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Sample records for lwrs vessel internal

  1. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun [POSTECH, Daejeon (Korea, Republic of)

    2015-10-15

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  2. Internal structure of an ex-vessel corium debris bed during severe accidents of LWRs

    International Nuclear Information System (INIS)

    Kim, Eunho; Park, Jin Ho; Moriyama, Kiyofumi; Park, Hyun Sun

    2015-01-01

    In the aspect of the coolability assessment the configuration of the debris bed, including internal and external characteristics, has significant importance as boundary conditions for simulations, however, relatively little investigation of the sedimentation process. For the development of a debris bed, recently there have been several studies that focused on thermal characteristics of corium particles. Yakush et al. performed simulation studies and showed that two phase natural convection affects the particle settling trajectory and changes the final arrival location of particles to result more flattened bed. Those simulation results have been supported by the experimental studies of Kim et al. using simulant particles and air bubble injection. For the internal structure of a debris bed, there have been several simulation and experimental studies, which investigated the effect of internal structure on debris bed coolability. Magallon has reported the particle size distribution at three elevations of the debris bed of FARO L-31 case, where the mean particle size was bigger for the lower elevation. However, there is a lack of detailed information on the characteristics of the debris bed, including the local structure and porosity. In this study, we investigated the internal structure of the debris bed using a mixture of stainless steel particles and air bubble injection. Local particle sedimentation quantity, particle size distribution change in radial direction and axial direction, and bed porosity was measured to investigate a relationship between the internal structure and the accident condition. An experimental investigation was carried out for the internal structure of ex-vessel corium debris bed in the flooded cavity during sever accident. Moderate corium discharge in high flooding level was assumed for full fragmentation of melt jet. The test particle mixture was prepared by following an empirical correlation, which reflects the particle size distribution of

  3. Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

    International Nuclear Information System (INIS)

    Heddleson, F.A.

    1975-01-01

    Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

  4. Analysis of the surveillance test data on irradiation embrittlement of the reactor pressure vessel steels in LWRs

    International Nuclear Information System (INIS)

    Lee, Gyoeng Geun; Kwon, Jun Hyun

    2010-11-01

    The surveillance test data in Korean LWRs were analyzed from a viewpoint of materials science. TTS change with the neutron irradiation were compared to the model values of the RG1.99/2 and NUREG/CR-6551. The model values of TTS were higher than the actual values of TTS. It was impossible to find a relationship between TTS and neutron fluence in weld data. The correlation of the increase in YS (yield strength) and TTS with neutron irradiation was also investigated. Like the result of TTS change, the YS/TTS showed the correlations in plate/forgings metals, however no correlation in weld metals. The data were similar to Odette's result about US surveillance tests. From the empirical relationships, the TTS curve change could be predicted using the CVN test result of the unirradiated specimen and the change in YS with neutron irradiation of the specimen

  5. EDF studies on PWR vessel internal loading

    International Nuclear Information System (INIS)

    Bellet, S.; Vallat, S.

    1998-01-01

    EDF has undertaken some mechanics and thermal-hydraulics studies with the objective of mastering plant phenomena today and in order to numerically predict the behaviour of vessel internals on units planned for the future. From some justifications already underway after in operation incidents (wear and drop time of RCCA rods, fuel deflection, adapter cracks, baffle bolt cracks) we intend to control reactor vessel flows and mechanical behaviour of internal structures. During normal operation, thermal-hydraulic is the main load of vessel internals. The current approach consists of acquiring the capacity to link different calculations, taking care that codes are qualified for physical phenomena and complex 3D geometries. For baffle assembly, a more simple model of this structure has been used to treat the physical phenomena linked to the LOCA transient. Results are encouraging mainly due to code capacity progression (resolution and models), which allows more and more complex physical phenomena to be treated, like turbulence flow and LOCA. (author)

  6. BWR vessel and internals project (BWRVIP)

    International Nuclear Information System (INIS)

    Bilanin, W.J.; Dyle, R.L.

    1996-01-01

    Recent Boiling Water Reactor (BWR) inspections indicate that Intergranular Stress Corrosion Cracking (IGSCC) is a significant technical issue for some BWR internals. IN response, the Boiling Water Reactor Vessel and Internals Project (BWRVIP) was formed by an associated of domestic and international utilities which own and operate BWRs. The project is identifying or developing generic, cost-effective strategies for managing degradation of reactor internals from which each utility can select the alternative most appropriate for their plant. The Electric Power Research Institute manages the technical program, implementing the utility defined programs. The BWRVIP is organized into four technical tasks: Assessment, Inspection, Repair and Mitigation. An Integration task coordinates the work. The goal of the Assessment task is to develop methodologies for evaluation of vessel and internal components in support of decisions for operation, inspection, mitigation or repair. The goal of the Inspection task is to develop and assess effective and predictable inspection techniques which can be used to determine the condition of BWR vessel and internals that are potentially susceptible to service-related SCC degradation. The goal of the Repair task is to assure the availability of cost-effective repair/replacement alternatives. The goal of the Mitigation task is to develop and demonstrate countermeasures for SCC degradation. This paper summarizes the BWRVIP approach for addressing BWR internals SCC degradation and illustrates how utilities are utilizing BWRVIP products to successfully manage the effect of SCC on core shrouds

  7. Nondestructive examination requirements for PWR vessel internals

    International Nuclear Information System (INIS)

    Spanner, J.

    2015-01-01

    This paper describes the requirements for the nondestructive examination of pressurized water reactor (PWR) vessel internals in accordance with the requirements of the EPRI Material Reliability Program (MRP) inspection standard for PWR internals (MRP-228) and the American Society of Mechanical Engineers Section XI In-service Inspection. The MRP vessel internals examinations have been performed at nuclear plants in the USA since 2009. The objective of the inspection standard is to provide the requirements for the nondestructive examination (NDE) methods implemented to support the inspection and evaluation of the internals. The inspection standard contains requirements specific to the inspection methodologies involved as well as requirements for qualification of the NDE procedures, equipment and personnel used to perform the vessel internals inspections. The qualification requirements for the NDE systems will be summarized. Six PWR plants in the USA have completed inspections of their internals using the Inspection and Evaluation Guideline (MRP-227) and the Inspection Standard (MRP-228). Examination results show few instances of service-induced degradation flaws, as expected. The few instances of degradation have mostly occurred in bolting

  8. Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

    International Nuclear Information System (INIS)

    Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

    1991-10-01

    This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

  9. Irradiation experiments on materials for core internals, pressure vessel and fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Tsukada, Takashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Materials degradation due to the aging phenomena is one of the key issues for the life assessment and extension of the light water reactors (LWRs). This presentation introduces JAERI`s activities in the field of LWR material researches which utilize the research and testing reactors for irradiation experiments. The activities are including the material studies for the core internals, pressure vessel and fuel cladding. These materials are exposed to the neutron/gamma radiation and high temperature water environments so that it is worth reviewing their degradation phenomena as the continuum. Three topics are presented; For the core internal materials, the irradiation assisted stress corrosion cracking (IASCC) of austenitic stainless steels is the present major concern. At JAERI the effects of alloying elements on IASCC have been investigated through the post-irradiation stress corrosion cracking tests in high-temperature water. The radiation embrittlement of pressure vessel steels is still a significant issue for LWR safety, and at JAERI some factors affecting the embrittlement behavior such as a dose rate have been investigated. Waterside corrosion of Zircaloy fuel cladding is one of the limiting factors in fuel rod performance and an in-situ measurement of the corrosion rate in high-temperature water was performed in JMTR. To improve the reliability of experiments and to extent the applicability of experimental techniques, a mutual utilization of the technical achievements in those irradiation experiments is desired. (author)

  10. Segmentation and packaging reactor vessels internals

    International Nuclear Information System (INIS)

    Boucau, Joseph

    2014-01-01

    Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

  11. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

  12. Internal Friction of Pressure Vessel Steel Embrittlement

    International Nuclear Information System (INIS)

    Van Ouytsel, K.

    2001-01-01

    The contribution consists of an abstract of a PhD thesis. The thesis contains a literature study, a description of the construction details of a new inverted torsion pendulum. This device was designed to investigate pressure-vessel steels at high amplitudes (10 -4 to 10 -2 ) and over a wide temperature range (90-700K) at approximately 1 Hz in the irradiated condition. Results of measurements on a variety of reactor pressure vessel steels by means of the torsion penduli are reported and interpreted

  13. Starting procedure for internal combustion vessels

    Science.gov (United States)

    Harris, Harry A.

    1978-09-26

    A vertical vessel, having a low bed of broken material, having included combustible material, is initially ignited by a plurality of ignitors spaced over the surface of the bed, by adding fresh, broken material onto the bed to buildup the bed to its operating depth and then passing a combustible mixture of gas upwardly through the material, at a rate to prevent back-firing of the gas, while air and recycled gas is passed through the bed to thereby heat the material and commence the desired laterally uniform combustion in the bed. The procedure permits precise control of the air and gaseous fuel mixtures and material rates, and permits the use of the process equipment designed for continuous operation of the vessel.

  14. Pressure vessel failure at high internal pressure

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1995-01-01

    A RPV failure due to plastic instability was investigated using the ABAQUS finite element code together with a material model of thermal plasticity for large deformations. Not only rotational symmetric temperature distributions were studied, but also 'hot spots'. Calculations show that merely by the depletion of strength of the material - even at internal wall temperatures well below the melting point of the fuel elements of about 2000/2400 C - the critical internal pressure can decrease to values smaller than the operational pressure of 16 Mpa. (orig.)

  15. The BWR vessel and internals project - 2001 and beyond

    International Nuclear Information System (INIS)

    Wagoner, V.; Mulford, T.

    2001-01-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  16. The BWR vessel and internals project - 2001 and beyond

    Energy Technology Data Exchange (ETDEWEB)

    Wagoner, V. [Carolina Power and Light, Progress Energy Building, NC (United States); Mulford, T. [Electric Power Research Institute, Palo Alto, CA (United States)

    2001-07-01

    The BWR Vessel and Internals Project (BWRVIP) is an international association of utilities owning boiling water reactors (BWRs). Figure 1 contains a list of current BWRVIP member utilities. The association was formed in 1994 with the following objectives: to lead the BWR industry toward generic resolution of reactor pressure vessel and internals material condition issues; to identify or develop generic, cost-effective strategies from which each operating plant will select the most appropriate alternative; to serve as the focal point for the regulatory interface with the industry on BWR vessel and internals issues; and to share information and promote communication and cooperation among participating utilities. The initial issue faced by the BWRVIP was core shroud cracking that had been observed in a number of BWRs. The BWRVIP reacted by quickly developing a set of industry guidelines to assist utilities in inspecting, evaluating, and, if necessary, repairing cracked shrouds. Subsequently, the BWRVIP pro-actively developed a comprehensive set of guidelines for managing degradation in other reactor internal components, including the reactor pressure vessel itself. The major components addressed by the BWRVIP are included. (author)

  17. Radiology trainer. Torso, internal organs and vessels. 2. ed.

    International Nuclear Information System (INIS)

    Staebler, Axel; Erlt-Wagner, Birgit

    2013-01-01

    The radiology training textbook is based on case studies of the clinical experience, including radiological imaging and differential diagnostic discussion. The scope of this volume covers the torso, internal organs and vessels. The following issues are discussed: lungs, pleura, mediastinum; heart and vascular system; upper abdomen organs; gastrointestinal tract; urogenital system.

  18. Vessel-related problems in severe accidents, International Research Projects

    International Nuclear Information System (INIS)

    Figueras, J. M.

    2000-01-01

    The paper describes those most relevant aspects of research programmes and projects, on the behavior of vessel during severe accidents with partial or total reactor core fusion, performed during the last twenty years or still on-going projects, by countries or international organizations in the nuclear community, presenting the most important technical aspects, in particular the results achieved, as well as the financial and organisational aspects. The paper concludes that, throughout a joint effort of the international nuclear community, in which Spain has been present via private and public organizations, actually exist a reasonable technical and experimental knowledge of the vessel in case of severe accidents, but still there are aspects not fully solved which are the basis for continuing some programmes and for proposal of new ones. (Author)

  19. Baffle-former arrangement for nuclear reactor vessel internals

    International Nuclear Information System (INIS)

    Machado, O.J.; Berringer, R.T.

    1978-01-01

    A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

  20. Optimization of reactor pressure vessel internals segmentation in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

    2017-11-15

    One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

  1. Fuel rod pressure in nuclear power reactors: Statistical evaluation of the fuel rod internal pressure in LWRs with application to lift-off probability

    Energy Technology Data Exchange (ETDEWEB)

    Jelinek, Tomas

    2001-02-01

    In this thesis, a methodology for quantifying the risk of exceeding the Lift-off limit in nuclear light water power reactors is outlined. Due to fission gas release, the pressure in the gap between the fuel pellets and the cladding increases with burnup of the fuel. An increase in the fuel-clad gap due to clad creep would be expected to result in positive feedback, in the form of higher fuel temperatures, leading to more fission gas release, higher rod pressure, etc, until the cladding breaks. An increase in the fuel-clad gap that leads to this positive feedback is a phenomenon called Lift-off and is a limitation that must be considered in the fuel core management. Lift-off is a consequence of very high internal fuel rod pressure. The internal fuel rod pressure is therefore used as a Lift-off indicator. The internal fuel rod pressure is closely connected to the fission gas release into the fuel rod plenum and is thus used to increase the database. It is concluded that the dominating error source in the prediction of the pressure in Boiling Water Reactors (BWR), is the power history. There is a bias in the fuel pressure prediction that is dependent on the fuel rod position in the fuel assembly for BWRs. A methodology to quantify the risk of the fuel rod internal pressure exceeding a certain limit is developed; the risk is dependent of the pressure prediction and the fuel rod position. The methodology is based on statistical treatment of the discrepancies between predicted and measured fuel rod internal pressures. Finally, a methodology to estimate the Lift-off probability of the whole core is outlined.

  2. Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

    Energy Technology Data Exchange (ETDEWEB)

    Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

    2013-07-01

    Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

  3. JAIF report on small- and medium-size LWRs

    International Nuclear Information System (INIS)

    Anon.

    1983-01-01

    The Ministry of International Trade and Industry has made a report on the results of the study on the utilization of small and medium size LWRs. The Japan Atomic Industrial Forum undertook this research work by the commission of the Ministry. It studied six cases of PWRs and BWRs with thermal output power ranging from 20 MWt to 700 MWt as the hypothetical design for their siting conditions, reactor capacity and the examples of utilization of the corresponding steam output. A large housing estate of 5,000 units, a local core city with population of 100,000, industrial utilization for petrochemical plants and paper and pulp plants and the combined utilization for a factory and a housing estate of 5,000 units were examined. The report said that a city with population of 100,000 would be the best object to supply heat from a nuclear reactor, and it is important to develop the nuclear reactors with 200 MWt capacity. This type is versatile in the siting conditions, and can be constructed near a city, accordingly, it offers the possibility of extensive utilization. The significance of small and medium size LWRs as the alternative energy for oil, the regional development accompanying the use of small and medium size LWRs, the technical development toward scale reduction, easy operation and versatile utilization, the economical evaluation of the utilization models and the tasks for the future are reported. (Kako, I.)

  4. Best practices for preparing vessel internals segmentation projects

    International Nuclear Information System (INIS)

    Boucau, Joseph; Segerud, Per; Sanchez, Moises

    2016-01-01

    Westinghouse has been involved in reactor internals segmentation activities in the U.S. and Europe for 30 years. Westinghouse completed in 2015 the segmentation of the reactor vessel and reactor vessel internals at the Jose Cabrera nuclear power plant in Spain and a similar project is on-going at Chooz A in France. For all reactor dismantling projects, it is essential that all activities are thoroughly planned and discussed up-front together with the customer. Detailed planning is crucial for achieving a successful project. One key activity in the preparation phase is the 'Segmentation and Packaging Plan' that documents the sequential steps required to segment, separate, and package each individual component, based on an activation analysis and component characterization study. Detailed procedures and specialized rigging equipment have to be developed to provide safeguards for preventing certain identified risks. The preparatory work can include some plant civil structure modifications for making the segmentation work easier and safer. Some original plant equipment is sometimes not suitable enough and need to be replaced. Before going to the site, testing and qualification are performed on full scale mock-ups in a specially designed pool for segmentation purposes. The mockup testing is an important step in order to verify the function of the equipment and minimize risk on site. This paper is describing the typical activities needed for preparing the reactor internals segmentation activities using under water mechanical cutting techniques. It provides experiences and lessons learned that Westinghouse has collected from its recent projects and that will be applied for the new awarded projects. (authors)

  5. Advanced thorium cycles in LWRs and HWRs

    International Nuclear Information System (INIS)

    Radkowsky, A.

