McDermott, H.J.; Hermens, G.A.

    1980-07-01

    A system for routine testing of self-contained and airline breathing equipment, developed by Shell Oil Co., for testing breathing equipment at one of its refineries, consists of an 80 psig air supply for airline respirators; a 500-2100 psig air supply for self-contained units; a regulator test system which uses a mannequin head that simulates human inhalation and which tests the ability of the regulator to keep the mask interior at the correct positive pressure; and an exhalation valve test system which identifies a leaky or sticking valve. The testing system has been in use for about 30 mo and has led to increased acceptance of respiratory protective equipment by workers.

  19. Sensitivity evaluation of human factors for reliability of the containment spray system

    International Nuclear Information System (INIS)

    Tsujimura, Yasuhiro; Suzuki, Eiji

    1988-01-01

    Evaluation of the human reliability is one of the most difficult problems that deal with the safety and reliability of large systems, especially of the Engineered Safety Features (ESF) of the nuclear power plant. Influences of human factors on the reliability of the Containment Spray System in the ESF were estimated by using the FTA method in this paper. As a result, the adequacy of the system structure and the effects of human factors on variations of the design of the system structure were explained. (author)

  20. The research on rectification and amplification of the image in mobile large container inspection system

    International Nuclear Information System (INIS)

    Jin Hui; Cheng Jianping; Chen Zhiqiang; Zhang Li

    2001-01-01

    The author introduces a geometrical rectification algorithm of the image in mobile large container inspection system. The comparison and discussion of the image before and after the rectification have been given. Amplification algorithms of the images are discussed. With all the algorithms, the quality of the images has been improved

  1. A method of providing a barrier in a fracture-containing system

    DEFF Research Database (Denmark)

    2014-01-01

    The present invention relates to a method of providing a barrier in a fracture-containing system, comprising: i) Providing a treatment fluid comprising: a) a base fluid; b) an elastomeric material, wherein said elastomeric material comprises at least one polymer capable of crosslinking into an el......The present invention relates to a method of providing a barrier in a fracture-containing system, comprising: i) Providing a treatment fluid comprising: a) a base fluid; b) an elastomeric material, wherein said elastomeric material comprises at least one polymer capable of crosslinking...... into an elastomer, and c) at least one crosslinking agent; ii) Placing the treatment fluid in a fracture-containing system; iii) Allowing the elastomeric material to crosslink with itself to form a barrier in said fracture-containing system; wherein the elastomeric material and/or the crosslinking agent...... are of neutral buoyancy with regard to the base fluid. The invention is contemplated to having utility not only in the oil-drilling industry but also in the plugging of fractures in sewer drains, pipelines etc....

  2. 21 CFR 864.9900 - Cord blood processing system and storage container.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Cord blood processing system and storage container. 864.9900 Section 864.9900 Food and Drugs FOOD AND DRUG ADMINISTRATION, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL DEVICES HEMATOLOGY AND PATHOLOGY DEVICES Products Used In Establishments That...

  3. EPIC 219217635: A Doubly Eclipsing Quadruple System Containing an Evolved Binary

    DEFF Research Database (Denmark)

    Borkovits, T.; Albrecht, S.; Rappaport, S.

    2018-01-01

    We have discovered a doubly eclipsing, bound, quadruple star system in the field of K2 Campaign 7. EPIC 219217635 is a stellar image with Kp = 12.7 that contains an eclipsing binary (‘EB’) with PA = 3.59470 d and a second EB with PB = 0.61825 d. We have obtained followup radial-velocity (‘RV’) sp...

  4. Evaluation of Flammable Gas Monitoring and Ventilation System Alternatives for Double-Contained Receiver Tanks

    International Nuclear Information System (INIS)

    GUSTAVSON, R.D.

    1999-01-01

    This study identifies possible flammable gas monitoring and ventilation system alternatives to ensure adequate removal of flammable gases from the Double-Contained Receiver Tank (DCRT) primary tanks during temporary storage of small amounts of waste. The study evaluates and compares these alternatives to support closure of the Flammable Gas Unreviewed Safety Question (USQ TF-96-04330)

  5. Results from modernization of the containment localization systems of the Kozloduy NPP units 3 and 4

    International Nuclear Information System (INIS)

    Sabinov, S.

    2005-01-01

    The improvements of the Accident Localization System (SLA) of units 3 and 4 systematically implemented by Kozloduy NPP shows an important direction for increasing the safety of NPP with WWER-440/V-230 containments. During the years Kozloduy NPP implemented a large scope of activities aimed in full resolution of all generic shortcomings identified in the original design of these containments. These activities allowed already in 2002 to justify that integrity of the containment is assured for all postulated events and the radiological consequences for all DBAs and BDBAs without core degradation are within the regulatory limits. The last phase of this modernization was oriented toward achieving the same goals in case of severe accidents by installation of systems for avoiding long term pressurization of the SG compartments and by installation of a system for keeping of negative pressure (slight vacuum) during the late phases of development of the accidents thus minimizing significantly the uncontrolled radioactive releases from the containment and assured controlled purified release of radioactivity to environment, and for elimination of conditions for H2 deflagration within the localization system. This paper summarizes the results of the whole modernization process with an emphasis of the implementation of the latest phase successfully completed by Kozloduy NPP in first quarter of 2005 which allowed the plant to demonstrate remarkable mitigation capability for a comprehensive set of very low probability severe accidents scenarios in line with the approach now being promoted for the modern design NPPs

  6. Luminescence properties of Si-containing porous matrix–PbS nanoparticle systems

    International Nuclear Information System (INIS)

    Tarasov, S. A.; Aleksandrova, O. A.; Lamkin, I. A.; Maksimov, A. I.; Maraeva, E. V.; Mikhailov, I. I.; Moshnikov, V. A.; Musikhin, S. F.; Nalimova, S. S.; Permyakov, N. V.; Spivak, Yu. M.; Travkin, P. G.

    2015-01-01

    The luminescence properties of systems that contain lead-sulfide nanoparticles deposited onto substrates fabricated from porous silicon, oxidized porous silicon, and porous (tin-oxide)–(silicon-oxide) layers are studied. It is shown that the structure and composition of the matrix induce a strong effect on the luminescence spectra of colloidal quantum dots, defining their emission wavelength

  7. Experimental study and phase equilibrium modeling of systems containing acid gas and glycol

    DEFF Research Database (Denmark)

    Afzal, Waheed; Breil, Martin P.; Tsivintzelis, Ioannis

    2012-01-01

    In this work, we study phase equilibria of systems containing acid gases and glycols. The acid gases include carbonyl sulfide (COS), hydrogen sulfide (H2S), and carbon dioxide (CO2) while glycols include monoethylene glycol (MEG), diethylene glycol (DEG), and triethylene glycol (TEG). A brief lit...

  8. Experience on a BWR plant diagnosis system

    International Nuclear Information System (INIS)

    Tanabe, A.; Kawai, K.; Hashimoto, Y.