    The main aspects of advanced thorium cycles in LWRs and HWRs are reviewed. New concepts include the seed blanket close packed heavy water breeder, the light water seed blanket thorium burner and self-induced thorium cycle in CANDU type reactors. (author)

  6. Analytical and experimental vibration analysis of BWR pressure vessel internals

    International Nuclear Information System (INIS)

    Krutzik, N.; Schad, O.

    1975-01-01

    This report attempts to evaluate the validity as well as quality of several analytical methods in the light of presently available experimental data for the internals of pressure vessels of boiling-water-reactor-types. The experimental checks were performed after the numerical analysis was completed and showed the accuracy of the numerical results. The analytical investigations were done by finite element programmes - 2-dimensional as well as 3-dimensional, where the effect of the mass distribution with parts of virtual masses on the dynamic response could be studied in depth. The experimental data were collected at various different plants and with different mass correlations. Besides evaluating the dynamic characteristics of the components, tests were also performed to evaluate the vibrations of the pressure vessel relative to the main structure. After analysing extensive recorded data much better understanding of the response under a variety of loading- and boundary conditions could be gained. The comparison of the results of analytical studies with the experimental results made a broad qualitative evaluation possible. (Auth.)

  7. Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

    Energy Technology Data Exchange (ETDEWEB)

    Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

  8. GCR dismantling: corrosion of vessel internals during decay storage

    International Nuclear Information System (INIS)

    Gras, J.M.

    1991-06-01

    Gas-cooled reactor decommissioning confronts EDF with the problem of the corrosion resistance of vessel internals over a decay storage period fixed at 50 years. The layer of magnetite previously formed in the C0 2 should protect structural steelwork from atmospheric corrosion. In any case, estimated steel corrosion after 50 years may be put at below or equal to 0.1 mm and the corresponding swelling induced by corrosion products at 0.2 mm. There should be no risk of hydrogen embrittlement or stress corrosion cracking of threaded fasteners. Corrosion tests aimed at providing further insight into the effects of the magnetite layer and a program for the surveillance of post-decommissioning structural corrosion should nevertheless be envisaged

  9. Computer system for International Reactor Pressure Vessel Materials Database support

    International Nuclear Information System (INIS)

    Arutyunjan, R.; Kabalevsky, S.; Kiselev, V.; Serov, A.

    1997-01-01

    This report presents description of the computer tools for support of International Reactor Pressure Vessel Materials Database developed at IAEA. Work was focused on raw, qualified, processed materials data, search, retrieval, analysis, presentation and export possibilities of data. Developed software has the following main functions: provides software tools for querying and search of any type of data in the database; provides the capability to update the existing information in the database; provides the capability to present and print selected data; provides the possibility of export on yearly basis the run-time IRPVMDB with raw, qualified and processed materials data to Database members; provides the capability to export any selected sets of raw, qualified, processed materials data

  10. Seismic Response Analysis of Assembled Reactor Vessel Internals

    International Nuclear Information System (INIS)

    Je, Sang-Yun; Chang, Yoon-Suk; Kang, Sung-Sik

    2015-01-01

    RVIs (Reactor Vessel Internals) perform important safe-related functions such as upholding the nuclear fuel assembly as well as providing the coolant passage of the reactor core and supporting the control rod drive mechanism. Therefore, the components including RVIs have to be designed and constructed taking into account the structural integrity under various accident scenarios. The reliable seismic analysis is essentially demanded to maintain the safe-related functions of RVIs. In this study, a modal analysis was performed based on the previous researches to examine values of frequencies, mode shapes and participation factors. Subsequently, the structural integrity respecting to the earthquake was evaluated through a response spectrum analysis by using the output variables of modal analysis. In this study, the structural integrity of the assembled RVIs was carried out against the seismic event via the modal and response spectrum analyses. Even though 287MPa of the maximum stress value occurred at the connected region between UGS and CSA, which was lower than its allowable value. At present, fluid-structure interaction effects are being examined for further realistic simulation, which will be reported in the near future

  11. Visual inspection of vessel internals; Visuelle Inspektion von Kerneinbauten

    Energy Technology Data Exchange (ETDEWEB)

    Rabe, G. [Siemens AG KWU, Erlangen (Germany)

    1999-08-01

    Visual inspection has matured to a qualified testing method and has become a standard method for inspection of reactor pressure vessels. Until today, all known defects in RPV internals have been detected by visual inspection. The codes KTA 3204 and DIN 25435-4 describe the framework conditions and requirements for visual inspections, which should be adhered to to the most possible extent. Visual inspections are carried by now at all RPV internals, also at those where access is difficult and limited. The inspection robot SUSI is applied in most cases. The camera and manipulator technology meanwhile has been upgraded to a standard performance quality allowing reliable, fast and easy visual inspection. The personnel is trained accordingly, so as to keep abreast with enhancements. Qualification of the inspection system has been simplified and standardised to a large extent. (orig/CB) [Deutsch] Die Sichtpruefung ist zu einem qualifizierten Pruefverfahren gereift und hat bei der Inspektion der RDB-Einbauten einen festen Platz eingenommen. Bisher wurden alle bekannten Schaeden an den RDB-Einbauten bei der Sichtpruefung festgestellt. In der KTA 3204 und der DIN 25435-4 sind die Rahmenbedingungen und Anforderungen an die Sichtpruefung beschrieben, die es gilt, weitestgehend einzuhalten. Mittlerweile werden an allen RDB-Einbauten, auch an den nur bedingt zugaenglichen, Sichtpruefungen vorgenommen. Dabei hat das Inspektionsfahrzeug SUSI inzwischen den breitesten Raum eingenommen. Die Entwicklung der Kamera- und Manipulatortechnik hat inzwischen einen Stand erreicht, der eine sichere, schnelle und einfache Sichtpruefung zulaesst. Das Pruefpersonal wird laufend fuer die Sichtpruefung geschult und qualifiziert. Die Qualifizierung des Inspektionssystems wurde weitestgehend vereinfacht und standardisiert. (orig.)

  12. General consideration of effective plutonium utilization in future LWRs

    International Nuclear Information System (INIS)

    Ishikawa, Nobuyuki; Okubo, Tsutomu

    2009-01-01

    In this study, the potential of mixed oxide fueled light water reactors (MOX-LWRs), especially focusing on the high conversion type LWRs (HC-LWRs) such as FLWR are evaluated in terms of both economic aspect and effective use of plutonium. For economics consideration, relative economics positions of MOX-LWRs are clarified comparing the cost of electricity for uranium fueled LWRs (U-LWRs), MOX-LWRs and fast breeder reactors (FBRs) assuming future natural uranium price raise and variation of parameters such as construction cost and capacity factor. Also the economic superiority of MOX utilization against the uranium use is mentioned from the view point of plutonium credit concerning to the front-end fuel cycle cost. In terms of effective use of plutonium, comparative evaluations on plutonium mass balance in the cases of HC-LWR and high moderation type LWRs (HM-LWRs) taking into account plutonium quality (ratio of fissile to total plutonium) constraint in multiple recycling are performed as representative MOX utilization cases. Through this evaluation, the advantageous features of plutonium multiple recycling by HC-LWR are clarified. From all these results, merits of the introduction of HC-LWRs are discussed. (author)

  13. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, K.I.; Kim, B.S.; Kim, D.H. [Korea Atomic Energy Research Inst., Thermal Hydraulic Safety Research, Taejon (Korea, Republic of)

    2001-07-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  14. Effect of the in- and ex-vessel dual cooling on the retention of an internally heated melt pool in a hemispherical vessel

    International Nuclear Information System (INIS)

    Ahn, K.I.; Kim, B.S.; Kim, D.H.

    2001-01-01

    A concept of in-vessel melt retention (IVMR) by in-vessel reflooding and/or reactor cavity flooding has been considered as one of severe accident management strategies and intensive researches to be performed worldwide. This paper provides some results of analytical investigations on the effect of both in- / ex-vessel cooling on the retention of an internally heated molten pool confined in a hemispherical vessel and the related thermal behavior of the vessel wall. For the present analysis, a scale-down reactor vessel for the KSNP reactor design of 1000 MWe (a large dry PWR) is utilized for a reactor vessel. Aluminum oxide melt simulant is also utilized for a real corium pool. An internal power density in the molten pool is determined by a simple scaling analysis that equates the heat flux on the the scale-down vessel wall to that estimated from KSNP. Well-known temperature-dependent boiling heat transfer curves are applied to the in- and ex-vessel cooling boundaries and radiative heat transfer has been only considered in the case of dry in-vessel. MELTPOOL, which is a computational fluid dynamics (CFD) code developed at KAERI, is applied to obtain the time-varying heat flux distribution from a molten pool and the vessel wall temperature distributions with angular positions along the vessel wall. In order to gain further insights on the effectiveness of in- and ex-vessel dual cooling on the in-vessel corium retention, four different boundary conditions has been considered: no water inside the vessel without ex-vessel cooling, water inside the vessel without ex-vessel cooling, no water inside the vessel with ex-vessel cooling, and water inside the vessel with ex-vessel cooling. (authors)

  15. Lining up device for the internal structures of a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Silverblatt, B.L.

    1977-01-01

    The invention concerns a nuclear reactor of the type with a vessel, a vessel head carried at the top of this vessel by a core cylinder comprising a flange internally supported by the vessel, and an upper support structure supported between the core cylinder flange and the vessel head to align laterally the head, vessel, flange and support structure. A bottom key device is provided for lining up the flange, support structure and vessel, and an upper key device for laterally lining up support structure and the vessel head and for maintaining this alignment when they are removed simultaneously from the core cylinder and vessel. When re-assembling the reactor, the top support structure and the vessel head are lowered simultaneously so that an opening in the top alignment structure engages in the upper extension of the bottom alignment structure. A plurality of alignment stuctures may be utilised round the circumference of the reactor vessel. The disposition of the invention also facilitates the removal of the core cylinder from the reactor vessel. In this way, the alignment on re-assembly is ensured by the re-entry of the bottom extension under the flange of the core cylinder with the groove or keyway of the reactor vessel [fr

  16. Components of the primary circuit of LWRs

    International Nuclear Information System (INIS)

    1980-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives. (orig.) [de

  17. Reducing occupational radiation exposures at LWRs

    International Nuclear Information System (INIS)

    Lattanzi, D.; Neri, C.; Papa, C.; Paribelli, S.

    1980-01-01

    The occupational radiation doses received by nuclear power plant personnel during a period of several years of operation are briefly reviewed. Comparisons are made between the data for BWRs and PWRs in order to identify the more ''critical'' reactor type, from a radiological poin; of view. Attention is also devoted to GCRs. Furthermore the areas which contribute most to personnel doses are considered and briefly reviewed. The main steps to be taken in order to reduce occupational radiation exposures at LWRs are discussed. (H.K.)

  18. Recycling of MOX fuel for LWRs

    International Nuclear Information System (INIS)

    Joo, Hyung Kook; Oh, Soo Youl

    1992-01-01

    The status and issues related to the thermal recycling of reprocessed nuclear fuels have been reviewed. It is focused on the use of reprecessed plutonium in the form of mixed oxide (MOX) for a light water reactor and the review on reprocessing and fabrication processes is beyond the scope. In spite of the difference in the nuclear characteristics between plutonium and uranium isotopes, the neutronics behavior in a core with MOX fuels is similar to that with normal uranium fuels. However, since the neutron spectrum is hardened in a core with MOX, the Doppler, viod, and moderator temperature coefficients become more negative and the control rod and boron worths are slightly reduced. Therefore, the safety will be evaluated carefully in addition to the core neutronics analysis. The MOX fuel rod behavior related to the rod performance such as the pellet to clad interaction and fission gas release is also similar to that of uranium rods, and no specific problem arises. Substituting MOX fuels for a portion of uranium fuels, it is estimated that the savings be about 25% in uranium ore and 10% in uranium enrichment service requirements. The use of MOX fuel in LWRs has been commercialized in European countries including Germany, France, Belgium, etc., and a demonstration program has been pursued in Japan for the commercial utilization in the late 1990s. Such a worldwide trend indicates that the utilization of MOX fuel in LWRs is a proven technology and meets economics criteria. (Author)

  19. Guiding device for a manipulator mast for internal inspection of a reactor pressure vessel

    International Nuclear Information System (INIS)

    Seifert, W.; Schlueter, H.

    1977-01-01

    A remote-controlled supporting device centering a manipulator mast is described which is mounted and operated above a reactor pressure vessel under water in such a way that rotations and vertical movements necessary for the internal inspection of the pressure vessel remain possible. (RW) [de

  20. Sampling of reactor pressure vessel and core internals

    International Nuclear Information System (INIS)

    Oberhaeuser, Ralf

    2012-01-01

    Decommissioning and dismantling of nuclear power plants is a growing business as a huge number of plants built in the 1970's have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Beside the dismantling works for itself, the disposal of activated components and other nuclear waste is very expensive. What comes next is the fact that final disposal facilities are not available yet in most countries meaning a need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. AREVA is proud of versatile experience in successfully dismantling nuclear components like core internals and reactor pressure vessel (RPV). The basis of a well-founded and optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be and in the best case a 3D activation- model. For keeping the necessary sampling effort as small as possible, but simultaneously as efficient as possible, representative sampling positions are defined in advance by theoretical radiological examinations. For this, a detailed 3D-CAD-model of the components to be dismantled has proven very helpful and effective. Under these aspects a sampling of RPV and its components is necessary to verify the theoretically calculated radiological data. The obtained results of activation and contamination are taken into account for the optimized dismantling and packaging strategy. The precise 3D-activation-model will reduce the necessary number and type of final disposal containers as security factors are minimized leading to a lower shielding effort, too. Besides, components or even parts of components may be subject of release measurement. In the end, costs can be reduced. In this context

  1. Cost reduction of LWRs - The main features

    International Nuclear Information System (INIS)

    Board, J.A.; Norman, D.

    1991-01-01

    For effective generation cost reductions to be achieved while maintaining safety levels, the impact of each of the above element and their interactions must be fully assessed, together with the effects of uncertainty on commercial risk. the amount of interest during construction which must be paid can be minimised by adopting designs, output ratings (unit size), and construction methods which minimise construction time. However this must be done without unduly increasing specific capital cost. Total capital costs can be reduced by sharing design and project launching costs and front-end design and licensing costs, across a series of identical plants. The paper is reviewing costs and performance factors such as those above with the aim of identifying the strategies which might be necessary within and between countries in order to create an environment which would enable cost reductions on LWRs to be made. (author)

  2. Verification and Validation Strategy for LWRS Tools

    Energy Technology Data Exchange (ETDEWEB)

    Carl M. Stoots; Richard R. Schultz; Hans D. Gougar; Thomas K Larson; Michael Corradini; Laura Swiler; David Pointer; Jess Gehin

    2012-09-01

    One intension of the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to create advanced computational tools for safety assessment that enable more accurate representation of a nuclear power plant safety margin. These tools are to be used to study the unique issues posed by lifetime extension and relicensing of the existing operating fleet of nuclear power plants well beyond their first license extension period. The extent to which new computational models / codes such as RELAP-7 can be used for reactor licensing / relicensing activities depends mainly upon the thoroughness with which they have been verified and validated (V&V). This document outlines the LWRS program strategy by which RELAP-7 code V&V planning is to be accomplished. From the perspective of developing and applying thermal-hydraulic and reactivity-specific models to reactor systems, the US Nuclear Regulatory Commission (NRC) Regulatory Guide 1.203 gives key guidance to numeric model developers and those tasked with the validation of numeric models. By creating Regulatory Guide 1.203 the NRC defined a framework for development, assessment, and approval of transient and accident analysis methods. As a result, this methodology is very relevant and is recommended as the path forward for RELAP-7 V&V. However, the unique issues posed by lifetime extension will require considerations in addition to those addressed in Regulatory Guide 1.203. Some of these include prioritization of which plants / designs should be studied first, coupling modern supporting experiments to the stringent needs of new high fidelity models / codes, and scaling of aging effects.

  3. Outlines of guidelines for the inspection and evaluation of reactor vessel internals

    International Nuclear Information System (INIS)

    Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

    2014-01-01

    'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

  4. Computing radiation dose to reactor pressure vessel and internals

    International Nuclear Information System (INIS)

    1996-01-01

    Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

  5. Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

    2008-09-15

    This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

  6. Development of glia and blood vessels in the internal capsule of rats.

    Science.gov (United States)

    Earle, K L; Mitrofanis, J

    1998-02-01

    We have explored two aspects of internal capsule development that have not been described previously, namely, the development of glia and of blood vessels. To these ends, we used antibodies to glial fibrillary acidic protein (GFAP) and to vimentin (to identify astrocytes and to radial glia) and Griffonia simplicifolia (lectin; to identify microglia and blood vessels). Further, we made intracardiac injections of Evans Blue to examine the permeability of this dye in the vessels of the internal capsule during neonatal development. Our results show that large numbers of radial glia, astrocytes and microglia are not labelled with these markers in the white matter of the internal capsule until about birth; very few are labelled earlier, during the critical stages of corticofugal and corticopetal axonal ingrowth (E15-E20). The large glial labelling in the internal capsule at birth is accompanied by a dense vascular innervation of the capsule; as with the glia, very few labelled patent vessels are seen earlier. After intracardiac injections of Evans Blue, we find that the blood vessels of the internal capsule are not particularly permeable to Evans Blue. At each age examined (P0, P5, P15), blood vessels are outlined very clearly and there is no diffuse haze of fluorescence within the extracellular space, which is indicative of a leaky vessel. There are three striking differences between the glial environment of the internal capsule and that of the adjacent thalamus. First, the internal capsule is never rich with radial glial fibres (vimentin- and GFAP-immunoreactive) during development (except at P0), whereas the thalamus has many radial fibres from very early development (E15-E17). Second, astrocytes (vimentin- and GFAP-immunoreactive) first become apparent in the internal capsule (E20-P0) well before they do in the thalamus (P15). Third, the internal capsule houses a large transient population of amoeboid microglia (P0-P22), whereas the thalamus does not; only ramified

  7. Age-related degradation of boiling water reactor vessel internals

    International Nuclear Information System (INIS)

    Ware, A.G.; Shah, V.N.

    1992-01-01

    Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

  8. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1994-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (74-90 mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments. Under severe accident loading conditions, the steel containment vessel in a typical Mark-I or Mark-II plant may deform under internal pressurization such that it contacts the inner surface of a shield building wall. (Thermal expansion from increasing accident temperatures would also close the gap between the SCV and the shield building, but temperature effects are not considered in these analyses.) The amount and location of contact and the pressure at which it occurs all affect how the combined structure behaves. A preliminary finite element model has been developed to analyze a model of a typical steel containment vessel con-ling into contact with an outer structure. Both the steel containment vessel and the outer contact structure were modelled with axisymmetric shell finite elements. Of particular interest are the influence that the contact structure has on deformation and potential failure modes of the containment vessel. Furthermore, the coefficient of friction between the two structures was varied to study its effects on the behavior of the containment vessel and on the uplift loads transmitted to the contact structure. These analyses show that the material properties of an outer contact structure and the amount

  9. Lifetime assessment on PWR reactor vessel internals in Korea

    International Nuclear Information System (INIS)

    Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

    2002-01-01

    In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

  10. Analyses of a steel containment vessel with an outer contact structure under severe internal overpressurization conditions

    International Nuclear Information System (INIS)

    Porter, V.L.

    1993-01-01

    Many Mark-I and Mark-II BWR plants are designed with a steel vessel as the primary containment. Typically, the steel containment vessel (SCV) is enclosed within a reinforced concrete shield building with only a small gap (50--90mm) separating the two structures. This paper describes finite element analyses performed to evaluate the effects of contact and friction between a steel containment vessel and an outer contact structure when the containment vessel is subjected to large internal pressures. These computations were motivated by a joint program on containment integrity involving the Nuclear Power Engineering Corporation (NUPEC) of Japan, the US Nuclear Regulatory Commission (NRC), and Sandia National Laboratories for testing model containments

  11. Inspection and repair of reactor pressure vessel (RPV) internals

    International Nuclear Information System (INIS)

    Bohmann, W.; Poetz, F.; Nicolai, M.

    1996-01-01

    The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

  12. Safeguards on MOX assemblies at LWRs

    International Nuclear Information System (INIS)

    Arenas Carrasco, J.; Koulikov, I.; Heinonen, O.J.; Arlt, R.; Grigoleit, K.; Clarke, R.; Swinhoe, M.

    2000-01-01

    Operating within the framework of the New Partnership Approach (NPA) for unirradiated MOX fuel assemblies in LWRs, the IAEA and EURATOM have gained experience in safeguarding 13 LWRs licensed to operate with MOX assemblies. In order to fulfil SIR requirements, verification methods and techniques capable of measuring MOX assemblies under water have been and are still being developed. These encompass both qualitative tests for the detection of plutonium (gross attribute tests) and quantitative tests for the measurement of the amount of plutonium (partial defect tests) and are based on gamma and neutron detection techniques. There are nine PWR and two BWR where the reactor and the spent fuel pond can be covered by the same surveillance device. These are Type I reactors where the reactor and the pond are located in the same hall. In these types of facilities relying on surveillance during the MOX refuelling is especially difficult at the BWRs due to the depth of the core pond. There are two PWR type facilities where the reactor and the spent fuel pond are located in different halls and cannot be covered by the same surveillance device (Type II). An open core camera has not been installed during refuelling and therefore indirect surveillance is currently used to survey MOX loading. Improvements are therefore required and are under consideration. After receipt at the facility, there are a few facilities which must keep the received fresh MOX fuel in wet storage, not only for a short period prior to refuelling, but for more than a year, until the next refuelling campaign. In these cases timely inspections for direct use fresh nuclear material require considerable inspection effort. Additionally, where human surveillance of core loading and finally core closure are necessary there is also a large demand for manpower. Either an agreement should be reached with the operators to delay the MOX loading until the end of the fuelling campaign, or alternative approaches should be

  13. Assessment of integrity for the pressure vessel internals of PWRs under blowdown loadings

    International Nuclear Information System (INIS)

    Geiss, M.; Benner, J.; Ludwig, A.

    1984-01-01

    In safety analysis of pressurized water reactors the loss-of-coolant accident plays a central role. Thereby a sudden break of a cold primary coolant pipe close to the reactor pressure vessel is postulated. The sudden pressure release of the primary system (blowdown) causes high dynamic loading on the pressure vessel internals. The resulting deformations must not impair shut down of the reactor and decay heat removal in an inadmissible way. For this assessment a blowdown analysis for a 1300 MW pressurized water reactor is carried out. These investigations are completed with a detailed stress analysis for the highly loaded core barrel clamping. The results show that the reactor pressure vessel internals are able to withstand blowdown loading. Even in case of a sudden and complete break of the primary coolant pipe the loading has to be twice as high to endanger the structural integrity. (orig.) [de

  14. A study on detection of internal defects of pressure vessel by digital shearography

    International Nuclear Information System (INIS)

    Kang, Young Jun; Park, Sung Tae; Lee, Hae Moo; Nam, Seung Hun

    1999-01-01

    Pipelines in power plants, nuclear facilities and chemical industries are often affected by corrosion effects. The inspection of internal defects of these pipelines is important to guarantee safe operational condition. Conventional NDT methods have been taken relatively much time, money, and manpower because of performing as the method of contact with objects to be inspected. Digital shearography is a laser-based optical method which allows full-field observation of surface displacement derivatives. This method has many advantages in practical use, such as low sensitivity to environmental noise, simple optical configuration and real time measurement. Therefore it is a good method to use for detecting internal defects. In this paper, the experiment was performed with some pressure vessels which has different internal cracks. We detected internal cracks of the pressure vessels at a real time and evaluated qualitatively these results. We also performed qualitative measurement of shearo fringe by using phase shifting method.

  15. 46 CFR 27.205 - What are the requirements for internal communication systems on towing vessels?

    Science.gov (United States)

    2010-10-01

    ... systems on towing vessels? 27.205 Section 27.205 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY... fitted with a communication system between the engine room and the operating station that— (1) Consists... required to have internal communication systems. (c) When the operating-station's engine controls and the...

  16. Safety design philosophy of the ABWR for the next generation LWRs

    International Nuclear Information System (INIS)

    Sato, Takashi; Akinaga, Makoto; Kojima, Yoshihiro

    2009-01-01

    The paper presents safety design philosophy of the advanced boiling water reactor (ABWR) to be reflected in developing the next generation light water reactors (LWRs). The basic policy of the ABWR safety design was to improve safety and reduce cost simultaneously by reflecting lessons learned of precursors, incidents and accidents that were beyond the design basis such as the Three Mile Island Unit 2 (TMI 2) accident. The ABWR is a fully active safety plant. The ABWR enhanced redundancy and diversity of active safety systems using probabilistic safety assessment (PSA) insights. It adopted a complete three division active emergency core cooling system (ECCS) and attained a very low core damage frequency (CDF) value of less than 10 -7 /ry for internal events. Only very small residual risks, if any, rather exist in external events such as an extremely large earthquake beyond the design basis. This is because external events can constitute a common cause that disables all the redundant active safety systems. Therefore, it is useless to add one more ECCS train and make a four division active ECCS for external events. Nowadays, however, fully passive safety LWRs are already established. Incorporating some of these passive safety systems we can also establish the next generation LWRs that are truly strong against external events. We can establish a plant that can survive a giant earthquake at least three days without AC power source, SA proof safety design that enables no containment failure and no evacuation to eliminate the residual risks. The same basic policy as the ABWR to improve safety and reduce cost simultaneously is again effective for the next generation LWRs. (author)

  17. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    International Nuclear Information System (INIS)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y.

    2014-01-01

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10 22 to 3*10 24 n/m 2 depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties), most of the

  18. Strategic Assessment of Causes, Impacts and Mitigation of Radiation Embrittlement of RPV steel in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Shamim, Jubair Ahmed; Bhowmik, Palash Kumar; Gairola, Abhinav; Suh, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of)

    2014-10-15

    Nuclear power has been emerged as a proven technology in the present day world to beget electricity after its first successful demonstration in 1942. Due to world's increasing concern over the augmented concentration of 'Greenhouse Gas' emissions primarily caused by burning of fossil fuel, it is not surprising that there will be a galloping demand for nuclear power in near future. As per data of World Nuclear Association, there are currently 435 operable civil nuclear power reactors around the world, with a further 71 under construction, among which the most common type is LWR. Pressure vessel of LWR is the most vital pressure boundary component of Nuclear Steam Supply System (NSSS) as it houses the core under elevated pressure and temperature. It also provides structural support to RPV internals and attempts to protect against possible rupture under all postulated transients that the NSSS may undergo. LWR pressure vessel experiences service at a temperature of 250-320 .deg. C and receives significant level of fast neutron fluence, ranging from about 5*10{sup 22} to 3*10{sup 24} n/m{sup 2} depending on plant design. There are also differences in materials used for various designed reactors. Weldments also vary in type and impurity level. Accordingly, the assessment of degradation of major components such as RPV steel caused by aging and corrosion is a common objective for safe operation of all LWRs. The purpose of this paper is to assess how neutron irradiation contributes to the degradation of mechanical properties of RPV steel and how these effects can be minimized. Since RPV is the only irreplaceable component in NPPs, the degradation of mechanical properties of RPV is the life-limiting feature of LWR nuclear power plant operation. Although there are a number of ways (e.g. thermal neutrons, fast neutrons and gamma-ray irradiation) that may contribute to the displacement of atoms (hence RPV embrittlement and degradation of mechanical properties

  19. 50 CFR 216.46 - U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation...

    Science.gov (United States)

    2010-10-01

    ... 50 Wildlife and Fisheries 7 2010-10-01 2010-10-01 false U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation Program. 216.46 Section 216.46 Wildlife and Fisheries....46 U.S. citizens on foreign flag vessels operating under the International Dolphin Conservation...

  20. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    The TEMP-STRESS FEM represents an axisymmetric simulation of the reinforced concrete vessel to internal pressurization. The information shows the global deformation, the state of strain/stress within the containment vessel with respect to the imposed pressures. Thus, the location and progress of concrete cracking, the stretching of the liner and the reinforcing bars and final failure are indicated through the entire loading range. Equilibrium of the entire system is assured at definite loading increments. With the progress of concrete cracking, the resisting load is continuously transferred to the reinforcing bars and the liner. Thus, after the tensile strength is exceeded and the concrete stress is set to zero, the internal pressures are entirely resisted by the liner and the reserve strength of the reinforcing bars. The reinforcing bars are mechanically connected to each other by splices, the ultimate strength of which is less than that of the rebars themselves. The corresponding strain at this limiting stress is lower than the ultimate strain of the liner. Therefore, the specified ultimate strength of the splices limits the pressurization of the vessel. Furthermore, once any of the splices fail, then load is transferred to the adjacent members, causing their failure and general failure of the vessel. (orig./HP)

  1. Experimental and theoretical investigation on the depressurization of a vessel with internals

    International Nuclear Information System (INIS)

    Vigni, P.; Oriolo, F.; Rosa, U.

    1978-01-01

    This paper is about some blow-down experiments performed at the Scalbatraio Center of the University of Pisa. The blow-down tests have been made to investigate the depressurization of a vessel with internal structures, reproducing the geometry of a BWR. The experimental data have been compared with calculations performed by the RELAP program, in order to evaluate the scaling effects related to their application to large scale units. (author)

  2. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

    International Nuclear Information System (INIS)

    2010-01-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  3. Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

    International Nuclear Information System (INIS)

    Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

    1987-01-01

    On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

  4. Load bearing capacities and elastic-plastic behavior of reactor vessel internals

    International Nuclear Information System (INIS)

    Watanabe, Keita; Nagase, Ryuichi

    2017-01-01

    Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

  5. Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

    Science.gov (United States)

    Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

    2003-06-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. The measurements were made using Red Perspex™ polymethyl methacrylate high-level film dosimeters, a Radcal "peanut" ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

  6. RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR

    International Nuclear Information System (INIS)

    HOLDEN, N.E.; RECINIELLO, R.N.; HU, J.P.; RORER, D.C.

    2002-01-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex(trademark) polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup

  7. International Cooperation for the Dismantling of Chooz A Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Grenouillet, J.J.; Posivak, E.

    2009-01-01

    Chooz A is the first PWR that is being decommissioned in France. The main issue that is conditioning the success of the project is the Reactor Pressure Vessel (RPV) and Reactor Vessel Internals (RVI) segmentation. Whereas Chooz A is the first and unique RPV and RVI being dismantled in France, there are many similar experiences available in the world. Thus the project team was eager to cooperate with other teams facing or being faced with the same issue. A cooperation programme was established in two separate ways: - Benefiting from experience feedback from completed RPV and RVI dismantling projects, - Looking for synergy with future RPV dismantling projects for activities such as segmentation tools design, qualification and manufacturing for example. This paper describes the implementation of this programme and how the outcome of the cooperation was used for the implementation of Chooz-A RPV and RVI segmentation project. It shows also the limits of such a cooperation. (authors)

  8. Complete resection of locally advanced ovarian carcinoma fixed to the pelvic sidewall and involving external and internal iliac vessels.

    Science.gov (United States)

    Nishikimi, Kyoko; Tate, Shinichi; Matsuoka, Ayumu; Shozu, Makio

    2017-08-01

    Locally advanced ovarian carcinomas may be fixed to the pelvic sidewall, and although these often involve the internal iliac vessels, they rarely involve the external iliac vessels. Such tumors are mostly considered inoperable. We present a surgical technique for complete resection of locally advanced ovarian carcinoma fixed to the pelvic sidewall and involving external and internal iliac vessels. A 69-year-old woman presented with ovarian carcinoma fixed to the right pelvic sidewall, which involved the right external and internal iliac arteries and veins and the right lower ureter, rectum, and vagina. We cut the external iliac artery and vein at the bifurcation and at the inguinal ligament to resect the external artery and vein. Then, we reconstructed the arterial and venous supplies of the right external artery and vein with grafts. After creating a wide space immediately inside of the sacral plexus to allow the tumor fixed to pelvic sidewall with the internal iliac vessels to move medially, we performed total internal iliac vessel resection. We achieved complete en bloc tumor resection with the right external and internal artery and vein, right ureter, vagina, and rectum adhering to the tumor. There were no intra- or postoperative complications, such as bleeding, graft occlusion, infection, or limb edema. Exfoliation from the sacral plexus and total resection with external and internal iliac vessels enables complete resection of the tumor fixed to the pelvic sidewall. Copyright © 2017 Elsevier Inc. All rights reserved.

  9. General Description of the Mechanic Design of the Pressure Vessel and the Internal Mechanical Component of the CAREM Reactor

    International Nuclear Information System (INIS)

    Diez, F.; Horro, R.

    2000-01-01

    This paper presents a brief description of the CAREM reactor pressure vessel and its main internal mechanical components and summarizes the functional requirements and approaches applied for their design, together with a review of the normative applicable in each case

  10. PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

    International Nuclear Information System (INIS)

    Schwirian, R.; Robison, G.

    1994-07-01

    The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

  11. Development of advanced design features for KNGR reactor vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-12-01

    Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

  12. Development of advanced design features for KNGR reactor vessel and internals

    International Nuclear Information System (INIS)

    Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

    1995-12-01

    Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

  13. Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

    International Nuclear Information System (INIS)

    Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

    2012-01-01

    With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

  14. An overview of reactor vessel internals segmentation for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Litka, T.J.

    1994-01-01

    Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

  15. Fast neutron fluence calculations as support for a BWR pressure vessel and internals surveillance program

    International Nuclear Information System (INIS)

    Lucatero, Marco A.; Palacios-Hernandez, Javier C.; Ortiz-Villafuerte, Javier; Xolocostli-Munguia, J. Vicente; Gomez-Torres, Armando M.