    1981-01-01

    It is important to watch plant dynamics and equipment condition for avoiding a big transient or avoiding damage to a system by equipment failure. After the TMI accident the necessity of a diagnosis system has been recognized and such development activities have become of primary importance in many organizations. A diagnosis system has two kinds of function. One is the early detection of an anomaly before detection by a conventional instrumentation system. The other is appropriate instruction after alarm or scram has occurred. The authors have been developing the former system by a noise analysis technique and a feasibility study has been undertaken in recent years as a joint research programme of several electric power companies and the Toshiba Corporation. A prototype diagnosis system has been installed on a BWR plant in Japan. This diagnosis system concerns reactor core, jet pumps and three main control systems. Many data from normal operation have been accumulated using this system and a variation pattern of normal noise data is clarified. On this basis, anomally detection criteria have been determined using statistical decision theory. It is confirmed that this system performance is satisfactory, and that the system will be of great use for surveillance of core and control systems without artificial disturbances. (author)

  9. CLASSIFICATION OF THE MGR NON-FUEL COMPONENTS DISPOSAL CONTAINER SYSTEM

    International Nuclear Information System (INIS)

    J.A. Ziegler

    1999-01-01

    The purpose of this analysis is to document the Quality Assurance (QA) classification of the Monitored Geologic Repository (MGR) non-fuel components disposal container system structures, systems and components (SSCs) performed by the MGR Safety Assurance Department. This analysis also provides the basis for revision of YMP/90-55Q, Q-List (YMP 1998). The Q-List identifies those MGR SSCs subject to the requirements of DOE/RW-0333P, ''Quality Assurance Requirements and Description'' (QARD) (DOE 1998)

  10. Gas-handling system for studies of tritium-containing materials

    International Nuclear Information System (INIS)

    Carstens, D.H.W.

    1975-01-01

    A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

  11. On the detection of chemical reactions in the systems containing tritium and luminescent labels

    International Nuclear Information System (INIS)

    Krasnyanskij, A.V.

    1997-01-01

    Features of detecting chemical processes in scintillation systems containing tritium, are considered on the base of a model, connecting the counting rate with the mass of cenverted substance. It is shown that the character of the cunting rate dependence on the mass of the converted phase is determined by the spatial distribution of scintillating and radioactive phases in microheterogeneous systems. Calculation results can be used for designing sensor elements, based on radionuclide luminescent probe

  12. Stability Enhancement of a Power System Containing High-Penetration Intermittent Renewable Generation

    OpenAIRE

    Morel, Jorge; Obara, Shin’ya; Morizane, Yuta

    2015-01-01

    This paper considers the transient stability enhancement of a power system containing large amounts of solar and wind generation in Japan. Following the Fukushima Daiichi nuclear disaster there has been an increasing awareness on the importance of a distributed architecture, based mainly on renewable generation, for the Japanese power system. Also, the targets of CO2 emissions can now be approached without heavily depending on nuclear generation. Large amounts of renewable generation leads to...

  13. Effect of Liquid Crystalline Systems Containing Antimicrobial Compounds on Infectious Skin Bacteria.

    Science.gov (United States)

    Souza, Carla; Watanabe, Evandro; Aires, Carolina Patrícia; Lara, Marilisa Guimarães

    2017-08-01

    This study aimed (i) to prepare liquid crystalline systems (LCS) of glyceryl monooleate (GMO) and water containing antibacterial compounds and (ii) to evaluate their potential as drug delivery systems for topical treatment of bacterial infections. Therefore, LCS containing CPC (cetylpyridinium chloride) (LCS/CPC) and PHMB (poly(hexamethylene biguanide) hydrochloride) (LCS/PHMB) were prepared and the liquid crystalline phases were identified by polarizing light microscopy 24 h and 7 days after preparation. The in vitro drug release profile and in vitro antibacterial activity of the systems were assessed using the double layer agar diffusion method against Staphylococcus aureus, methicillin-resistant S. aureus, Staphylococcus epidermidis, Escherichia coli, and Enterococcus faecalis. The interaction between GMO and the drugs was evaluated by a drug absorption study. Stable liquid crystalline systems containing CPC and PHMB were obtained. LCS/PHMB decreased the PHMB release rate and exerted strong antibacterial activity against all the investigated bacteria. In contrast, CPC interacted with GMO so strongly that it became attached to the system; the amount released was not sufficient to exert antibacterial activity. Therefore, the studied liquid crystalline systems were suitable to deliver PHMB, but not CPC. Accordingly, it was demonstrated that GMO interacts with each drug differently, which may interfere in the final efficiency of GMO/water LCS.

  14. Properies of binder systems containing cement, fly ash, and limestone powder

    Directory of Open Access Journals (Sweden)

    Krittiya Kaewmanee

    2014-10-01

    Full Text Available Fly ash and limestone powder are two major widely available cement replacing materials in Thailand. However, the current utilization of these materials is still not optimized due to limited information on properties of multi-binder systems. This paper reports on the mechanical and durability properties of mixtures containing cement, fly ash, and limestone powder as single, binary, and ternary binder systems. The results showed that a single binder system consisting of only cement gave the best carbonation resistance. A binary binder system with fly ash exhibited superior performances in long-term compressive strength and many durability properties except carbonation and magnesium sulfate resistances, while early compressive strength of a binary binder system with limestone powder was excellent. The ternary binder system, taking the most benefit of selective cement replacing materials, yielded, though not the best, satisfactory performances in almost all properties. Thus, the optimization of binders can be achieved through a multi-binder system.

  15. Modeling Users' Experiences with Interactive Systems

    CERN Document Server

    Karapanos, Evangelos

    2013-01-01

    Over the past decade the field of Human-Computer Interaction has evolved from the study of the usability of interactive products towards a more holistic understanding of how they may mediate desired human experiences.  This book identifies the notion of diversity in usersʼ experiences with interactive products and proposes methods and tools for modeling this along two levels: (a) interpersonal diversity in usersʽ responses to early conceptual designs, and (b) the dynamics of usersʼ experiences over time. The Repertory Grid Technique is proposed as an alternative to standardized psychometric scales for modeling interpersonal diversity in usersʼ responses to early concepts in the design process, and new Multi-Dimensional Scaling procedures are introduced for modeling such complex quantitative data. iScale, a tool for the retrospective assessment of usersʼ experiences over time is proposed as an alternative to longitudinal field studies, and a semi-automated technique for the analysis of the elicited exper...

  16. Muon-catalyzed fusion experiment target and detector system. Preliminary design report

    International Nuclear Information System (INIS)

    Jones, S.E.; Watts, K.D.; Caffrey, A.J.; Walter, J.B.

    1982-03-01

    We present detailed plans for the target and particle detector systems for the muon-catalyzed fusion experiment. Requirements imposed on the target vessel by experimental conditions and safety considerations are delineated. Preliminary designs for the target vessel capsule and secondary containment vessel have been developed which meet these requirements. In addition, the particle detection system is outlined, including associated fast electronics and on-line data acquisition. Computer programs developed to study the target and detector system designs are described

  17. Effect of long-lived containers on the postclosure performance of a reference disposal system

    International Nuclear Information System (INIS)

    Goodwin, B.W.; Hajas, W.C.; LeNeveu, D.M.