    2010-01-01

    Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are

  16. Development of Safety Review Guide for the Periodic Safety Review of Reactor Vessel Internals

    International Nuclear Information System (INIS)

    Park, Jeongsoon; Ko, Hanok; Kim, Seonjae; Jhung, Myungjo

    2013-01-01

    Aging management of the reactor vessel internals (RVIs) is one of the important issues for long-term operation of nuclear power plants (NPPs). Safety review on the assessment and management of the RVI aging is conducted through the process of a periodic safety review (PSR). The regulatory body should check that reactor facilities sustain safety functions in light of degradation due to aging and that the operator of a nuclear power reactor establishes and implements management program to deal with degradation due to aging in order to guarantee the safety functions and the safety margin as a result of PSR. KINS(Korea Institute of Nuclear Safety) has utilized safety review guides (SRG) which provide guidance to KINS staffs in performing safety reviews in order to assure the quality and uniformity of staff safety reviews. The KINS SRGs for the continued operation of pressurized water reactors (PWRs) published in 2006 contain areas of review regarding aging management of RVIs in chapter 2 (III.2.15, Appendix 2.0.1). However unlike the SRGs for the continued operation, KINS has not officially published the SRGs for the PSR of PWRs, but published them as a form of the research report. In addition to that, the report provides almost same review procedures for aging assessment and management of RVIs with the ones provided in the SRGs for the continued operation, it cannot provide review guidance specific to PSRs. Therefore, a PSR safety review guide should be developed for RVIs in PWRs. In this study, a draft PSR safety review guide for reactor vessel internals in PWRs is developed and provided. In this paper, a draft PSR safety review guide for reactor vessel internals (PSR SRG-RVIs) in PWRs is introduced and main contents of the draft are provided. However, since the PSR safety review guides for areas other than RVIs in the pressurized water reactors (PWRs) are expected to be developed in the near future, the draft PSR SRG-RVIs should be revisited to be compatible with

  17. Experience in dismantling and packaging of pressure vessel and core internals

    International Nuclear Information System (INIS)

    Pillokat, Peter; Bruhn, Jan Hendrik

    2011-01-01

    Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA completed the order for dismantling of all remaining Reactor Pressure Vessel internals at German Boiling Water Reactor Wuergassen NPP in October '08. During the onsite activities about 121 tons of steel were successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA dismantled the cylindrical part of the Wuergassen Pressure Vessel. During this project approximately 320 tons of steel were cut and packaged for final disposal, as dismantling was mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment was the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this, an elevator was installed to transport cut segments into the turbine hall, where a special housing was built for final storage conditioning. At the beginning of 2007, another complex dismantling project of great importance was acquired by AREVA. The contract included dismantling and conditioning for final storage of the complete RPV Internals of the German Pressurized Water Reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. On-site activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of Core Internals were successfully dismantled at Stade NPP until September '09. To accomplish the best possible on

  18. Ion transport membrane module and vessel system with directed internal gas flow

    Science.gov (United States)

    Holmes, Michael Jerome; Ohrn, Theodore R.; Chen, Christopher Ming-Poh

    2010-02-09

    An ion transport membrane system comprising (a) a pressure vessel having an interior, an inlet adapted to introduce gas into the interior of the vessel, an outlet adapted to withdraw gas from the interior of the vessel, and an axis; (b) a plurality of planar ion transport membrane modules disposed in the interior of the pressure vessel and arranged in series, each membrane module comprising mixed metal oxide ceramic material and having an interior region and an exterior region; and (c) one or more gas flow control partitions disposed in the interior of the pressure vessel and adapted to change a direction of gas flow within the vessel.

  19. MOX fuel fabrication and utilisation in LWRs worldwide

    International Nuclear Information System (INIS)

    Provost, J.-L.; Schrader, M.; Nomura, S.

    2000-01-01

    Early in the development of the nuclear programme, a large part of the countries using nuclear energy has studied the reprocessing and recycling option in order to develop a safe conditioning of fission products and to recycle fissile materials in reactors. In the sixties, the feasibility of recycling plutonium in LWRs has been successfully demonstrated by several experimentations of MOX rod irradiations in different countries. Based on the background of the MOX behaviour collected during the seventies and on the results of the important MOX experimentation program implemented during this period, a large part of the European utilities decided at the beginning of the eighties to use MOX fuel in LWRs on an industrial scale. The main goals of the utilities were to use as a fuel an available fissile material and to control the stockpile of separated plutonium. Today, the understanding of the behaviour of plutonium fuel has grown significantly since the launch of the first R and D programmes on LWR and FR MOX fuels. Plutonium oxide physical and neutron behaviour is well known, its modelling is now available as well as experimentally validated. Up to now, more than 750 tHM MOX fuel (more than 2000 FAs) have been loaded in 29 PWRs and in 2 BWRs in Europe, corresponding to the recycling of about 35 t of plutonium. Reprocessing/recycling technology has reached maturity in the main nuclear industry countries. Spent fuel reprocessing and recycling of the separated fissile materials remains the main option for the back-end cycle. Today, the operation of MOX-recycling LWRs is considered satisfactory. Experience feedback shows that, in global terms, MOX cores behaviour is equivalent to that of UO 2 cores in terms of operation and safety. (author)

  20. An internal-friction study of reactor-pressure-vessel steel embrittlement

    International Nuclear Information System (INIS)

    Ouytsel, K. van; Fabry, A.; Batist, R. de; Schaller, R.

    1997-01-01

    Within an enhanced commercial surveillance strategy, the nuclear-research institute SCK.CEN in Mol, Belgium is investigating, by means of internal friction, the microstructural processes responsible for embrittlement of pressure-vessel steels. The experiments were carried out using a torsion pendulum at the Ecole Polytechnique Federale de Lausanne in Switzerland. Amplitude-independent internal-friction experiments teach us that neutron irradiation induces defects which interact with mobile dislocations. Thermal ageing of JRQ and Doel-IV steel does not cause major embrittlement effects. Amplitude-dependent internal-friction experiments allow us to determine a critical amplitude which corresponds to the yield stress of the material as obtained from static tensile tests. The results also correspond to a three-component model for the yield strength taking into account both hardening and non-hardening embrittlement. Investigations of Doel-I-II weld material in different conditions reveal that embrittlement due to irradiation or thermal ageing can be interpreted in terms of a fine interplay between long- and short-range phenomena. (author)

  1. Dr. Mainte. Integrated simulator of maintenance optimization of LWRs

    International Nuclear Information System (INIS)

    Isobe, Yoshihiro; Sagisaka, Mitsuyuki; Etoh, Junji; Matsunaga, Takashi; Kosaka, Toru; Matsumoto, Satoshi; Yoshimura, Shinobu

    2014-01-01

    Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) has been developed based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for representative components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability and economic efficiency, environmental impact and social acceptance. For the further improvement of the safety and availability, the effect of human error and its reduction on the optimization of plant maintenance activities and approaches of reducing it have been studied. (author)

  2. Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

    International Nuclear Information System (INIS)

    Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

    2010-01-01

    Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

  3. Production and quality control of fuel cladding tubes for LWRs

    International Nuclear Information System (INIS)

    Matsuda, Katsuhiko; Hagi, Shigeki; Anada, Hiroyuki; Abe, Hideaki; Hyodo, Shigetoshi

    1994-01-01

    This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

  4. Radiological characterization of the pressure vessel internals of the BNL High Flux Beam Reactor.

    Science.gov (United States)

    Holden, Norman E; Reciniello, Richard N; Hu, Jih-Perng

    2004-08-01

    In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, measurements and calculations of the decay gamma-ray dose-rate were performed in the reactor pressure vessel and on vessel internal structures such as the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. Measurements of gamma-ray dose rates were made using Red Perspex polymethyl methacrylate high-dose film, a Radcal "peanut" ion chamber, and Eberline's RO-7 high-range ion chamber. As a comparison, the Monte Carlo MCNP code and MicroShield code were used to model the gamma-ray transport and dose buildup. The gamma-ray dose rate at 8 cm above the center of the Transition Plate was measured to be 160 Gy h (using an RO-7) and 88 Gy h at 8 cm above and about 5 cm lateral to the Transition Plate (using Red Perspex film). This compares with a calculated dose rate of 172 Gy h using Micro-Shield. The gamma-ray dose rate was 16.2 Gy h measured at 76 cm from the reactor core (using the "peanut" ion chamber) and 16.3 Gy h at 87 cm from the core (using Red Perspex film). The similarity of dose rates measured with different instruments indicates that using different methods and instruments is acceptable if the measurement (and calculation) parameters are well defined. Different measurement techniques may be necessary due to constraints such as size restrictions.

  5. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  6. Radiology trainer. Torso, internal organs and vessels. 2. ed.; Radiologie-Trainer. Koerperstamm, innere Organe und Gefaesse

    Energy Technology Data Exchange (ETDEWEB)

    Staebler, Axel [Orthopaedische Klinik Harlaching, Muenchen (Germany). Radiologische Praxis; Erlt-Wagner, Birgit (eds.) [Klinikum der Universitaet Muenchen (Germany). Inst. fuer Klinische Radiologie

    2013-11-01

    The radiology training textbook is based on case studies of the clinical experience, including radiological imaging and differential diagnostic discussion. The scope of this volume covers the torso, internal organs and vessels. The following issues are discussed: lungs, pleura, mediastinum; heart and vascular system; upper abdomen organs; gastrointestinal tract; urogenital system.

  7. Absorbed dose calculations to blood and blood vessels for internally deposited radionuclides

    International Nuclear Information System (INIS)

    Akabani, G.; Poston, J.W. Sr.

    1992-01-01

    At present, absorbed dose calculations for radionuclides in the human circulatory system use relatively simple models and are restricted in their applications. To determine absorbed doses to the blood and to the surface of the blood vessel wall, Monte Carlo calculations were performed using the code Electron Gamma Shower (EGS4). Absorbed doses were calculated for the blood and the blood vessel wall (lumen) for different blood vessel sizes. The radionuclides chosen for this study were those commonly used in nuclear medicine. No diffusion of the radionuclide into the blood vessel was or cross fire between blood vessels was assumed. Results are useful in assessing the doses to blood and blood vessel walls for different nuclear medicine procedures

  8. Absorbed dose calculations to blood and blood vessels for internally deposited radionuclides

    International Nuclear Information System (INIS)

    Akabani, G.; Poston, J.W.

    1991-05-01

    At present, absorbed dose calculations for radionuclides in the human circulatory system used relatively simple models and are restricted in their applications. To determine absorbed doses to the blood and to the surface of the blood vessel wall, EGS4 Monte Carlo calculations were performed. Absorbed doses were calculated for the blood and the blood vessel wall (lumen) for different blood vessels sizes. The radionuclides chosen for this study were those commonly used in nuclear medicine. No diffusion of the radionuclide into the blood vessel was assumed nor cross fire between vessel was assumed. Results are useful in assessing the dose in blood and blood vessel walls for different nuclear medicine procedures. 6 refs., 6 figs., 5 tabs

  9. Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water

    International Nuclear Information System (INIS)

    Maruyama, Yu; Yamano, Norihiro; Moriyama, Kiyofumi; Park, Hyun Sun; Kudo, Tamotsu; Yang, Yanhua; Sugimoto, Jun

    1999-01-01

    In-vessel debris coolability experiments were performed in ALPHA program at JAERI. Molten aluminum oxide (Al 2 O 3 ) was poured into a pool of water in a lower head experimental vessel. Post-test observation and measurement using an ultrasonic technique indicated the formation of the interfacial gap between the solidified Al 2 O 3 and the vessel wall. Thermal responses of the vessel wall implied that the interfacial gap acted initially as a thermal resistance and water subsequently penetrated into the interfacial gap. The maximum heat flux at the inner surface of the vessel facing to the solidified Al 2 O 3 was roughly evaluated to be ranged from 320 kW/m 2 to 600 kW/m 2 . A post-test analysis was conducted with CAMP code. The influence of the interfacial gap on thermal behavior of Al 2 O 3 and the vessel wall was examined. (authors)

  10. Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

    International Nuclear Information System (INIS)

    Love, E.F.; Pauley, K.A.; Reid, B.D.

    1995-09-01

    This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

  11. An investigation of major influences on the seismic response of APR1400 reactor vessel internals - 15145

    International Nuclear Information System (INIS)

    Byun, Y.J.; Kim, J.G.; Sung, K.K.; Lee, D.H.

    2015-01-01

    This paper deals with 3 topics concerning the APR1400 reactor vessel internals (RVI) seismic analysis: nonlinear problems, approaches to account for uncertainties of seismic model, and dynamic responses to various seismic excitations. First, the noticeable nonlinear characteristics of the RVI seismic model are discussed, and the modeling methods for properly simulating the nonlinear behaviors of RVI under seismic loads are presented. By applying these methods to the seismic model, the seismic analysis can correctly predict the dynamic response of RVI. Next, two approaches to account for the uncertainties of seismic model are evaluated: the time history broadening method, and the sensitivity analysis based on NUREG-0800, Section 4.2, Appendix A. From the evaluation results, it is confirmed that the time history broadening method employed in the seismic analysis of APR1400 RVI sufficiently accounts for the uncertainty of seismic model. Finally, the response characteristics of APR1400 RVI to various seismic excitations are investigated. The seismic excitations corresponding to various soil profiles, including the effects of cracked and un-cracked concrete stiffness on the reactor containment building structure, are used as forcing functions. From this study, the effects of various site conditions on the dynamic response of APR1400 RVI are identified. As a result, the enveloped seismic responses obtained from this study will contribute to the development of RVI seismic design that covers a wide range of potential site conditions. (authors)

  12. Decreased hyperintense vessels on FLAIR images after endovascular recanalization of symptomatic internal carotid artery occlusion

    International Nuclear Information System (INIS)

    Liu Wenhua; Yin Qin; Yao Lingling; Zhu Shuanggen; Xu Gelin; Zhang Renliang; Ke Kaifu; Liu Xinfeng

    2012-01-01

    Background and purpose: Hyperintense vessels (HV) on fluid-attenuated inversion recovery (FLAIR) images were assumed to be explained by slow antegrade or retrograde leptomeningeal collateral flow related to extracranial or intracranial artery steno-occlusion. The aim of this study was to investigate the effect of recanalization after endovascular therapy of symptomatic internal carotid artery (ICA) occlusion on the presence of HV. Methods: Eleven patients with symptomatic ICA occlusion were retrospectively enrolled. Changes in the HV on FLAIR images were examined in affected hemisphere of each patient after successful treatment with endovascular recanalization (angioplasty, n = 3; stent-assisted angioplasty, n = 8). The relationship between postoperative changes in the HV and Thrombolysis In Cerebral Ischemia (TICI) scale (I-III) was assessed. Results: After operation, HV of the 11 affected hemispheres were showed to be decreased (n = 3) or disappeared (n = 8) in treated side. The median interval between pre- and postoperative MRI examinations was 97.0 h (range, from 69. to 48.7 h). Of the 8 patients with disappeared HV, 7 achieved high TICI grade flow (III) and 1 had relatively low TICI grade flow (IIc) in treated side. However, all the 3 patients with decreased HV were found to be relatively low TICI grade flow (IIc). Conclusion: Our data indicate that endovascular recanalization of ICA occlusion was effective for decreasing HV. Postoperative decrease in HV can be considered as a marker for hemodynamic improvement.

  13. The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

    International Nuclear Information System (INIS)

    Hermansky, P.; Krajcovic, M.

    2012-01-01

    The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

  14. Design of a supercritical water-cooled reactor. Pressure vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Kai

    2008-08-15

    The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

  15. Assessment of nonbackfittable concepts for improving uranium utilization in LWRs

    International Nuclear Information System (INIS)

    Newman, D.F.; Goldsmith, S.; Fleischman, R.M.

    1980-01-01

    Recent efforts to improve uranium utilization in light water reactors (LWRs) have involved backfittable changes to fuel or operations. The Advanced Reactor Design Study sponsored by the US Department of Energy identified and evaluated nonbackfittable LWR concepts to provide a basis for selecting and demonstrating specific improvements that have good implementation potential. Because the application of nonbackfittable concepts necessitates modifications to contemporary reactor designs, it was apparent that the most qualified organizations to assess implementation potential would be LWR designers/vendors. Accordingly, Babcock and Wilcox, Combustion Engineering, and General Electric were the principal participants in selecting, assessing, and evaluating the nonbackfittable concepts included in this study. The results of the industrial assessments of nonbackfittable LWR concepts are shown

  16. Fuel -coolant interactions in LWRs and LMFBRs: relationships and distinctions

    Energy Technology Data Exchange (ETDEWEB)

    Duffey, R B; Lellouche, G S [Nuclear Safety and Analysis Department, Electric Power Research Institute, Palo Alto, CA (United States)

    1979-10-15

    The question of fuel-coolant interaction and of potential vapor explosion is raised here. lt is the contention of the authors that there is in fact no need to study this question vis a vis Light Water Reactors (LWR) except from an academic point of view since it does not impact on safety considerations. As for LMFBRs, the design basis whole core accidents for LWRs are derived from the fundamental concern of maintaining core geometry to provide for convective cooling. However, the important distinction is that the core is in its most reactive configuration, and core and fuel rearrangement is therefore not of such concern. The author's thesis is that even if the probability of steam explosion following core melt were two orders of magnitude greater than currently assumed (10{sup -2}) the total LWR risk would increase only by a factor of 2-6 for BWRs and less a factor of 10 for PWRs

  17. The Integrated Approach to the Accident Evaluation for Advanced LWRs

    International Nuclear Information System (INIS)

    Oriolo, F.; Paci, S.

    1998-01-01

    The present paper discusses some relevant phenomena occurring in advanced LWRs during postulated accident scenarios. In particular, the operation of ESF is the starting point for analysis of those phenomena that cause the mutual influence between PS and containment in these plants. As a consequence, it is highlighted as accident analyses which treat PS and containment phenomena completely separated may be not adequate when applied to innovate reactors. Exemplified thermal-hydraulic analysis are presented for AP600 and SBWR, using the FUMO integrated model, for highlight accident evolution taking into account these interactions. The architecture of this integrated code is presented highlighting the importance of an integrated approach to the safety analysis in innovative reactors. (author)

  18. Acoustic emission/flaw relationships for inservice monitoring of LWRs

    International Nuclear Information System (INIS)

    Hutton, P.H.; Kurtz, R.J.; Friesel, M.A.; Skorpik, J.R.; Dawson, J.F.

    1991-10-01

    The program concerning Acoustic Emission/Flaw Relationships for Inservice Monitoring of LWRs was initiated in FY76 with the objective of validating the application of acoustic emission (AE) to monitor nuclear reactor pressure-containing components during operation to detect cracking. The program has been supported by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Research and development has been performed by Pacific Northwest Laboratory, operated for the Department of Energy by Battelle Memorial Institute. The program has shown the feasibility of continuous, on-line AE monitoring to detect crack growth and produced validated methods for applying the technology. Included are relationships for estimating flaw severity from AE data and field applications at Watts Bar Unit 1 Reactor, Limerick Unit 1 Reactor, and the High Flux Isotope Reactor. This report discusses the program scope and organization, the three program phases and the results obtained, standard and code activities, and instrumentation and software developed under this program

  19. Propertis of solidified radioactive wastes from commercial LWRs

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.; Colombo, P.

    1978-01-01

    A study has been performed to characterize the properties of solidified radioactive wastes generated in the liquid radwaste treatment systems at LWRs. The properties which have been studied are those which are pertinent in defining the relative potential for the release of radionuclides to the environment as well as others relating to the evaluation of various solidification agents on an economic and feasibility basis. The use of standard testing procedures in measuring these properties allows an intercomparison of respective properties between various types of solidified waste forms. The leachability, mechanical properties, thermal stability, radiation stability, and thermal properties of hydraulic cement, ureaformaldehyde, bitumen, and addition type polymer waste forms have been measured. In addition, the chemical sensitivity, volumetric efficiency and radiation shielding characteristics of these waste forms have been studied. Emphasis in this paper is placed on the results of studies concerning chemical compatibility of solidification agents with specific waste streams, volumetric efficiency, free standing water, and leachability

  20. A review of multiple recycle of plutonium in LWRs

    International Nuclear Information System (INIS)

    Hesketh, K.

    1997-01-01

    This paper presents a summary of the current status of the nuclear physics of multiple recycle in Light Water Reactors (LWRs). The paper draws attention to the principal nuclear physics issues of relevance and highlights where present knowledge is inadequate. This paper reviews the existing literature critically and provides a clearer picture of the assumptions underlying the past work and will highlight areas where past work has been deficient. From this comes a concise analysis of the technicalities of multiple MOX recycle and the areas of uncertainty which remain. The paper concludes that at present the question remains one for which there is as yet no satisfactory answer, largely because the basic nuclear data and lattice codes essential to define the technical limits are not yet sufficiently developed and validated. 6 refs, 2 figs, 3 tabs

  1. Investigation of the design of a metal-lined fully wrapped composite vessel under high internal pressure

    Science.gov (United States)

    Kalaycıoğlu, Barış; Husnu Dirikolu, M.

    2010-09-01

    In this study, a Type III composite pressure vessel (ISO 11439:2000) loaded with high internal pressure is investigated in terms of the effect of the orientation of the element coordinate system while simulating the continuous variation of the fibre angle, the effect of symmetric and non-symmetric composite wall stacking sequences, and lastly, a stacking sequence evaluation for reducing the cylindrical section-end cap transition region stress concentration. The research was performed using an Ansys® model with 2.9 l volume, 6061 T6 aluminium liner/Kevlar® 49-Epoxy vessel material, and a service internal pressure loading of 22 MPa. The results show that symmetric stacking sequences give higher burst pressures by up to 15%. Stacking sequence evaluations provided a further 7% pressure-carrying capacity as well as reduced stress concentration in the transition region. Finally, the Type III vessel under consideration provides a 45% lighter construction as compared with an all metal (Type I) vessel.

  2. Design and development of in-vessel viewing periscope for ITER (International Thermonuclear Experimental Reactor)

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Ito, Akira; Shibanuma, Kiyoshi; Tada, Eisuke

    1999-02-01

    An in-vessel viewing system is essential not only to detect and locate damage of components exposed to plasma, but also to monitor and assist in-vessel maintenance operation. In ITER, the in-vessel viewing system must be capable of operating at high temperature (200degC), under intense gamma radiation (30 kGy/h) and high vacuum or 1 bar inert gas. A periscope-type in-vessel viewing system has been chosen as a reference of the ITER in-vessel viewing system due to its wide viewing capability and durability for sever environments. According to the ITER research and development program, a full-scale radiation hard periscope with a length of 15 m has been successfully developed by the Japan Home Team. The performance tests have been shown sufficient capability at high temperature up to 250degC and radiation resistance over 100 MGy. This report describes the design and R and D results of the ITER in-vessel viewing periscope based on the development of 15-m-length radiation hard periscope. (author)

  3. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  4. Structural analysis of vacuum vessel and blanket support system for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Kitamura, Kazunori; Koizumi, Kouichi; Takatsu, Hideyuki; Tada, Eisuke; Shimane, Hideo.

    1996-11-01

    Structural analyses of vacuum vessel and blanket support system have been performed to examine their integrated structural behavior under the design loads and to assess their structural feasibility, with two kinds of three-dimensional (3-D) FEM models; a detailed model with 18deg sector region to investigate the detailed mechanical behaviors of the blanket and vessel components under the several symmetric loads, and a 180deg torus model with relatively coarser meshes to assess the structural responses under the asymmetric VDE load. The analytical results obtained by both models were also compared for the several symmetric loads to check the equivalent mechanical stiffness of the 180deg torus model. As the results, most of the vessel and blanket components have sufficient mechanical integrities with the stress level below the allowable limit of the materials, while the lower parts of inboard/outboard back plate need to be reinforced by increasing the thickness and/or mounting a toroidal ring support at the lower edge of the back plate. Two types of eigenvalue analyses were also conducted with the 180deg torus model to investigate natural frequencies of the vessel torus support system and to assess the mechanical integrity of the elastic stability under the asymmetric VDE load. Analytical results show that the mechanical stiffness of the vessel gravity support should be higher in the view point of a seismic response, and that those of the blanket support structures should also be increased for the buckling strength against the VDE vertical force. (author)

  5. International pressure vessels and piping codes and standards. Volume 2: Current perspectives; PVP-Volume 313-2

    International Nuclear Information System (INIS)

    Rao, K.R.; Asada, Yasuhide; Adams, T.M.

    1995-01-01

    The topics in this volume include: (1) Recent or imminent changes to Section 3 design sections; (2) Select perspectives of ASME Codes -- Section 3; (3) Select perspectives of Boiler and Pressure Vessel Codes -- an international outlook; (4) Select perspectives of Boiler and Pressure Vessel Codes -- ASME Code Sections 3, 8 and 11; (5) Codes and Standards Perspectives for Analysis; (6) Selected design perspectives on flow-accelerated corrosion and pressure vessel design and qualification; (7) Select Codes and Standards perspectives for design and operability; (8) Codes and Standards perspectives for operability; (9) What's new in the ASME Boiler and Pressure Vessel Code?; (10) A look at ongoing activities of ASME Sections 2 and 3; (11) A look at current activities of ASME Section 11; (12) A look at current activities of ASME Codes and Standards; (13) Simplified design methodology and design allowable stresses -- 1 and 2; (14) Introduction to Power Boilers, Section 1 of the ASME Code -- Part 1 and 2. Separate abstracts were prepared for most of the individual papers

  6. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    International Nuclear Information System (INIS)

    Lindley, B.A.; Lillington, J.N.; Kotlyar, D.; Parks, G.T.; Petrovic, B.

    2016-01-01

    The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO_2/PuO_2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R and D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U_3Si_2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R and D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I"2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I"2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I"2S-LWR design are U_3Si_2 and advanced stainless steel respectively. In addition, the I"2S-LWR design

  7. ITER vacuum vessel fabrication plan and cost study (D 68) for the international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    1995-01-01

    ITER Task No. 8, Vacuum Vessel Fabrication Plan and Cost Study (D68), was initiated to assess ITER vacuum vessel fabrication, assembly, and cost. The industrial team of Raytheon Engineers ampersand Constructors and Chicago Bridge ampersand Iron (Raytheon/CB ampersand I) reviewed the current vessel basis and prepared a manufacturing plan, assembly plan, and cost estimate commensurate with the present design. The guidance for the Raytheon/CB ampersand I assessment activities was prepared by the ITER Garching Work Site. This guidance provided in the form of work descriptions, sketches, drawings, and costing guidelines for each of the presently identified vacuum vessel Work Breakdown Structure (WBS) elements was compiled in ITER Garching Joint Work Site Memo (Draft No. 9 - G 15 MD 01 94-17-05 W 1). A copy of this document is provided as Appendix 1 to this report. Additional information and clarifications required for the Raytheon/CB ampersand I assessments were coordinated through the US Home Team (USHT) and its technical representative. Design details considered essential to the Task 8 assessments but not available from the ITER Joint Central Team (JCT) were generated by Raytheon/CB ampersand I and documented accordingly

  8. 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.)

  9. Contribution of materials investigations and operating experience of reactor vessel internals to PWRs' safety, performance and reliability

    International Nuclear Information System (INIS)

    Lemaire, E.; Monteil, N.; Jardin, N.; Doll, M.

    2015-01-01

    The Reactor Pressure Vessel Internals (RVI) include all the components inside the pressure vessel, except the nuclear fuel, the rod cluster assemblies and the instrumentation. The RVI consist of bolted and welded structures that are divided into two sub-assemblies: the upper internals which are removed at every refueling outage and the lower internals which are systematically removed for inspection at every 10-year outage. The main functions of the RVI are to position the core, to support it, and to provide a coolant flow by channeling the fluid. Moreover, the lower internals contribute to a neutron protection of the reactor pressure vessel by absorbing most of the neutron flux from the core. Depending on their location and material composition, the RVI components can face different ageing phenomena, that are actual or potential (such as wear, fatigue, stress corrosion cracking, irradiation assisted stress corrosion cracking, hardening and loss of ductility due to neutron irradiation, irradiation creep and irradiation swelling). EDF has developed a strategy for managing ageing and demonstrating the capacity of the RVI to perform their design functions over 40 years of operation. This overall approach is periodically revisited to take into account the most recent knowledge obtained from the following main topics: Safety Analyses, Research-Development programs, In-Service Inspection (ISI) results, Maintenance programs and Metallurgical Examinations. Based on continuous improvements in those fields, the goal of this paper is to present the way that materials investigations and operating experience obtained on RVI are managed by EDF to improve RVI safety, performance and reliability. It is shown that a perspective of 60 years of operation of RVI components is supported by large Research-Development efforts combined with field experience. (authors)

  10. 46 CFR 32.50-35 - Remote manual shutdown for internal combustion engine driven cargo pump on tank vessels-TB/ALL.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 1 2010-10-01 2010-10-01 false Remote manual shutdown for internal combustion engine... for Cargo Handling § 32.50-35 Remote manual shutdown for internal combustion engine driven cargo pump on tank vessels—TB/ALL. (a) Any tank vessel which is equipped with an internal combustion engine...

  11. Experimental tests on buckling of ellipsoidal vessel heads subjected to internal pressure

    International Nuclear Information System (INIS)

    Roche, R.L.; Alix, M.

    1980-05-01

    Tests were performed on 17 ellipsoidal vessel heads of three different materials and different geometries. The results include the following: 1) Accurate definition of the geometry and particularly a direct measurement of the thickness along the meridian. 2) The properties of the material of each head, obtained from test specimens cut from the head itself after the test. 3) The recording of deflection/pressure curves with indication of the pressure at which buckling occurred. These results can be used for validation and qualification of methods for calculating the buckling load when plasticity occurs before buckling. It was possible to develop an empirical equation representing the experimental results obtained with satisfactory accuracy. This equation may be useful in pressure vessel design

  12. Remote handling and robotic inspections of Palo Verde reactor vessel internals

    International Nuclear Information System (INIS)

    Ryder, W.

    1998-01-01

    Remote visual examinations and handling evolutions in high radiation field environments have required the use of radiation tolerant video systems. These systems involve significant expense and potentially require large envelope deployment structures. Recent events at Palo Verde including Upper Guide Structure damage and Reactor Vessel In-Service Inspections have provided opportunities for research, design and utilization of alternative approaches. Most significant of these, utilization of CCD modules with high magnification capabilities, have produced higher quality viewing, reduced maintenance expenditures, and rapid deployment intervals. (orig.) [de

  13. International workshop on WWER-440 reactor pressure vessel embrittlement and annealing. Working material

    International Nuclear Information System (INIS)

    1994-01-01

    The purpose of the Workshop was essentially to discuss the WWER 440 model 230 reactor pressure vessel integrity in terms of the measures already taken, current activities and future plans. The meeting was arranged in two parts, namely, the Scientific programme followed by the consideration, review and revision of the IAEA Consultancy report on RPV Embrittlement and Annealing. This particular report covers the first part of the meeting i.e., the Scientific Programme, in the form of proceedings of the meeting, while the re-drafted Consultancy report will be issued later. The meeting was attended by sixty-six representatives from thirteen countries. Refs, figs and tabs

  14. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  15. Structural criteria for extreme dynamic internal pressure loadings of vessels and closure heads

    International Nuclear Information System (INIS)

    Bitner, J.L.

    1985-01-01

    The criteria protect against tensile plastic instability and local ductile rupture failure modes. To minimize the number of critical areas that may need more rigorous analytical methods, a screening criterion for limiting the membrane, bending and local stresses is defined. The stresses for this criterion are calculated from either simple and economical elastic dynamic or equivalent static methods. For the critical areas that remain, a strain-based criterion for strains derived from dynamic, inelastic methods is given. To assure that the criteria are properly applied, guidelines are outlined for controlling methods for deriving stresses and strains, for selecting appropriate material properties and for addressing specific dominating parameters that affect the validity of the analysis. The application of the criteria to a complex liquid metal fast breeder reactor vessel and closure head and the subsequent experimental verification of the results by several scale model experiments are summarized. (orig./HP)

  16. COMPARATIVE STUDY THROUGH FINITE ELEMENT METHOD OF LIDS USED IN CYLINDRICAL VESSEL IN HORIZONTAL POSITION SUBJECT TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Eusebio V. Ibarra-Hernández

    2017-07-01

    Full Text Available In this work a study of the cylindrical vessels in horizontal position and subject to internal pressure is carried out, where lids are one of the main components of this equipment. The Autodesk Inventor pro. 2016 is used to make the geometrical characterization of these elements: parametric solid modeler, assembles and surfaces for the mechanical design of complex parts. The different geometric forms of the lids and bottoms analyzed in this work are: flat-circular with or without flange, elliptical with different values of the K factor, torispherical with different values of the M factor and the hemispherical bottoms. Using the Finate Element Method (FEM, a comparative study is made about the behavior of the stress and strain in the different geometrical forms mentioned before, being demonstrated that although the best resistance and rigidity values are presented by the hemispherical bottoms and the best options of production by the flat-circulars, they are not the bottoms used the most in this vessels, being the elliptic bottoms those of more use. The results obtained allow optimizing the design and knowing the thickness limit in the most requested areas.