    1996-05-01

    The concept for disposal of Canada's nuclear fuel waste involves isolating the waste in corrosion-resistant containers emplaced in a scaled vault at a depth of 500 to 1000 m in plutonic rock of the Canadian Shield. The concept permits a choice of methods, materials, site locations, and designs. The technical feasibility of this concept and its impact on the environment and human health are summarized in an Environmental Impact Statement (AECL 1994a,b), supported by nine detailed reference documents (Davis et al. 1993; Davison et al. 1994a,b; Goodwin et al. 1994; Greber et al. 1994; Grondin et al. 1994; Johnson et al. 1994a,b; Simmons and Baumgartner 1994). In the assessment of the reference disposal system, we assumed the containers encapsulating the nuclear fuel waste were constructed from Grade-2 titanium. In this report, we investigate the effect of a different choice, and assume the use of long-lived containers constructed from materials such as high-purity copper or Grades-12 or -16 titanium alloys. These alternative materials would provide much longer periods of protection, based on the expectation that the only container failure mechanism, for times up to 10 5 a, involves initial fabrication defects. We explore the effects of long-lived containers for the same vault layout and orientation that were assumed for the reference disposal vault. We also explore effects for two less favourable situations, in which the vault is closer to a nearby fracture zone and in which the vault is extended to have emplacement rooms on both sides of the fracture zone. Our analyses use the probabilistic assessment computer code, SYVAC3-CC3, an acronym for SYstems Variability Analysis Code, generation 3. with a system model describing the Canadian Concept, generation 3, for the disposal of nuclear fuel waste. The input data for the code have been adjusted to approximate the expected protection characteristics of alternative container materials. (author). 31 refs., 1 tab., 16 figs

  18. Apollo experience report: Earth landing system

    Science.gov (United States)

    West, R. B.

    1973-01-01

    A brief discussion of the development of the Apollo earth landing system and a functional description of the system are presented in this report. The more significant problems that were encountered during the program, the solutions, and, in general, the knowledge that was gained are discussed in detail. Two appendixes presenting a detailed description of the various system components and a summary of the development and the qualification test programs are included.

  19. Power control and management of the grid containing largescale wind power systems

    Science.gov (United States)

    Aula, Fadhil Toufick

    The ever increasing demand for electricity has driven many countries toward the installation of new generation facilities. However, concerns such as environmental pollution and global warming issues, clean energy sources, high costs associated with installation of new conventional power plants, and fossil fuels depletion have created many interests in finding alternatives to conventional fossil fuels for generating electricity. Wind energy is one of the most rapidly growing renewable power sources and wind power generations have been increasingly demanded as an alternative to the conventional fossil fuels. However, wind power fluctuates due to variation of wind speed. Therefore, large-scale integration of wind energy conversion systems is a threat to the stability and reliability of utility grids containing these systems. They disturb the balance between power generation and consumption, affect the quality of the electricity, and complicate load sharing and load distribution managing and planning. Overall, wind power systems do not help in providing any services such as operating and regulating reserves to the power grid. In order to resolve these issues, research has been conducted in utilizing weather forecasting data to improve the performance of the wind power system, reduce the influence of the fluctuations, and plan power management of the grid containing large-scale wind power systems which consist of doubly-fed induction generator based energy conversion system. The aims of this research, my dissertation, are to provide new methods for: smoothing the output power of the wind power systems and reducing the influence of their fluctuations, power managing and planning of a grid containing these systems and other conventional power plants, and providing a new structure of implementing of latest microprocessor technology for controlling and managing the operation of the wind power system. In this research, in order to reduce and smooth the fluctuations, two

  20. 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.)

  1. FLUENT calculations of the hydrogen distribution in a containment during the OECD-NEA THAI HM-2 experiment

    International Nuclear Information System (INIS)

    Visser, D.C.; Komen, E.M.J.; Houkema, M.; Siccama, N.B.; Kyttaelae, Juha; Huhtanen, Risto; Takasuo, Eveliina

    2009-01-01

    Hydrogen may be released into the containment atmosphere of a nuclear power plant during a severe accident. Locally, high hydrogen concentrations may be reached that can possibly cause fast deflagration or even detonation and put the integrity of the containment at risk. Therefore, the distribution and mixing of hydrogen is an important safety issue for nuclear power plants. Computer codes can be applied to predict the hydrogen distribution in the containment within the course of a hypothetical severe accident and get an estimate of the local hydrogen concentration in the various zones of the containment. In this way the risk associated with the hydrogen safety issue can be determined, and safety related measurements and procedures could be assessed. In order to validate the existing computer codes in the context of hydrogen distribution in the containment of a nuclear power plant, experimental benchmark studies have been performed in the German Thermal-hydraulics, Hydrogen, Aerosols and Iodine (THAI) facility in the framework of the OECD-NEA THAI project. In order to demonstrate the capabilities of the commercial Computational Fluid Dynamics (CFD) code FLUENT the THAI HM-2 test was simulated independently by NRG and VTT. In the first phase of the HM-2 test a stratified hydrogen rich light gas layer was established in the upper part of the THAI containment. In the second phase steam was injected at a lower position inducing a rising plume that gradually dissolved the stratified hydrogen-rich layer from below. Thermo-dynamic phenomena like natural convection, mixing, condensation, heat transfer and distribution in different zones that are expected in severe accidents are involved. The calculated results by NRG and VTT (on hydrogen concentration, temperature, pressure and flow velocity) are compared to the experimental results. The most important differences between the CFD model of NRG and VTT are the computational mesh, condensation model and treatment of the solid

  2. Study on vent containment filtering for the Spanish NPPS systems; Estudio sobre Sistemas de Centeo Filtrado de Contencion para las CCNN Espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Peinado, A.; Serrano, C.; Garcia-Serrano, J. L.

    2013-07-01

    The study discusses filtering systems on the market, and its suppliers, taking into account aspects such as ease of integration into the current plant design, characteristics of the process of filtering, operational range, autonomy of the system, maintenance, qualification and proven experiences, among others. The study, also contains an analysis of sequences kind of accident that serve to define the design parameters of the system.

  3. CTL Industries simplifies mining processes : fill containment system a simple solution for backfilling stopes

    Energy Technology Data Exchange (ETDEWEB)

    Larmour, A.

    2010-12-01

    Several large mining companies are interested in a patent-pending fill containment system designed to replace shotcrete fences after a stope has been filled. The existing process is extremely labour and material intensive. This article discussed how the new system has made work easier, safer and increased the bottom line. The article described the design and specifications of the shotcrete containment system. The advantages of the system include no wait time for customers for shotcrete teams to build fences, and it is a quick, easy process that takes about 3 hours to set up, inflate and fill. The article also outlined a shaft guide electronic profiling mechanism that had been developed to complement the shaft guide laminating device. The product electronically measures shaft guides in mine shafts to determine how much wear has occurred on the timber. Last, the article described a refuge station door sealant kit with a cabinet that contains items required to seal the refuge station door should an emergency occur. This quick and easy system replaces the previous method in which clay slugs are soaked in a five-gallon pail of water and applied by hand. 1 ref., 1 fig.

  4. Passive containment cooling system performance in the simplified boiling water reactor

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Gamble, R.E.; Yadigaroglu, G.

    1997-01-01

    The Simplified Boiling Water Reactor (SBWR) incorporates a passive system for decay heat removal from the containment in the event of a postulated Loss-of-Coolant Accident (LOCA). Decay heat is removed by condensation of the steam discharged from the reactor pressure vessel (RPV) in three condensers which comprise the Passive Containment Cooling System (PCCS). These condensers are designed to carry the heat load while transporting a mixture of steam and noncondensible gas (primarily nitrogen) from the drywell to the suppression chamber. This paper describes the expected LOCA response of the SBWR with respect to the PCCS performance, based on analysis and test results. The results confirm that the PCCS has excess capacity for decay heat removal and that overall system performance is very robust. 12 refs., 8 figs

  5. Enviro-geotechnical considerations in waste containment system design and analysis

    International Nuclear Information System (INIS)

    Fang, H.Y.; Daniels, J.L.; Inyang, H.I.

    1997-01-01

    The effectiveness of waste control facilities hinges on careful evaluation of the overall planning, analysis and design of the entire system prior to construction. At present, most work is focused on the waste controlling system itself, with little attention given to the local environmental factors surrounding the facility sites. Containment materials including geomembranes, geotextiles and clay amended soils have received intense scrutiny. This paper, however, focuses on three relatively important issues relating to the characterization of the surrounding geomedia. Leakage through naturally occurring low-permeability soil layers, shrinkages swelling, cracking and effects of dynamic loads on system components are often responsible for a waste containment breach. In this paper, these mechanisms and their synergistic effects are explained in terms of the particle energy field theory. It is hoped that this additional information may assist the designer to be aware or take precaution to design safer future waste control facilities

  6. Results of laboratory tests on a robust filtration system for PWR containments in the case of a serious accident

    International Nuclear Information System (INIS)

    L'Homme, A.; Berlin, M.; Beraud, G.

    1986-01-01

    A study is currently in progress in France on a simple filtration process using sand as a filtration medium which, in the event of a serious accident leading to core meltdown in a pressurized water reactor, will permit controlled and filtered releases from the containment. Laboratory tests on sand filters for aerosols have been conducted. The tests involved the use of columns of sand, 80 cm high and 20 cm in diameter, under conditions which were similar to those inside the containment of a PWR in which a serious accident has occurred. The sand granulometry, the aerosol particle size and the flow rate and steam content of the fluid to be filtered were variable parameters. The results obtained from the experiment showed that as a filtration medium for this simple filter system for reactors a sand obtainable from the Cattenom quarry was most suitable. For this sand the filtration coefficient for aerosols is greater than 10 and the pressure drop is less than 10 4 pascals. Experience has also shown that there is no risk, under the operating conditions envisaged, that the filter will become clogged by aerosols or steam from condensed water or that there will be any major escape of aerosols retained during long-term operation of the filter or caused by the vaporisation of the condensed water. A larger scale experiment is already being carried out. (author)

  7. Operational experience with the CEBAF control system

    International Nuclear Information System (INIS)

    Hovater, C.; Chowdhary, M.; Karn, J.; Tiefenback, M.; Zeijts, J. van; Watson, W.

    1996-01-01

    The CEBAF accelerator at Thomas Jefferson National Accelerator Facility (Jefferson Lab) successfully began its experimental nuclear physics program in November of 1995 and has since surpassed predicted machine availability. Part of this success can be attributed to using the EPICS (Experimental Physics and Industrial Control System) control system toolkit. The CEBAF control system is one of the largest accelerator control system now operating. It controls approximately 338 SRF cavities, 2,300 magnets, 500 beam position monitors and other accelerator devices, such as gun hardware and other beam monitoring devices. All told, the system must be able to access over 125,000 database records. The system has been well received by both operators and the hardware designers. The EPICS utilities have made the task of troubleshooting systems easier. The graphical and test-based creation tools have allowed operators to custom build control screens. In addition, the ability to integrate EPICS with other software packages, such as Tcl/Tk, has allowed physicists to quickly prototype high-level application programs, and to provide GUI front ends for command line driven tools. Specific examples of the control system applications are presented in the areas of energy and orbit control, cavity tuning and accelerator tune up diagnostics

  8. Investigation of light gas effects on passive containment cooling system in ALWR

    International Nuclear Information System (INIS)

    Paladino, D.; Auban, O.; Huggenberger, M.; Andreani, M.

    2003-01-01

    The large-scale thermal-hydraulic PANDA facility has been used for the last years for investigating passive decay-heat removal systems and related containment phenomena relevant for current and next generation of light water reactors. Passive Containment Cooling System (PCCS) systems operate by transferring heat from the containment to a water pool located outside the containment by steam condensation, and serve to mitigate long-term pressure build-up in the event of steam discharge from the primary circuit. As part of the 5 th Euratom framework program project TEMPEST, a new series of tests was performed in the PANDA facility to experimentally investigate the distribution of non-condensable gases inside the containment and their effect on the performance of PCCS of the European Simplified Boiling Water Reactor (ESBWR). The influence of light gas(hydrogen) on the PCCs performance is of special interest. Hydrogen release caused by the metalwater reaction in case of severe accident was simulated in PANDA by injecting helium into the lines feeding the break flow from the reactor pressure vessel to the Drywells. The paper combines the presentation of experimental results for a number of PANDA tests and the analysis performed using the GOTHIC code. As GOTHIC has 3-D modeling capabilities, gas distribution effects could be studied. The comparison of GOTHIC calculations (two pre-test and one post-test with the same model) with selected TEMPEST tests showed that the code is capable to predict well gas stratification in the drywell, while the system pressure increase due to the release of light gas is slightly overestimated. The analysis aiming to clarify the discordance between the GOTHIC simulation and the experimental results is included in this paper

  9. 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

  10. Development of integrated containment and surveillance system for fast critical facility FCA. Portal and penetration monitors

    International Nuclear Information System (INIS)

    Mukaiyama, Takehiko; Ogawa, Hironobu; Yokota, Yasuhiro.

    1998-01-01

    Manpower and radiation exposure problems, accompanied by frequent Non Destructive Assay (NDA) based inspections at the Fast Critical Facility FCA of Japan Atomic Energy Research Institute (JAERI), are a burden for both the inspectorates and the facility operator. In the hope of alleviating these burdens, the development of containment and surveillance measures for the FCA was initiated in 1979. The integrated containment and surveillance system consists of a portal monitor and a penetration monitor. The reactor building provides an ideal containment measure because of its explosion-proof, airtight structure and limited number of penetrations. The function of the portal monitor is to detect undeclared removal of nuclear material from the reactor building through the doorway. The penetration monitor is designed for surveillance of diversion routes through containment boundaries, and of safeguards related activities for bypassing the portal monitor. The combination of monitoring by the penetration monitor of containment boundaries and all their penetrations except for the doorway, and monitoring by the portal monitor, provides complete coverage of realistic diversion routes. The development of the system was completed in 1988 and the field trial test was conducted for the period of twelve running months. The final report on the field trial was concluded on January 1990. The major conclusion of the report was that the system is effective, reliable and efficient. Following this successful conclusion, the International Atomic Energy Agency (IAEA) accepted the system for meeting its safeguards goals at the FCA on condition that an independent IAEA authentication equipment is provided. The development of the authentication equipment is accomplished as an separate Japan Support Programme for Agency Safeguards (JASPAS) task. (author)