  17. Maury Journals - German Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — German vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  18. Loads on reactor pressure vessel internals induced by low-pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-02-01

    Departing from the conservation theorems for mass and impulse the computer code DRUWE has been developed which allows to calculate loads on the core shell with simplifying assumptions for the first period just after the rupture has opened. It can be supposed that the whole rupture cross section is set free within 15 msec. The calculation progresses in a way that for a core shell the local, timely pressure- and load development, respectively, the total dynamic load as well as the moments acting on the fixing of the core shell, can be calculated. The required input data are merely geometric data on the concept of the pressure vessel and its components as well as the effective subcooling of the fluid. By means of some parameters the programm development can be controlled in a way that the results are available in form of listings or diagrams, respectively, as well as in form of card decks for following investigations, e.g. solidity calculations. (orig./RW) [de

  19. Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

    1992-09-01

    Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

  20. Pressure vessel failure at high internal pressure; Untersuchungen zum Versagen des Reaktordruckbehaelters unter hohem Innendruck

    Energy Technology Data Exchange (ETDEWEB)

    Laemmer, H.; Ritter, B.

    1995-08-01

    A RPV failure due to plastic instability was investigated using the ABAQUS finite element code together with a material model of thermal plasticity for large deformations. Not only rotational symmetric temperature distributions were studied, but also `hot spots`. Calculations show that merely by the depletion of strength of the material - even at internal wall temperatures well below the melting point of the fuel elements of about 2000/2400 C - the critical internal pressure can decrease to values smaller than the operational pressure of 16 Mpa. (orig.)

  1. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals: 2007 update

    International Nuclear Information System (INIS)

    2007-06-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that effective control of the ageing degradation of the major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance or manufacturing errors) is one of the most important issues for plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wearout of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. IAEA-TECDOC-1119 documents ageing assessment and management practices for PWR Reactor Vessel Internals (RVIs) that were current at the time of its finalization in 1997-1998. Safety significant operating events have occurred since the finalization of the TECDOC, e.g. irradiation assisted stress corrosion cracking (IASCC) of baffle-former bolts, which threatened the integrity of the vessel internals. In addition, concern of fretting wear of control rod guide tubes has been raised in Japan. These events led to new ageing management actions by both NPP operators and regulators. Therefore it was recognized that IAEA-TECDOC-1119 should be updated by incorporating those new events and their countermeasures. The objective of this report is to update relevant sections of the existing IAEA-TECDOC- 1119 in order to provide current ageing management guidance for PWR RVIs to all involved in the operation and regulation of PWRs and thus to help ensure PWR safety in IAEA Member States throughout their entire service life

  2. FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE

    Directory of Open Access Journals (Sweden)

    Entin Hartini

    2016-06-01

    Full Text Available ABSTRACT FRACTURE MECHANICS UNCERTAINTY ANALYSIS IN THE RELIABILITY ASSESSMENT OF THE REACTOR PRESSURE VESSEL: (2D SUBJECTED TO INTERNAL PRESSURE. The reactor pressure vessel (RPV is a pressure boundary in the PWR type reactor which serves to confine radioactive material during chain reaction process. The integrity of the RPV must be guaranteed either  in a normal operation or accident conditions. In analyzing the integrity of RPV, especially related to the crack behavior which can introduce break to the reactor pressure vessel, a fracture mechanic approach should be taken for this assessment. The uncertainty of input used in the assessment, such as mechanical properties and physical environment, becomes a reason that the assessment is not sufficient if it is perfomed only by deterministic approach. Therefore, the uncertainty approach should be applied. The aim of this study is to analize the uncertainty of fracture mechanics calculations in evaluating the reliability of PWR`s reactor pressure vessel. Random character of input quantity was generated using probabilistic principles and theories. Fracture mechanics analysis is solved by Finite Element Method (FEM with  MSC MARC software, while uncertainty input analysis is done based on probability density function with Latin Hypercube Sampling (LHS using python script. The output of MSC MARC is a J-integral value, which is converted into stress intensity factor for evaluating the reliability of RPV’s 2D. From the result of the calculation, it can be concluded that the SIF from  probabilistic method, reached the limit value of  fracture toughness earlier than SIF from  deterministic method.  The SIF generated by the probabilistic method is 105.240 MPa m0.5. Meanwhile, the SIF generated by deterministic method is 100.876 MPa m0.5. Keywords: Uncertainty analysis, fracture mechanics, LHS, FEM, reactor pressure vessels   ABSTRAK ANALISIS KETIDAKPASTIAN FRACTURE MECHANIC PADA EVALUASI KEANDALAN

  3. Subaquatic, pressure vessels and LPG storage spheres internal inspection; Inspecao interna de esfera utilizando mergulho como acesso

    Energy Technology Data Exchange (ETDEWEB)

    Filgueira Filho, Rafael; Monteiro, Ayres [PETROBRAS, Rio de Janeiro, RJ (Brazil)

    2005-07-01

    Minimizing shut-down costs is a widespread target in the oil and gas industry. The use of new inspection techniques is one of the ways for that. This work presents a new procedure for internal inspections in pressure vessels by the non destructive testing - NDT, ACFM, using industrial diving techniques. As a pioneer experience, this method was applied in the inspection of the internal parts of the LPG sphere tank 5101 at PETROBRAS Transporte S.A. - TRANSPETRO, in Jequie's Terminal, in the state of Bahia, in december, 2003. This new method allows the reduction of indirect costs related to operational unavailability of the equipment, by the reduction of the shut-down time in approximately 50%, when compared to the demanded shut down time, when using scaffolds for accessing the internal parts. Despite of direct costs are still higher with the new methodology, this paper demonstrates the economical feasibility of this new method, based on the savings obtained with the fastest return of the equipment to operation. (author)

  4. Computer-generated vibratory signatures for EDF PWR reactor vessel internals

    International Nuclear Information System (INIS)

    Trenty, A.; Lefevre, F.; Garreau, D.

    1992-07-01

    This paper presents a device for generation of characteristic signatures for normal or faulty vibrations on EDF PWR internal structures. The objective is to test the efficiency of methods for diagnosing faults in these structures. With this device, it is possible to build an entire PSD in several phases: choice of a general basic shape, localized addition of several kinds of background noise, generation of peaks of variable shapes, adjustment of local or global amplifications... It also offers the possibility of distorting real PSDs acquired from the reactor: shifting frequency or modifying peak shape, eliminating or adding existing shapes or shapes to be created, smoothing curves... One example is given of simulated loss of function in a hold-down spring on a computer-generated PSD of ex-core neutron noise. The device is now being used to test the potential of neural networks in recognizing faults on internal structures

  5. Characterization of Decommissioned PWR Vessel Internals Material Samples: Tensile and SSRT Testing (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.

    2004-01-01

    Pressurized water reactor (PWR) cores operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs requires detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel (internals) subjected to such conditions. This project studied the effects of reactor service on the mechanical and corrosion properties of samples of baffle plate, former plate, and core barrel from a decommissioned PWR

  6. Characterization of Decommissioned PWR Vessel Internals Materials Samples: Material Certification, Fluence, and Temperature (Nonproprietary Version)

    International Nuclear Information System (INIS)

    Krug, M.; Shogan, R.; Fero, A.; Snyder, M.

    2004-01-01

    Pressurized water reactor (PWR) cores, operate under extreme environmental conditions due to coolant chemistry, operating temperature, and neutron exposure. Extending the life of PWRs require detailed knowledge of the changes in mechanical and corrosion properties of the structural austenitic stainless steel components adjacent to the fuel. This report contains basic material characterization information of the as-installed samples of reactor internals material which were harvested from a decommissioned PWR

  7. Petrous internal carotid aneurysm causing epistaxis: Balloon embolization with preservation of the parent vessel

    Energy Technology Data Exchange (ETDEWEB)

    Willinsky, R.; Lasjaunias, P.; Pruvost, P.; Boucherat, M.

    1987-11-01

    A patient with severe, recurrent posterior epistaxis was shown at angiography to have an aneurysm of the petrous portion of the internal carotid artery (ICA). Since childhood, she had had pain related to eustachian tube blockage by the aneurysm. An endovascular balloon embolization of the aneurysm was successful with preservation of the parent artery. The treatment resulted in resolution of the symptoms. The report confirms the usefulness of an angiographic protocol in evaluating vascular problems.

  8. Petrous internal carotid aneurysm causing epistaxis: Balloon embolization with preservation of the parent vessel

    International Nuclear Information System (INIS)

    Willinsky, R.; Lasjaunias, P.; Pruvost, P.

    1987-01-01

    A patient with severe, recurrent posterior epistaxis was shown at angiography to have an aneurysm of the petrous portion of the internal carotid artery (ICA). Since childhood, she had had pain related to eustachian tube blockage by the aneurysm. An endovascular balloon embolization of the aneurysm was successful with preservation of the parent artery. The treatment resulted in resolution of the symptoms. The report confirms the usefulness of an angiographic protocol in evaluating vascular problems. (orig.)

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

    International Nuclear Information System (INIS)

    Carvalho, E.A. de

    2005-01-01

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

  10. TEMP-STRESS analysis of a reinforced concrete vessel under internal pressure

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Kennedy, J.M.; Pfeiffer, P.A.

    1987-01-01

    Prediction of the response of the Sandia National laboratory 1/6-scale reinforced concrete containment model test was obtained by Argonne National Laboratory (ANL) employing a computer program developed by ANL. The test model was internally pressurized to failure. The two-dimensional code TEMP-STRESS [1-5] has been developed at ANL for stress analysis of plane and axisymmetric 2-D reinforced structures under various thermal conditions. The program is applicable to a wide variety of nonlinear problems, and is utilized in the present study. The comparison of these pretest computations with test data on the containment model should be a good indication of the state of the code

  11. Analysis of the fluid-structure dynamic interaction of reactor pressure vessel internals during blowdown

    International Nuclear Information System (INIS)

    Schlechtendahl, E.G.; Krieg, R.; Schumann, U.

    1977-01-01

    The loadings on reactor internal structures (in particular the core barrel) induced during a PWR-blowdown must not result in excessive stresses and strains. The deformations are strongly influenced by the coupling of fluid and structure dynamics and it is necessary, therefore, to develop and apply new coupled analysis tools. In this paper a survey is given over work currently in progress in the Nuclear Research Center Karlsruhe and the Los Alamos Scientific Laboratory which aim towards 'best estimate codes'. The new methods will be verified by means of the HDR-blowdown tests and other experiments. The results of several scoping calculations are presented and illustrated by movie films. (orig.) [de

  12. Modernization of Cross Section Library for VVER-1000 Type Reactors Internals and Pressure Vessel Dosimetry

    Directory of Open Access Journals (Sweden)

    Voloschenko Andrey

    2016-01-01

    Full Text Available The broad-group library BGL1000_B7 for neutron and gamma transport calculations in VVER-1000 internals, RPV and shielding was carried out on a base of fine-group library v7-200n47g from SCALE-6 system. The comparison of the library BGL1000_B7 with the library v7-200n47g and the library BGL1000 (the latter is using for VVER-1000 calculations is demonstrated on several calculation and experimental tests.

  13. Inspection, evaluation and maintenance guidelines for reactor vessel internals in JAPAN

    International Nuclear Information System (INIS)

    Sakashita, Akihiro; Goto, Tomoya; Hirano, Shinro; Dozaki, Koji

    2010-01-01

    Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows. Repair: methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized. Replacement: methodology to replace the existing structure with new one when the degradation of structure is actualized. Preventive maintenance : methodology to mitigate the damaged condition. When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management. (author)

  14. Vessel Operating Units (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data for vessels that are greater than five net tons and have a current US Coast Guard documentation number. Beginning in1979, the NMFS...

  15. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessel internals

    International Nuclear Information System (INIS)

    2005-10-01

    . The guidance reports are directed at technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The report addresses the reactor pressure vessel internals in BWRs. Maintaining the structural integrity of these reactor pressure vessel internals throughout NPP service life, in spite of several ageing mechanisms, is essential for plant safety

  16. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    OpenAIRE

    KO, DO-YOUNG; KIM, KYU-HYUNG

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

  17. Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

    International Nuclear Information System (INIS)

    Wang, J.A.; Kam, F.B.K.

    1998-02-01

    The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs

  18. MRP-227 Reactor vessel internals inspection planning and initial results at the Oconee nuclear station unit 2

    International Nuclear Information System (INIS)

    Davidsaver, S.B.; Fyfitch, S.; Whitaker, D.E.; Doss, R.L.

    2015-01-01

    The U.S. PWR industry has pro-actively developed generic inspection requirements and standards for reactor vessel (RV) internals. The Electric Power Research Institute (EPRI) Pressurized Water Reactor (PWR) Materials Reliability Program (MRP) has issued MRP-227-A and MRP-228 with mandatory and needed requirements based on the Nuclear Energy Institute (NEI) document NEI 03-08. The inspection and evaluation guidelines contained in MRP-227-A consider eight age-related degradation mechanisms: stress corrosion cracking (SCC), irradiation-assisted stress corrosion cracking (IASCC), wear, fatigue, thermal aging embrittlement, irradiation embrittlement, void swelling and irradiation growth, and thermal and irradiation-enhanced stress relaxation or irradiation-enhanced creep. This paper will discuss the decision planning efforts required for implementing the MRP-227-A and MRP-228 requirements and the results of these initial inspections at the Oconee Nuclear power station (ONS) units. Duke Energy and AREVA overcame a significant technology and NDE challenge by successfully completing the first-of-a-kind MRP-227-A scope requirements at ONS-1 in one outage below the estimated dose and with zero safety issues or events. This performance was repeated at ONS-2 a year later. The remote NDE tooling and processes developed to examine the MRP-227-A scope for ONS-1 and ONS-2 are transferable to other PWRs

  19. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    International Nuclear Information System (INIS)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

    2016-01-01

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

  20. Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

  1. Plutonium recycle in LWRs - The programme of the Federal Republic of Germany

    International Nuclear Information System (INIS)

    1979-01-01

    This paper explains that in accordance with the Federal Republic of Germany's nuclear energy policy, spent fuel from their large programme of LWRs will be stored for some years in an interim storage pool until capacity for reprocessing, which is essential on environmental grounds, becomes available after 1990. The residual fissile content (uranium and plutonium) of the spent fuel can then be recovered and either be recycled into thermal LWRs or later on be burnt in Fast Reactors. A comprehensive R, D and D programme for plutonium recycle was launched in 1971. Up to the end of 1978 about 10,000 MOX fuel rods were designed, specified, fabricated and irradiated in several nuclear power plants. The methods used in this work and the results are described

  2. Computed tomography angiographic study of internal mammary perforators and their use as recipient vessels for free tissue transfer in breast reconstruction

    Directory of Open Access Journals (Sweden)

    Aditya V Kanoi

    2017-01-01

    Full Text Available Context: The internal mammary artery perforator vessels (IMPV as a recipient in free flap breast reconstruction offer advantages over the more commonly used thoracodorsal vessels and the internal mammary vessels (IMV. Aims: This study was designed to assess the anatomical consistency of the IMPV and the suitability of these vessels for use as recipients in free flap breast reconstruction. Patients and Methods: Data from ten randomly selected female patients who did not have any chest wall or breast pathology but had undergone a computed tomography angiography (CTA for unrelated diagnostic reasons from April 2013 to October 2013 were analysed. Retrospective data of seven patients who had undergone mastectomy for breast cancer and had been primarily reconstructed with a deep inferior epigastric artery perforator free flap transfer using the IMPV as recipient vessels were studied. Results: The CTA findings showed that the internal mammary perforator was consistently present in all cases bilaterally. In all cases, the dominant perforator arose from the upper four intercostal spaces (ICS with the majority (55% arising from the 2nd ICS. The mean distance of the perforators from the sternal border at the level of pectoralis muscle surface on the right side was 1.86 cm (range: 0.9–2.5 cm with a mode value of 1.9 cm. On the left side, a mean of 1.77 cm (range: 1.5–2.1 cm and a mode value of 1.7 cm were observed. Mean perforator artery diameters on the right and left sides were 2.2 mm and 2.4 mm, respectively. Conclusions: Though the internal mammary perforators are anatomically consistent, their use as recipients in free tissue transfer for breast reconstruction eventually rests on multiple variables.

  3. Assessment of NDE Technologies for Detection and Characterization of Stress Corrosion Cracking in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Ramuhalli, Pradeep; Toloczko, Mychailo B.; Bond, Leonard J.; Montgomery, Robert O.

    2012-12-31

    Stress corrosion cracking (SCC) in light water reactors (LWRs) has been a persistent form of degradation in the nuclear industry. Examples of SCC can be found for a range of materials in boiling and pressurized water reactor environments, including carbon steels, stainless steels, and nickel-base stainless alloys. The evolution of SCC is often characterized by a long initiation stage followed by a phase of more rapid crack growth to failure. This provides a relatively short window of opportunity to detect the start of observable SCC, and it is conceivable that SCC could progress from initiation to failure between subsequent examinations when managed by applying periodic in-service inspection techniques. Implementation of advanced aging management paradigms in the current fleet of LWRs will require adaptation of existing measurement technologies and development of new technologies to perform on-line measurements during reactor operation to ensure timely detection of material degradation and to support the implementation of advanced diagnostics and prognostics. This paper considers several non-destructive examination (NDE) technologies with known sensitivity to detection of indicators for SCC initiation and/or propagation, and assesses these technologies with respect to their ability to detect and accurately characterize the significance of an SCC flaw. Potential strategies to improve SCC inspection or monitoring performance are offered to benefit management of SCC degradation in LWRs.