  11. The scaling of experiments on volcanic systems

    Directory of Open Access Journals (Sweden)

    Olivier eMERLE

    2015-06-01

    Full Text Available In this article, the basic principles of the scaling procedure are first reviewed by a presentation of scale factors. Then, taking an idealized example of a brittle volcanic cone intruded by a viscous magma, the way to choose appropriate analogue materials for both the brittle and ductile parts of the cone is explained by the use of model ratios. Lines of similarity are described to show that an experiment simulates a range of physical processes instead of a unique natural case. The pi theorem is presented as an alternative scaling procedure and discussed through the same idealized example to make the comparison with the model ratio procedure. The appropriateness of the use of gelatin as analogue material for simulating dyke formation is investigated. Finally, the scaling of some particular experiments such as pyroclastic flows or volcanic explosions is briefly presented to show the diversity of scaling procedures in volcanology.

  12. Transport of 152Eu colloids in a system of fine sand and water containing humic substances

    International Nuclear Information System (INIS)

    Klotz, D.

    1995-01-01

    The migration of 152 Eu in a system of fine sand and water containing humic substances was investigated in a flow column system under realistic conditions. In this system, the trivalent Eu forms colloids with the water. These Eu humates are transported without retardation at recovery rates significantly below 100 per cent. Recovery is more or less a measure of the physical process of filtration of Eu bonded to particulates. In the range of natural filtering rates, the recovery rates decrease with decreasing filtering rate. (orig.) [de

  13. Code for calculation of spreading of radioactivity in reactor containment systems

    International Nuclear Information System (INIS)

    Vertes, P.

    1992-09-01

    A detailed description of the new version of TIBSO code is given, with applications for accident analysis in a reactor containment system. The TIBSO code can follow the nuclear transition and the spatial migration of radioactive materials. The modelling of such processes is established in a very flexible way enabling the user to investigate a wide range of problems. The TIBSO code system is described in detail, taking into account the new developments since 1983. Most changes improve the capabilities of the code. The new version of TIBSO system is written in FORTRAN-77 and can be operated both under VAX VMS and PC DOS. (author) 5 refs.; 3 figs.; 21 tabs

  14. Experience with Nuclear Medicine Information System

    Directory of Open Access Journals (Sweden)

    Bilge Volkan-Salanci

    2012-12-01

    Full Text Available Objective: Radiology information system (RIS is basically evolved for the need of radiologists and ignores the vital steps needed for a proper work flow of Nuclear Medicine Department. Moreover, CT/MRI oriented classical PACS systems are far from satisfying Nuclear Physicians like storing dynamic data for reprocessing and quantitative analysis of colored images. Our purpose was to develop a workflow based Nuclear Medicine Information System (NMIS that fulfills the needs of Nuclear Medicine Department and its integration to hospital PACS system. Material and Methods: Workflow in NMIS uses HL7 (health level seven and steps include, patient scheduling and retrieving information from HIS (hospital information system, radiopharmacy, acquisition, digital reporting and approval of the reports using Nuclear Medicine specific diagnostic codes. Images and dynamic data from cameras of are sent to and retrieved from PACS system (Corttex© for reprocessing and quantitative analysis. Results: NMIS has additional functions to the RIS such as radiopharmaceutical management program which includes stock recording of both radioactive and non-radioactive substances, calculation of the radiopharmaceutical dose for individual patient according to body weight and maximum permissible activity, and calculation of radioactivity left per unit volume for each radionuclide according their half lives. Patient scheduling and gamma camera patient work list settings were arranged according to specific Nuclear Medicine procedures. Nuclear Medicine images and reports can be retrieved and viewed from HIS. Conclusion: NMIS provides functionality to standard RIS and PACS system according to the needs of Nuclear Medicine. (MIRT 2012;21:97-102

  15. Personnel protection and beam containment systems for the 3 GeV Injector

    International Nuclear Information System (INIS)

    Yotam, R.; Cerino, J.; Garoutte, R.; Hettel, R.; Horton, M.; Sebek, J.; Benson, E.; Crook, K.; Fitch, J.; Ipe, N.; Nelson, G.; Smith, H.

    1991-01-01

    The 3 GeV Injector is the electron beam source for the SPEAR Storage Ring, and its personnel safety system was designed to protect personnel from both radiation exposure and electrical hazards. The Personnel Protection System (PPS) was designed and implemented with complete redundancy and is a relay based interlock system completely independent from the machine protection system. A comprehensive monitoring of the system status, and control of the Injector PPS from the SPEAR Control Room via the control computer is a feature. The Beam Containment System (BCS) is based on beam current measurements along the Linac and on Beam Shut Off Ion Chambers (BSOIC) installed outside the Linac, at several locations around the Booster, and around the SPEAR storage ring. An outline of the design criteria is presented with more detailed description of the philosophy of the PPS logic and the BCS

  16. Stability Enhancement of a Power System Containing High-Penetration Intermittent Renewable Generation

    Directory of Open Access Journals (Sweden)

    Jorge Morel

    2015-06-01

    Full Text Available This paper considers the transient stability enhancement of a power system containing large amounts of solar and wind generation in Japan. Following the Fukushima Daiichi nuclear disaster there has been an increasing awareness on the importance of a distributed architecture, based mainly on renewable generation, for the Japanese power system. Also, the targets of CO2 emissions can now be approached without heavily depending on nuclear generation. Large amounts of renewable generation leads to a reduction in the total inertia of the system because renewable generators are connected to the grid by power converters, and transient stability becomes a significant issue. Simulation results show that sodium-sulfur batteries can keep the system in operation and stable after strong transient disturbances, especially for an isolated system. The results also show how the reduction of the inertia in the system can be mitigated by exploiting the kinetic energy of wind turbines.

  17. Attempt at a Systemic Design of a Protocell: Connecting information, Metabolism and Container

    DEFF Research Database (Denmark)

    Albertsen, Anders N.; Maurer, Sarah; Cape, Jonathan

    many of the basic characteristics of a living system, but usually lack the gene mediated regulation functions that natural cells possess. To address this issue, we are attempting a systemic approach (Rasmussen, 2004) in implementing a simple, chemical system that contains three major types of molecules...... of the design and the already demonstrated advantages of the systemic approach in unravelling interactions (Declue, 2009; Maurer, 2011) between the components and their significances for the self-maintenance and self-replication of a protocell. References DeClue, MS, Monnard, P-A, Bailey, JA, Maurer, SE, Collis......The minimal requirements for a living system are often listed as follows: i) a living system must have a specific identity and be able to preserve it (compartmentalization) ; ii) it must sustain itself by using energy from its environment to manufacture at least some of its components from...

  18. Development of the environmental qualification safety requirement matrix for the containment system of in-service CANDU reactors

    International Nuclear Information System (INIS)

    Chun, R.M.; Low, J.; Sobolewski, J.