  4. Assessment of NDE technologies for detection and characterization of stress corrosion cracking in LWRs

    International Nuclear Information System (INIS)

    Meyer, R.M.; Ramuhalli, P.; Toloczko, M.B.; Bond, L.J.; Montgomery, R.O.

    2012-01-01

    Stress corrosion cracking (SCC) in light water reactors (LWRs) has been a persistent form of degradation in the nuclear industry. Examples of SCC can be found for a range of materials in boiling and pressurized water reactor environments, including carbon steels, stainless steels, and nickel-base alloys. The evolution of SCC is often characterized by a long initiation stage followed by a phase of more rapid crack growth to failure. This provides a relatively short window of opportunity to detect the start of observable SCC, and it is conceivable that SCC could progress from initiation to failure between subsequent examinations when managed by applying periodic in-service inspection techniques. Implementation of advanced aging management paradigms in the current fleet of LWRs will require adaptation of existing measurement technologies and development of new technologies to perform on-line measurements during reactor operation to ensure timely detection of material degradation and to support the implementation of advanced diagnostics and prognostics. This paper considers several non-destructive examination (NDE) technologies with known sensitivity to detection of indicators for SCC initiation and/or propagation, and assesses these technologies with respect to their ability to detect and accurately characterize the significance of an SCC flaw. Potential strategies to improve SCC inspection or monitoring performance are offered to benefit management of SCC degradation in LWRs. (author)

  5. Meeting the physics design challenges of modern LWRs being inducted into the Indian nuclear power programme

    International Nuclear Information System (INIS)

    Jagannathan, V.; Pal, Usha; Karthikeyan, R.; Raj, Devesh; Srivastava, Argala; Khan, Suhail Ahmad

    2007-01-01

    Indian nuclear power programme started with the two Boiling Water Reactors (BWR) of 210 MWe capacity at Tarapur. Two VVER-1000 MWe reactors which are Pressurized Water Reactors (PWR) of Russian design are being constructed at Kudankulam, Tamilnadu and are expected to be commissioned by end 2008. There may be also a possibility of inducting some western PWRs in future. These reactors belong to the category of light water reactors (LWR). The LWRs are compact and have complex physical characteristics distinctly different from those of the Pressurized Heavy Water Reactors (PHWR) which, currently form the mainstay of our indigenous nuclear power programme. The physics design and analysis capability for the modern LWRs (BWR, PWR and VVER) has been developed at Light Water Reactors Physics Section, BARC. This paper presents the current state of art in this key technology area to meet the physics design and operation challenges when LWRs would be inducted in a major way into the Indian nuclear power programme and commence operating in the coming decades. (author)

  6. Detection and sizing of inside-surface cracks in reactor pressure vessels

    International Nuclear Information System (INIS)

    Kamata, Hiroshi; Kanazawa, Katsuo; Satoh, Kunio; Honma, Takashi

    1984-01-01

    According to the past operational experience of LWRs, most of the defects arising in reactor pressure vessels accompanying operation are cracks occurring in the build up welding of austenitic stainless steel on the internal surfaces. The detection of these cracks has been carried out by ultrasonic flaw detection from outside in BWRs and from inside in PWRs as in-service inspection. However, there are difficulties such as ultrasonic echoes often occur though defects do not exist, and the quantitative evaluation of detected cracks is difficult by this method because of its accuracy. One of the means to reduce the first difficulty is to use eddy current method together to help the judgement, and for overcoming the second, the ultrasonic method catching end peak echo, that catching diffracted waves, eddy current method and electric resistance method were tried and compared. It is desirable to detect cracks in early stage before they reach parent material. The techniques to detect cracks on the internal surfaces of pressure vessels from the inside and to measure the depth are reported in this paper. The methods of flaw detection examined and the instruments used, the experimental method and the results are reported. It was concluded that eddy current method can be used as the backup for ultrasonic remote flaw detection, and the accuracy of depth measurement was the highest in ultrasonic diffraction wave method. (Kako, I.)

  7. FFTF and CRBRP reactor vessels

    International Nuclear Information System (INIS)

    Morgan, R.E.

    1977-01-01

    The Fast Flux Test Facility (FFTF) reactor vessel and the Clinch River Breeder Reactor Plant (CRBRP) reactor vessel each serve to enclose a fast spectrum reactor core, contain the sodium coolant, and provide support and positioning for the closure head and internal structure. Each vessel is located in its reactor cavity and is protected by a guard vessel which would ensure continued decay heat removal capability should a major system leak develop. Although the two plants have significantly different thermal power ratings, 400 megawatts for FFTF and 975 megawatts for CRBRP, the two reactor vessels are comparable in size, the CRBRP vessel being approximately 28% longer than the FFTF vessel. The FFTF vessel diameter was controlled by the space required for the three individual In-Vessel Handling Machines and Instrument Trees. Utilization of the triple rotating plug scheme for CRBRP refueling enables packaging of the larger CRBRP core in a vessel the same diameter as the FFTF vessel

  8. Assessment and management of ageing of major nuclear power plant components important to safety: PWR vessel internals

    International Nuclear Information System (INIS)

    1999-10-01

    components addressed in the reports. This report addresses the pressurized water reactor vessel internals (taken as a single component). The IAEA acknowledges the work of all contributors to drafting and review of this report

  9. Cerebral White Matter Hypoperfusion Increases with Small-Vessel Disease Burden. Data From the Third International Stroke Trial.

    Science.gov (United States)

    Arba, Francesco; Mair, Grant; Carpenter, Trevor; Sakka, Eleni; Sandercock, Peter A G; Lindley, Richard I; Inzitari, Domenico; Wardlaw, Joanna M

    2017-07-01

    Leukoaraiosis is associated with impaired cerebral perfusion, but the effect of individual and combined small-vessel disease (SVD) features on white matter perfusion is unclear. We studied patients recruited with perfusion imaging in the Third International Stroke Trial. We rated individual SVD features (leukoaraiosis, lacunes) and brain atrophy on baseline plain computed tomography or magnetic resonance imaging. Separately, we assessed white matter at the level of the lateral ventricles in the cerebral hemisphere contralateral to the stroke for visible areas of hypoperfusion (present or absent) on 4 time-based perfusion imaging parameters. We examined associations between SVD features (individually and summed) and presence of hypoperfusion using logistic regression adjusted for age, sex, baseline National Institutes of Health Stroke Scale, hypertension, and diabetes. A total of 115 patients with median (interquartile range) age of 81 (72-86) years, 78 (52%) of which were male, had complete perfusion data. Hypoperfusion was most frequent on mean transit time (MTT; 63 patients, 55%) and least frequent on time to maximum flow (19 patients, 17%). The SVD score showed stronger independent associations with hypoperfusion (e.g., MTT, odds ratio [OR] = 2.80; 95% confidence interval [CI] = 1.56-5.03) than individual SVD markers (e.g., white matter hypoattenuation score, MTT, OR = 1.49, 95% CI = 1.09-2.04). Baseline blood pressure did not differ by presence or absence of hypoperfusion or across strata of SVD score. Presence of white matter hypoperfusion increased with SVD summed score. The SVD summed score was associated with hypoperfusion more consistently than individual SVD features, providing validity to the SVD score concept. Increasing SVD burden indicates worse perfusion in the white matter. Copyright © 2017 National Stroke Association. Published by Elsevier Inc. All rights reserved.

  10. Evaluation Methodology for Void Swelling Susceptibility of APR1400 Reactor Vessel Internals for U.S. NRC Design Certification

    Energy Technology Data Exchange (ETDEWEB)

    Kweon, Hyeong Do; Lee, Do Hwan [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The APR1400 RVI (Reactor Vessel Internals) operates in harsh conditions, such as long term exposure to neutron irradiation, high temperatures, reactor coolant environment, and other operating loads. Therefore, even though the RVI components are mainly made of austenitic stainless steel which is well known to have good mechanical and corrosion-resistive properties, these operating conditions. The aging is characterized by a chromium depletion along grain boundaries of austenitic stainless steel, a decrease in ductility and fracture toughness of the steel, an increase in yield and ultimate strength of the steel, and a potential volume change due to void formation in the steel. For these reasons, under certain conditions of stress, temperature, and level of irradiation, the void swelling which is one of the challenging degradation mechanisms affecting the integrity of the RVI may appear at specific locations of the RVI, especially due to high neutron fluence and high temperature under localized gamma heating and low velocity of coolant flow. To assess the effects of operating neutron fluences, temperatures and stresses on the material properties changes and the susceptibility to the void swelling, the evaluation methodology of the APR1400 RVI components for U.S. NRC Design Certification was suggested in this paper. The approach to the evaluation is summarized as follows: 1. RVI component list of the APR1400 is collected. 2. Initial screening to determine the evaluation scope is completed using the design values of fluences. 3. Functionality assessments (radiation transport analysis, CFD analysis, structural analysis) are sequentially performed. 4. Susceptibility to the void swelling is identified through ANSYS/USERMAT module. KHNP believes that the proposed methodology which is based on the EPRI works for operating reactors is the best way to evaluate the void swelling for new reactors such as the APR1400.

  11. Evolution of thermal fatigue management of piping in US LWRs

    International Nuclear Information System (INIS)

    McDewitt, M.; Wolfe, K.; McGill, R.

    2015-01-01

    Fatigue usage caused by cyclic changes of thermally stratified reactor coolant in Light Water Reactor (LWR) pressure boundary piping was not an original consideration in US Nuclear Power Plant (NPP) designs. During the mid 1980's, several events involving cracking and leakage due to thermal cycling occurred in reactor coolant system branch piping at both US and International NPPs. In 1988, the US Nuclear Regulatory Commission (US NRC) issued Bulletin 88-08 to alert LWR licensees of the potential for piping failures due to stratified thermal cycling. In response to these events, the US nuclear industry developed initiatives to identify susceptible components and established measures to monitor and prevent future failures. These initiatives have been effective in preventing leakage events, but have also identified fewer defects than expected based on screening model predictions. Improved analytical techniques are being investigated to maintain program effectiveness while minimizing unnecessary non-destructive examinations. This paper discusses the evolution of the US thermal fatigue initiatives, and analytical concepts being evaluated to improve program efficiency. (authors)

  12. Proceedings of the 3rd international conference on heat exchangers, boilers and pressure vessels (HEB-97). Vol.1 (Research Papers)

    International Nuclear Information System (INIS)

    1997-04-01

    This conference was held on 5-6 Apr 1997 in Alexandria. the specialists discussed heat exchangers, boilers and pressure vessels. more than 200 papers were presented in the meetings. it contains of data, figures and tables

  13. Proceedings of the 3rd international conference on heat exchangers, boilers and pressure vessels (HEB-97). Vol.1 (Research Papers)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This conference was held on 5-6 Apr 1997 in Alexandria. the specialists discussed heat exchangers, boilers and pressure vessels. more than 200 papers were presented in the meetings. it contains of data, figures and tables.

  14. A computer code for analysis of severe accidents in LWRs

    International Nuclear Information System (INIS)

    2001-01-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  15. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  16. A computer code for analysis of severe accidents in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The ICARE2 computer code, developed and validated since 1988 at IPSN (nuclear safety and protection institute), calculates in a mechanistic way the physical and chemical phenomena involved in the core degradation process during possible severe accidents in LWR's. The coupling between ICARE2 and the best-estimate thermal-hydraulics code CATHARE2 was completed at IPSN and led to the release of a first ICARE/CATHARE V1 version in 1999, followed by 2 successive revisions in 2000 and 2001. This documents gathers all the contributions presented at the first international ICARE/CATHARE users'club seminar that took place in November 2001. This seminar was characterized by a high quality and variety of the presentations, showing an increase of reactor applications and user needs in this area (2D/3D aspects, reflooding, corium slumping into the lower head,...). 2 sessions were organized. The first one was dedicated to the applications of ICARE2 V3mod1 against small-scale experiments such as PHEBUS FPT2 and FPT3 tests, PHEBUS AIC, QUENCH experiments, NRU-FLHT-5 test, ACRR-MP1 and DC1 experiments, CORA-PWR tests, and PBF-SFD1.4 test. The second session involved ICARE/CATHARE V1mod1 reactor applications and users'guidelines. Among reactor applications we found: code applicability to high burn-up fuel rods, simulation of the TMI-2 transient, simulation of a PWR-900 high pressure severe accident sequence, and the simulation of a VVER-1000 large break LOCA scenario. (A.C.)

  17. Performance of R + D works to study the behavior during blowdown of LWRs

    International Nuclear Information System (INIS)

    Kaffanke, E.

    1980-03-01

    The aim of the RS 16B experiments is to study thermohydraulic behavior during blowdown of pressure vessels (with internals) of BWRS or PWR-steam generators. A small scale vessel will be used for simulation and measurement. Three series of experiments with varying paramters will be carried out in order to gain insight into flow behavior and loadings on internals caused by steam and feedwater line breaks. The experiments, besides supplying experimental results, will also serve to verify mathematical codes that have been developed to describe such highly complex phenomena. The contributions was as follows: 1) Planning and design of models to simulate BWR-internals. 2) Pre-test calculation of the effects of pressure differences on these models. 3) Post-test calculations upon completion of experiments. Since the beginning of the RS 16B-project in 1972, blowdown effects due to steam and feedwater line breaks have been calculated. A version of the short time computer program LAMB was used for the predictions. This program is an accepted code for the licensing procedure where blowdown analysis of BWR and PWR-steam generator is involved. (orig./HP) [de

  18. Providing Pressurized Gasses to the International Space Station (ISS): Developing a Composite Overwrapped Pressure Vessel (COPV) for the Safe Transport of Oxygen and Nitrogen

    Science.gov (United States)

    Kezirian, Michael; Cook, Anthony; Dick, Brandon; Phoenix, S. Leigh

    2012-01-01

    To supply oxygen and nitrogen to the International Space Station, a COPV tank is being developed to meet requirements beyond that which have been flown. In order to "Ship Full' and support compatibility with a range of launch site operations, the vessel was designed for certification to International Standards (ISO) that have a different approach than current NASA certification approaches. These requirements were in addition to existing NASA certification standards had to be met. Initial risk-reduction development tests have been successful. Qualification is in progress.

  19. Maury Journals - US Vessels

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — U.S. vessels observations, after the 1853 Brussels Conference that set International Maritime Standards, modeled after Maury Marine Standard Observations.

  20. Final report on the reactor pressure vessel pressurized-thermal-shock. International comparative assessment study (RPV PTS ICAS)

    International Nuclear Information System (INIS)

    Sievers, J.; Schulz, H.; Bass, R.; Pugh, C.

    1999-10-01

    A summary of the recently completed International Comparative Assessment Study of Pressurized-Thermal-Shock in Reactor Pressure Vessels (RPV PTS ICAS) is presented here to record the results in actual and comparative fashions. Within the DFM task, where account was taken of material properties and boundary conditions, reasonable agreement was obtained in linear-elastic and elastic-plastic analysis results. Linear elastic analyses and J-estimation schemes were shown to provide conservative estimates of peak crack driving force when compared with those obtained using complex three-dimensional (3D) finite element analyses. Predictions of RT NDT generally showed less scatter than that observed in crack driving force calculations due to the fracture toughness curve used for fracture assessment in the transition temperature region. Observed scatter in some analytical results could be traced mainly to a misinterpretation of the thermal expansion coefficient data given for the cladding and base metal. Also, differences in some results could be due to a quality assurance problem related to procedures for approximating the loading data given in the Problem Statement. For the PFM task, linear-elastic solutions were again shown to be conservative with respect to elastic-plastic solutions (by a factor of 2 to 4). Scatter in solutions obtained using the same computer code was generally attributable to differences in input parameters, e.g. standard deviations for the initial value of RT NDT , as well as for nickel and copper content. In the THM task, while there was a high degree of scatter during the early part of the transient, reasonable agreement in results was obtained during the latter part of the transient. Generally, the scatter was due to differences in analytical approaches used by participants, which included correlation-based engineering methods, system codes and three-dimensional computational fluids dynamics codes. Some of the models used to simulate condensation

  1. Application of Dr. Mainte, integrated simulator of maintenance optimization, to LWRs

    International Nuclear Information System (INIS)

    Isobe, Yoshihiro; Sagisaka, Mitsuyuki; Etoh, Junji; Matsunaga, Takashi; Kosaka, Toru; Matsumoto, Satoshi; Yoshimura, Shinobu

    2015-01-01

    Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) is based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for typical components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability, economic rationality, environmental impact and social acceptance. For the further improvement of the safety and availability of nuclear power plants, the effect of human error and its reduction on the optimization of maintenance activities have been studied. In addition, an approach of reducing human error is proposed. (author)

  2. PDX vacuum vessel stress analysis

    International Nuclear Information System (INIS)

    Nikodem, Z.D.