    1994-01-01

    Over the last several years, Ontario Hydro Nuclear (OHN) has placed increasing emphasis on environmental qualification (EQ) at its Pickering and Bruce NGS A and B nuclear generating stations (NGSs). The program currently underway (at the time of the conference) builds upon the experience gained from the extensive Darlington NGS EQ experience and from EQ programs conducted by other utilities. Some of the major steps of the OHN EQ program include: defining Safety Requirement Matrices (SRMs), establishing environmental conditions, developing an EQ List, conducting an EQ Assessment and maintaining Operational EQ Assurance during the plant life. The SRM identifies safety related components, their required safety functions and their mission times for each postulated design basis accident (DBA). This is a critical step, as the SRM defines the equipment that requires assurance of EQ and precise requirements must be provided to ensure a cost effective EQ program. This paper describes the development of the SRMs for the containment system of the Bruce stations. The introductory section briefly discusses how the industry has dealt with equipment qualification as it has evolved and the role of the SRMs in the OHN EQ Program. In Section 2, the preparation of the SRM is described along with the applicable ground rules used. The results of the application of the SRM preparation guidelines to the containment system are discussed in Section 3. A summary of the major findings and conclusions is presented. 3 refs., 3 figs

  19. Distributed finite-time containment control for double-integrator multiagent systems.

    Science.gov (United States)

    Wang, Xiangyu; Li, Shihua; Shi, Peng

    2014-09-01

    In this paper, the distributed finite-time containment control problem for double-integrator multiagent systems with multiple leaders and external disturbances is discussed. In the presence of multiple dynamic leaders, by utilizing the homogeneous control technique, a distributed finite-time observer is developed for the followers to estimate the weighted average of the leaders' velocities at first. Then, based on the estimates and the generalized adding a power integrator approach, distributed finite-time containment control algorithms are designed to guarantee that the states of the followers converge to the dynamic convex hull spanned by those of the leaders in finite time. Moreover, as a special case of multiple dynamic leaders with zero velocities, the proposed containment control algorithms also work for the case of multiple stationary leaders without using the distributed observer. Simulations demonstrate the effectiveness of the proposed control algorithms.

  20. Data processing system for NBT experiments

    International Nuclear Information System (INIS)

    Takahashi, C.; Hosokawa, M.; Shoji, T.; Fujiwara, M.

    1981-07-01

    Data processing system for Nagoya Bumpy Torus (NBT) has been developed. Since plasmas are produced and heated in steady state by use of high power microwaves, sampling and processing data prevails in long time scale on the order of one minute. The system, which consists of NOVA 3/12 minicomputer and many data acquisition devices, is designed to sample and process large amount of data before the next discharge starts. Several features of such long time scale data processing system are described in detail. (author)

  1. Operating experience feedback report - Air systems problems

    International Nuclear Information System (INIS)

    Ornstein, H.L.

    1987-12-01

    This report highlights significant operating events involving observed or potential failures of safety-related systems in U.S. plants that resulted from degraded or malfunctioning non-safety grade air systems. Based upon the evaluation of these events, the Office for Analysis and Evaluation of Operational Data (AEOD) concludes that the issue of air systems problems is an important one which requires additional NRC and industry attention. This report also provides AEOD's recommendations for corrective actions to deal with the issue. (author)

  2. General distributed control system for fusion experiments

    International Nuclear Information System (INIS)

    Klingner, P.L.; Levings, S.J.; Wilkins, R.W.

    1986-01-01

    A general control system using distributed LSI-11 microprocessors is being developed. Common software residues in each LSI-11 and is tailored to an application by control specifications downloaded from a host computer. The microprocessors, their control interfaces, and the micro-to-host communications are CAMAC based. The host computer also supports an operator interface, coordination of multiple microprocessors, and utilities to create and maintain the control specifications. Typical applications include monitoring safety interlocks as well as controlling vacuum systems, high voltage charging systems, and diagnostics

  3. Use of a FORTH-based PROLOG for real-time expert systems. 1: Spacelab life sciences experiment application

    Science.gov (United States)

    Paloski, William H.; Odette, Louis L.; Krever, Alfred J.; West, Allison K.

    1987-01-01

    A real-time expert system is being developed to serve as the astronaut interface for a series of Spacelab vestibular experiments. This expert system is written in a version of Prolog that is itself written in Forth. The Prolog contains a predicate that can be used to execute Forth definitions; thus, the Forth becomes an embedded real-time operating system within the Prolog programming environment. The expert system consists of a data base containing detailed operational instructions for each experiment, a rule base containing Prolog clauses used to determine the next step in an experiment sequence, and a procedure base containing Prolog goals formed from real-time routines coded in Forth. In this paper, we demonstrate and describe the techniques and considerations used to develop this real-time expert system, and we conclude that Forth-based Prolog provides a viable implementation vehicle for this and similar applications.

  4. A control system for a free electron laser experiment

    International Nuclear Information System (INIS)

    Giove, D.

    1992-01-01

    The general layout of a control and data acquisition system for a Free Electron Laser experiment will be discussed. Some general considerations about the requirements and the architecture of the whole system will be developed. (author)

  5. Fission products distributions in Candu primary heat transport and Candu containment systems during a severe accident

    International Nuclear Information System (INIS)

    Constantin, Marin; Rizoiu, Andrei

    2005-01-01

    The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) and CANDU Containment Systems by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity both of CANDU PHT and Containment System were strong motivations to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated and a simplified FPs inventory, some simplifications in the feeders geometry and containment were used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The containment model consists of 4 rooms connected between by 6 links. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by the CATHENA code and the source term of FPs introduced into the PHT was estimated by the ORIGEN code. The FPs distribution in the nodes of the circuit and the FPs mass transfer per isotope and chemical species are obtained by using SOPHAEROS module of ASTEC code. The distributions into the containment are obtained by the CPA module of ASTEC code (thermalhydraulics calculations in the containment and FPs aerosol transport). The results consist of mass distributions in the nodes of the circuit and the transferred mass to the containment through the break for different species (FPs and chemical species) and mass distributions in the different parts and

  6. The silicon tracking system of the CBM experiment

    Energy Technology Data Exchange (ETDEWEB)

    Balog, Tomas [GSI Helmholtzzentrum fuer Schwerionenforschung GmbH, Darmstadt (Germany); Collaboration: CBM-Collaboration

    2014-07-01

    The Compressed Baryonic Matter (CBM) experiment at FAIR will explore the phase diagram of strongly interacting matter at the highest net-baryon densities in nucleus-nucleus collisions with interaction rates up to 10 MHz. As the core tracking detector of CBM the Silicon Tracking System (STS) will be installed in the gap of the 1 T super conducting dipole magnet for reconstruction of charged particle trajectories and its momenta. The requirement on momentum resolution, Δp/p=1%, can only be achieved with an ultra-low material budget, imposing particular restrictions on the location of 2.5 million channel front-end electronics dissipating 40 KW in the fiducial volume of about 2 m{sup 3}. The concept of the STS is based on a modular structure containing 300 μm thick double-sided silicon microstrip sensors read out through ultra-thin multi-line micro-cables with fast self-triggering electronics. As central building blocks the modules consisting of each a sensor, micro-cable and front-end electronics will be mounted with lightweight carbon fiber support structures onto 8 detector stations. At the station periphery infrastructure such as power and cooling lines will be placed. The status of the STS development is summarized in the presentation, including an overview on sensors, read-out electronics, prototypes, and system integration.

  7. Adapting to Biology: Maintaining Container-Closure System Compatibility with the Therapeutic Biologic Revolution.

    Science.gov (United States)

    Degrazio, Dominick

    Many pharmaceutical companies are transitioning their research and development drug product pipeline from traditional small-molecule injectables to the dimension of evolving therapeutic biologics. Important concerns associated with this changeover are becoming forefront, as challenges develop of varying complexity uncommon with the synthesis and production of traditional drugs. Therefore, alternative measures must be established that aim to preserve the efficacy and functionality of a biologic that might not be implemented for small molecules. Conserving protein stability is relative to perpetuating a net equilibrium of both intrinsic and extrinsic factors. Key to sustaining this balance is the ability of container-closure systems to maintain their compatibility with the ever-changing dynamics of therapeutic biologics. Failure to recognize and adjust the material properties of packaging components to support compatibility with therapeutic biologics can compromise patient safety, drug productivity, and biological stability. This review will examine the differences between small-molecule drugs and therapeutic biologics, lay a basic foundation for understanding the stability of therapeutic biologics, and demonstrate potential sources of container-closure systems' incompatibilities with therapeutic biologics at a mechanistic level. Many pharmaceutical companies are transitioning their research and development drug product pipeline from traditional small-molecule injectables to recombinantly derived therapeutic biologics. Concerns associated with this transformation are becoming prominent, as therapeutic biologics are uncharacteristic to small-molecule drugs. Maintaining the stability of a therapeutic biologic is a combination of balancing intrinsic factors and external elements within the biologic's microenvironment. An important aspect of this balance is relegated to the overall compatibility of primary, parenteral container-closure systems with therapeutic biologics

  8. MTX [Microwave Tokamak Experiment] plasma diagnostic system

    International Nuclear Information System (INIS)

    Rice, B.W.; Hooper, E.B.; Brooksby, C.A.

    1987-01-01

    In this paper, a general overview of the MTX plasma diagnostics system is given. This includes a description of the MTX machine configuration and the overall facility layout. The data acquisition system and techniques for diagnostic signal transmission are also discussed. In addition, the diagnostic instruments planned for both an initial ohmic-heating set and a second FEL-heating set are described. The expected range of plasma parameters along with the planned plasma measurements will be reviewed. 7 refs., 5 figs

  9. Cost-benefit of the bubble tower concept as a containment passive safety system

    International Nuclear Information System (INIS)

    Iotti, R.C.; Bardach, H.; Shin, J.J.; Parnes, M.J.

    1994-01-01

    Containment system integrity for both PWRs and BWRs can be assured by passive measures highlighted the use of an accessory Bubble Tower. The utilization of the Bubble Tower precludes the possibility of containment overpressurization. From the thermodynamic standpoint, the Bubble Tower is simply water column of about 120 ft. height attached to the containment and connected to the air space above the suppression pool of a BWR, or a PWR In-containment Refueling Water Storage Tank. From the radiological protection standpoint, the Bubble Tower is a water column sufficient to effect decontamination factors of at least 100 for nuclide species other than the noble gases, and with the addition of organic solubilizers sufficient to effect decontamination factors of at least 10 iodides and at least 100 for other nuclide species. When containment steam or noncondensable gas passes through the Bubble Tower, a significant fraction of the radionuclides is absorbed by the water column. When a cost-benefit dose evaluation is performed relative to the utilization of a Bubble Tower, even under conditions where the dollars per man-rem is taken as $1000, the results are favorable. They are substantially more favorable when the dollars per man-rem is taken as $5000 or $10,000 as are the current trends. (author)

  10. Life cycle assessment of façade coating systems containing manufactured nanomaterials

    Energy Technology Data Exchange (ETDEWEB)

    Hischier, Roland, E-mail: roland.hischier@empa.ch; Nowack, Bernd [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland); Gottschalk, Fadri [ETSS (Switzerland); Hincapie, Ingrid [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland); Steinfeldt, Michael [University of Bremen FB 4/FG 10 Technological Design and Development (Germany); Som, Claudia [Swiss Federal Laboratories for Materials Science and Technology (Empa), Technology and Society Lab (TSL) (Switzerland)

    2015-02-15

    Nanotechnologies are expected to hold considerable potential for the development of new materials in the construction sector. Up to now the environmental benefits and risks of products containing manufactured nanomaterials (MNM) have been quantified only to a limited extent. This study aims to assess the potential environmental, health and safety impacts of coatings containing MNM using Life-cycle assessment: Do paints containing MNM result in a better environmental performance than paints not containing MNM? The study shows that the results depend on a number of factors: (i) The MNM have to substitute an (active) ingredient of the initial paint composition and not simply be an additional ingredient. (ii) The new composition has to extend the lifetime of the paint for such a time period that the consumption of paint along the life cycle of a building is reduced. (iii) Releases of MNM have to be reduced to the lowest level possible (in particular by dumping unused paint together with the packaging). Only when all these boundary conditions are fulfilled, which is the case only for one of the three paint systems examined, is an improved environmental performance of the MNM-containing paint possible for the paint compositions examined in this study.

  11. [Experimental study of percutaneous vertebroplasty with a novel bone void filling container system].

    Science.gov (United States)

    Wang, Tai-Ping; Zhang, Kui-bo; Zheng, Zhao-min; Liu, Hui; Yu, Bin-sheng

    2011-04-19

    To investigate vertebral augmentation with a novel reticulate bone filling container system by polymethyl methacrylate (PMMA) injection in cadaveric simulated vertebral compressive fracture and explore the effect of reticulate bone filling container on cement distribution controlling within vertebral body and the restoration of biomechanical properties after augmentation. A total of 28 freshly frozen human vertebrae specimens were randomly divided into 4 groups. After the measurements of bone mineral density (BMD) and vertebral height, each vertebra received an axle load by a MTS (material testing system) machine to test the initial strength and stiffness. Subsequently a simultaneous compressive fracture model was created to measure the stiffness and height of fractured vertebrae. Then the augmentation procedure was performed. Afterward the biomechanical properties and the vertebral height were similarly measured as pre-operatively. The expansion of bone filling container and the distribution of cement within vertebral body were morphologically observed by crossing the specimens in sagittal midline and also integrated with the radiographic results. Stiffness was significantly restored comparing with that of fractured level (P container groups while it was irregular in single-layer groups. After crossing, the double-layer version expanded well in vertebral body and could enwrap most of injected cement. There was only a little leakage near the vessel layer. But the single-layer version had a poor expansion and a large amount of cement leakage. This novel reticulate bone void filling container system with different layers may restore both the biomechanical properties and the height of fractured vertebrae. But, with the benefit of reducing cement leakage, a double-layer design can enwrap most of injected PMMA and has a brighter prospect of clinical application.

  12. The market-incentive recycling system for waste packaging containers in Taiwan

    International Nuclear Information System (INIS)

    Bor Yunchang, Jeffrey; Chien, Y.-L.; Hsu, Esher

    2004-01-01

    This paper presents a new market-incentive (MI) system to recycle waste-packaging containers in Taiwan. Since most used packaging containers have no or insufficient market value, the government imposes a combined product charge and subsidy policy to provide enough economic incentive for recycling various kinds of packaging containers, such as iron, aluminum, paper, glass and plastic. Empirical results show that the new MI approach has stimulated and established the recycling market for waste-packaging containers. The new recycling system has provided 18,356 employment opportunities and generated NT$ 6.97 billion in real-production value and NT$ 3.18 billion in real GDP during the 1998 survey year. Cost-effectiveness analysis constitutes the theoretical foundation of the new scheme, whereas data used to compute empirical product charge are from two sources: marketing surveys of internal conventional costs of solid-waste collection, disposal and recycling in Taiwan, and benefit transfer of external environmental costs in the United States. The new recycling policy designed by the authors provides a reasonable solution for solid-waste management in a country with limited land resources such as Taiwan

  13. Preliminary CFD Analysis for HVAC System Design of a Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Son, Sung Man; Choi, Choengryul [ELSOLTEC, Yongin (Korea, Republic of); Choo, Jae Ho; Hong, Moonpyo; Kim, Hyungseok [KEPCO Engineering and Construction, Gimcheon (Korea, Republic of)

    2016-10-15

    HVAC (Heating, Ventilation, Air Conditioning) system has been mainly designed based on overall heat balance and averaging concepts, which is simple and useful for designing overall system. However, such a method has the disadvantage that cannot predict the local flow and temperature distributions in a containment building. In this study, a CFD (Computational Fluid Dynamics) preliminary analysis is carried out to obtain detailed flow and temperature distributions in a containment building and to ensure that such information can be obtained via CFD analysis. This approach can be useful for hydrogen analysis in an accident related to hydrogen released into a containment building. In this study, CFD preliminary analysis has been performed to obtain the detailed information of the reactor containment building by using the CFD analysis techniques and to ensure that such information can be obtained via CFD analysis. We confirmed that CFD analysis can offer enough detailed information about flow patterns and temperature field and that CFD technique is a useful tool for HVAC design of nuclear power plants.

  14. Conceptual design of passive containment cooling system with air holdup tanks of improved APR+

    International Nuclear Information System (INIS)

    Jeon, Byong Guk; Cheon No, Hee

    2014-01-01

    In Korea, after the successful validation of passive auxiliary feedwater system (PAFS), a passive containment cooling system (PCCS) gets attention for future development. We suggested PCCS design based on APR+, an advanced PWR developed in Korea, and performed scoping analysis. On the extension of the simple scoping analysis, MARS simulation is performed to incorporate the behavior of water pool outside the containment as well as steam-air mixture inside the containment. Through the simulation we demonstrated the effectiveness of the air holdup tank (AHT). Also we investigated the effect of the models of heat transfer coefficients between steam-air mixture side and water side, and flow instability inside HX tubes. The presence of AHT enables us to reduce the number of required HX tubes more than half through an increase in the heat transfer coefficients due to the reduction of air fraction in the containment. Finally flow instability was observed and mitigated by putting orifice plates at the inlet of tubes, increasing height of return nozzle, and increasing a tube angle. (authors)

  15. A review of experiments comparing systems of grazing management ...

    African Journals Online (AJOL)

    Experiments comparing different systems of grazing management on natural pastures in various parts of the world are reviewed. In experiments in which various rotational systems were tested against continuous grazing, fewer than half revealed pasture improvement relative to continuous grazing. In the majority of ...

  16. Experiment design for identification of structured linear systems

    NARCIS (Netherlands)

    Potters, M.G.

    2016-01-01

    Experiment Design for system identification involves the design of an optimal input signal with the purpose of accurately estimating unknown parameters in a system. Specifically, in the Least-Costly Experiment Design (LCED) framework, the optimal input signal results from an optimisation problem in

  17. Experiences collected with a daylight system

    International Nuclear Information System (INIS)

    Krueger, P.

    1981-01-01

    The Kodak Daylight System is space-saving. It consists of the Kodak-X-Omat cassette, the identification camera and the loading and reloading device (cassette multiloader). All three units of this system are notable for extremely safe operation. Handling is simple and easily appreciated. All brands of film can be used. The data window in the cassette is too large; in some cases, important sections of the image can be covered up by the window. Automatic loading and reloading considerably reduces solling of the cassette, which is otherwise common, and also reduces abrasion of the intensifying screen. Introduction of this system has had a beneficial effect on overall working and has been welcomed by all concerned as a contribution towards facilitating routine work. (orig.) [de

  18. Analysis of Weld Fabrication Flaws in High-Level Radioactive Waste Disposal Containers: Experiences from the US Programme

    International Nuclear Information System (INIS)

    Bullen, Daniel; Apted, Mick

    2002-11-01

    The purpose of this report is to examine key issues regarding the fabrication, closure and defect detection in canisters for radioactive waste disposal in a deep geological repository. As a preliminary step, a review is made of the closure-weld design and non-destructive evaluation (NDE) of the closure seal for the US high-level waste repository programme. This includes statistical analysis of the data obtained by NDE and identification of key areas of investigation where additional data are required. Information from other industrial experiences on closure and flaw detection of metal containers is also reviewed. The canister material and closure methods for the US programme and industrial activities reviewed here differ from those of SKB's KBS-3 reference design. The issues and approaches to issue resolution identified from the US programme and industrial analogues, however, can provide an initial basis for preparing for independent review of SKB's canister closure plans and encapsulation facility

  19. System of accidents notification: the ROSIS experience

    International Nuclear Information System (INIS)

    Coffey, M.; Cunningham, J.

    2009-01-01

    ROSIS is short for 'Radiation Oncology Safety Information System' and it is a voluntary web-based safety information database for Radiotherapy. The system is based on professional front-line staff in radiotherapy clinics reporting incidents and corrective actions over the Internet to a database. On a six years period, 120 health establishments registered more than 1200 events. Almost 98% of statements concern external radiotherapy. The reports can be consulted on the Internet site (www.clin.radfys.lu.se/) besides, a mini training to the risk management in the field of radiotherapy based on the Rosis data has been finalized and proposed for six years. (N.C.)

  20. A PC-based discrete tomography imaging software system for assaying radioactive waste containers

    International Nuclear Information System (INIS)

    Palacios, J.C.; Longoria, L.C.; Santos, J.; Perry, R.T.

    2003-01-01

    A PC-based discrete tomography imaging software system for assaying radioactive waste containers for use in facilities in Mexico has been developed. The software system consists of three modules: (i) for reconstruction transmission tomography, (ii) for reconstruction emission tomography, and (iii) for simulation tomography. The Simulation Module is an interactive computer program that is used to create simulated databases for input to the Reconstruction Modules. These databases may be used in the absence of physical measurements to insure that the tomographic theoretical models are valid and that the coding accurately describes these models. Simulation may also be used to determine the detection limits of the reconstruction methodology. A description of the system, the theory, and a demonstration of the systems capabilities is provided in the paper. The hardware for this system is currently under development