    1975-01-01

    A stress analysis of PDX vacuum vessel is described and the summary of results is presented. The vacuum vessel is treated as a toroidal shell of revolution subjected to an internal vacuum. The critical buckling pressure is calculated. The effects of the geometrical discontinuity at the juncture of toroidal shell head and cylindrical outside wall, and the concavity of the cylindrical wall are examined. An effect of the poloidal field coil supports and the vessel outside supports on the stress distribution in the vacuum vessel is determined. A method evaluating the influence of circular ports in the vessel wall on the stress level in the vessel is outlined

  3. Research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Rao, P.S.

    The role of the research vessels as a tool for marine research and exploration is very important. Technical requirements of a suitable vessel and the laboratories needed on board are discussed. The history and the research work carried out...

  4. Status and subjects of thermal-hydraulic analysis for next-generation LWRs

    International Nuclear Information System (INIS)

    2000-03-01

    The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)

  5. Influence of localized plasticity on Stress Corrosion Cracking of austenitic stainless steel. Application to IASCC of internals reactor core vessels

    International Nuclear Information System (INIS)

    Cisse, Sarata

    2012-01-01

    The surface conditions of the 316L screw connecting vessel internals of the primary circuit of PWR (pressurized water reactor) corresponds to a grinding condition. These screws are affected by the IASCC (Irradiation Assisted Stress Corrosion Cracking). Initiation of cracking depends on the surface condition but also on the external oxidation and interactions of oxide layer with the deformation bands. The first objective of this study is to point the influence of surface condition on the growth kinetic of oxide layer, and the surface reactivity of 304, 316 stainless steel grade exposed to PWR primary water at 340 C. The second objective is to determine influence of strain localization on the SCC of austenitic stainless steels in PWR primary water. Indeed, the microstructure of irradiated 304, 316 grades correspond to a localized deformation in deformation bands free of radiation defects. In order to reproduce that microstructure without conducting irradiations, low cycle fatigue tests at controlled stain amplitude are implemented for the model material of the study (A286 austenitic stainless steel hardened by the precipitation of phase γ'Ni3(Ti, Al)). During the mechanical cycling (after the first hardening cycles), the precipitates are dissolved in slip bands leading to the localization of the deformation. Once the right experimental conditions in low cycle fatigue obtained (for localized microstructure), interactions oxidation / deformation bands are studied by oxidizing pre deformed samples containing deformation bands and non deformed samples. The tensile tests at a slow strain rate of 8 x 10 -8 /s are also carried out on pre deformed samples and undeformed samples. The results showed that surface treatment induces microstructural modifications of the metal just under the oxide layer, leading to slower growth kinetics of the oxide layer. However, surface treatment accelerates development of oxides penetrations in metal under the oxide layer. As example, for

  6. Mobile nuclear reactor containment vessel

    International Nuclear Information System (INIS)

    Thompson, R.E.; Spurrier, F.R.; Jones, A.R.

    1978-01-01

    A containment vessel for use in mobile nuclear reactor installations is described. The containment vessel completely surrounds the entire primary system, and is located as close to the reactor primary system components as is possible in order to minimize weight. In addition to being designed to withstand a specified internal pressure, the containment vessel is also designed to maintain integrity as a containment vessel in case of a possible collision accident

  7. Transient Safety Analysis of Fast Spectrum TRU Burning LWRs with Internal Blankets

    Energy Technology Data Exchange (ETDEWEB)

    Downar, Thomas [Univ. of Michigan, Ann Arbor, MI (United States); Zazimi, Mujid [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hill, Bob [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-31

    The objective of this proposal was to perform a detailed transient safety analysis of the Resource-Renewable BWR (RBWR) core designs using the U.S. NRC TRACE/PARCS code system. This project involved the same joint team that has performed the RBWR design evaluation for EPRI and therefore be able to leverage that previous work. And because of their extensive experience with fast spectrum reactors and parfait core designs, ANL was also part the project team. The principal outcome of this project was the development of a state-of-the-art transient analysis capability for GEN-IV reactors based on Monte Carlo generated cross sections and the US NRC coupled code system TRACE/PARCS, and a state-of-the-art coupled code assessment of the transient safety performance of the RBWR.

  8. THE ANALYSIS OF FOREIGN-VESSEL SINKING AS AN EFFORT BY THE GOVERNMENT OF INDONESIA TO COMBAT IUU FISHING PURSUANT TO INTERNATIONAL LAW

    Directory of Open Access Journals (Sweden)

    Kristiyanto - Kristiyanto

    2015-12-01

    Full Text Available As an archipelagic state, Indonesia possesses some of the most abundant fishery resources in the world. Geographically, Indonesia’s strategic location makes it a challenge, and it is a shared responsibility for all citizens to preserve and conserve these resources. The strategic location and rich biological as well as non-biological marine resources automatically attract foreign vessels to carry out IUU fishing activities, particularly in the area of ZEEI (Indonesian Exclusive Economic Zone. The Government of Indonesia has taken various preventive measures to combat IUU fishing practices through bilateral cooperations and various laws. In addition, the Government has also taken some repressive efforts by burning and sinking foreign vessels. In this study, the researcher will analyze the governmental action pursuant to international law and examine the extent to which the sinking of the ship is effective from the perspective of international law. This study will be conducted using normative and juridical approach by reviewing and analyzing various national and international legal instruments related to IUU fishing. We hope that this study will be able to deliver theoretical and practical benefits for students and other researchers who are interested in the issue of IUU fishing practices.   Keywords : IUU fishing, marine resources, archipelagic state.

  9. 33 CFR 82.3 - Pushing vessel and vessel being pushed: Composite unit.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INTERNATIONAL NAVIGATION RULES 72 COLREGS: INTERPRETATIVE RULES § 82.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the 72 COLREGS states that when a pushing vessel and a...

  10. Study of impact of the AP1000{sup Registered-Sign} reactor vessel upper internals design on fuel performance

    Energy Technology Data Exchange (ETDEWEB)

    Xu Yiban; Conner, Michael; Yuan Kun; Dzodzo, Milorad B.; Karoutas, Zeses; Beltz, Steven A.; Ray, Sumit; Bissett, Teresa A. [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China)

    2012-11-15

    One aspect of the AP1000{sup Registered-Sign} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is the reduction in the number of reactor vessel outlet nozzles/hot legs leaving the upper plenum from three to two. With regard to fuel performance, this design difference creates a different flow field in the AP1000 reactor vessel upper plenum (the region above the core). The flow exiting core and entering the upper plenum must turn 90 Degree-Sign , flow laterally through the upper plenum around support structures, and exit through one of the two outlet nozzles. While the flow in the top of the core is mostly axial, there is some lateral flow component as the core flow reacts to the flow field and pressure distribution in the upper plenum. The pressure distribution in the upper plenum varies laterally depending upon various factors including the proximity to the outlet nozzles. To determine how the lateral flow in the top of the AP1000 core compares to current Westinghouse reactors, a computational fluid dynamics (CFD) model of the flow in the upper portion of the AP1000 reactor vessel including the top region of the core, the upper plenum, the reactor vessel outlet nozzles, and a portion of the hot legs was created. Due to geometric symmetry, the computational domain was reduced to a quarter (from the top view) that includes Vulgar-Fraction-One-Quarter of the top of the core, Vulgar-Fraction-One-Quarter of the upper plenum, and Vulgar-Fraction-One-Half of an outlet nozzle. Results from this model include predicted velocity fields and pressure distributions throughout the model domain. The flow patterns inside and around guide tubes clearly demonstrate the influence of lateral flow due to the presence of the outlet nozzles. From these results, comparisons of AP1000 flow versus current Westinghouse plants were performed. Field performance

  11. NDE of stainless steel and on-line leak monitoring of LWRs. Annual report, October 1984-September 1985. Volume 2

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Claytor, T.N.; Mathieson, T.; Prine, D.W.

    1986-02-01

    This progress report summarizes work performed by the Argonne National Laboratory and GARD, Inc. (Division of Chamberlain Mfg. Corp.) as subcontractor on NDE of stainless steel and on-line leak monitoring of LWRs during the 12 months from October 1984 to September 1985. 15 refs., 36 figs

  12. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  13. Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Isabella J van Rooyen

    2012-09-01

    The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

  14. Key physical parameters and temperature reactivity coefficients of the deep burn modular helium reactor fueled with LWRs waste

    Energy Technology Data Exchange (ETDEWEB)

    Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Cetnar, Jerzy E-mail: jerzy@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

    2004-11-01

    We investigated some important neutronic features of the deep burn modular helium reactor (DB-MHR) using the MCNP/MCB codes. Our attention was focused on the neutron flux and its spectrum, capture to fission ratio of {sup 239}Pu and the temperature coefficient of fuel and moderator. The DB-MHR is a graphite-moderated helium-cooled reactor proposed by General Atomic to address the need for a fast and efficient incineration of plutonium for non-proliferation purposes as well as the management of light water reactors (LWRs) waste. In fact, recent studies have shown that the use of the DB-MHR coupled to ordinary LWRs would keep constant the world inventory of plutonium for a reactor fleet producing 400 TW{sub e}/y. In the present studies, the DB-MHR is loaded with Np-Pu driver fuel (DF) with an isotopic composition corresponding to LWRs spent fuel waste. DF uses fissile isotopes (e.g. {sup 239}Pu and {sup 241}Pu), previously generated in the LWRs, and maintains criticality conditions in the DB-MHR. After an irradiation of three years, the spent DF is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). TF mainly contains non-fissile actinides which undergo neutron capture and transmutation during the subsequent three-year irradiation in the DB-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. After extraction of the spent TF, irradiated for three years, over 94% of {sup 239}Pu and 53% of all actinides coming from LWRs waste will have been destroyed in the DB-MHR. In this paper we look at the operation conditions at equilibrium for the DB-MHR and evaluate fluxes and reactivity responses using state of the art 3-D Monte Carlo simulations.

  15. Completion of the fabrication and assembly of the internal parts and pressure vessel of the LABGENE reactor

    International Nuclear Information System (INIS)

    Guimaraes, Leonam dos Santos

    2005-01-01

    The Navy's Technological Center in Sao Paulo (CTMSP) has successfully concluded in 2005 the final assembly of the internals of the Laboratory of Energy Generation's Reactor (LABGENE). This structure together with the fuel elements and the control rods drives mechanisms are part of a PWR type Nuclear Reactor. (author)

  16. RPV-1: A Virtual Test Reactor to simulate irradiation effects in light water reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Jumel, Stephanie; Van-Duysen, Jean Claude

    2005-01-01

    Many key components in commercial nuclear reactors are subject to neutron irradiation which modifies their mechanical properties. So far, the prediction of the in-service behavior and the lifetime of these components has required irradiations in so-called 'Experimental Test Reactors'. This predominantly empirical approach can now be supplemented by the development of physically based computer tools to simulate irradiation effects numerically. The devising of such tools, also called Virtual Test Reactors (VTRs), started in the framework of the REVE Project (REactor for Virtual Experiments). This project is a joint effort among Europe, the United States and Japan aimed at building VTRs able to simulate irradiation effects in pressure vessel steels and internal structures of LWRs. The European team has already built a first VTR, called RPV-1, devised for pressure vessel steels. Its inputs and outputs are similar to those of experimental irradiation programs carried out to assess the in-service behavior of reactor pressure vessels. RPV-1 is made of five codes and two databases which are linked up so as to receive, treat and/or convey data. A user friendly Python interface eases the running of the simulations and the visualization of the results. RPV-1 is sensitive to its inputs (neutron spectrum, temperature, ...) and provides results in conformity with experimental ones. The iterative improvement of RPV-1 has been started by the comparison of simulation results with the database of the IVAR experimental program led by the University of California Santa Barbara. These first successes led 40 European organizations to start developing RPV-2, an advanced version of RPV-1, as well as INTERN-1, a VTR devised to simulate irradiation effects in stainless steels, in a large effort (the PERFECT project) supported by the European Commission in the framework of the 6th Framework Program

  17. International trade and air pollution: estimating the economic costs of air emissions from waterborne commerce vessels in the United States.

    Science.gov (United States)

    Gallagher, Kevin P

    2005-10-01

    Although there is a burgeoning literature on the effects of international trade on the environment, relatively little work has been done on where trade most directly effects the environment: the transportation sector. This article shows how international trade is affecting air pollution emissions in the United States' shipping sector. Recent work has shown that cargo ships have been long overlooked regarding their contribution to air pollution. Indeed, ship emissions have recently been deemed "the last unregulated source of traditional air pollutants". Air pollution from ships has a number of significant local, national, and global environmental effects. Building on past studies, we examine the economic costs of this increasing and unregulated form of environmental damage. We find that total emissions from ships are largely increasing due to the increase in foreign commerce (or international trade). The economic costs of SO2 pollution range from dollars 697 million to dollars 3.9 billion during the period examined, or dollars 77 to dollars 435 million on an annual basis. The bulk of the cost is from foreign commerce, where the annual costs average to dollars 42 to dollars 241 million. For NOx emissions the costs are dollars 3.7 billion over the entire period or dollars 412 million per year. Because foreign trade is driving the growth in US shipping, we also estimate the effect of the Uruguay Round on emissions. Separating out the effects of global trade agreements reveals that the trade agreement-led emissions amounted to dollars 96 to dollars 542 million for SO2 between 1993 and 2001, or dollars 10 to dollars 60 million per year. For NOx they were dollars 745 million for the whole period or dollars 82 million per year. Without adequate policy responses, we predict that these trends and costs will continue into the future.

  18. Multiphase flow in ex-vessel coolability: development of an innovative concept

    International Nuclear Information System (INIS)

    Corradini, Michael L.

    2006-01-01

    The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

  19. Disposition of weapons-grade plutonium in LWRs - a utility perspective

    International Nuclear Information System (INIS)

    Naughton, W.F.

    1996-01-01

    The optimal alternative for the disposition of weapons-grade plutonium (WPu) is to burn it in currently operating light water reactors (LWRs). There are three key aspects associated with the reactor burn option that lead to this conclusion. They are timeliness, acceptable solution, and equity or resource recovery. The National Academy of Sciences report on this subject has thoroughly covered the first two aspects by indicating that there is open-quotes a clear and present dangerclose quotes associated with WPu - the timeliness issue. Also, the report indicates that the spent-fuel standard is the acceptable long-term solution. Both of these aspects are met by the reactor burn option as it exists today; i.e., the reactor option is both timely and obviously meets the spent-fuel standard. The equity or resource recovery aspect is based on the fact that the resources for the development and use of this material as a nuclear deterrent for the past 50 yr was supplied by, in the U.S.'s case, the taxpayers. Only the reactor burn option offers an opportunity for the partial recovery of those resources in the form of electrical energy

  20. 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

  1. Recent operating experience issues with 17-4 PH in LWRs

    International Nuclear Information System (INIS)

    Olender, A.; Gorman, J.; Marks, C.; Ilevbare, G.

    2015-01-01

    The stainless steel 17-4 PH has been used successfully in LWRs for quite some time. Its properties as a precipitation hardening martensitic stainless steel are desirable for high strength, high hardness applications, such as valve stems, bolting, and turbine blades. However, some in-service issues with this material have arisen over time and are addressed in this paper. Although the use of 17-4 PH in high hardness conditions is discouraged as it offers less resistance to SCC than lower hardness conditions, the utilization, and subsequent failure, of such components has occurred as recently as 2007. Thermal embrittlement can increase susceptibility to SCC, another major issue affecting 17-4 PH. The recommended maximum operating temperature is, therefore, 243 C. degrees (470 F. degrees) due to recent failures attributed to this cause. Connections with dissimilar metals have also recently been a problem, as inadequate tolerances have been left between components to allow for differences in coefficients of thermal expansion as the temperature is ramped up to operating conditions. However, with the proper precautions, 17-4 PH can, and has, performed successfully during long-term plant operation. (authors)

  2. Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Croucher, D.W.

    1979-01-01

    The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

  3. Design of A Vibration and Stress Measurement System for an Advanced Power Reactor 1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program

    International Nuclear Information System (INIS)

    Ko, Doyoung; Kim, Kyuhyung

    2013-01-01

    In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea

  4. DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

    Directory of Open Access Journals (Sweden)

    DO-YOUNG KO

    2013-04-01

    Full Text Available In accordance with the US Nuclear Regulatory Commission (US NRC, Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP has been developed for an Advanced Power Reactor 1400 (APR1400. The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment. Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

  5. Irradiated stainless steel material constitutive model for use in the performance evaluation of PWR pressure vessel internals

    Energy Technology Data Exchange (ETDEWEB)

    Rashid, J.Y.; Dunham, R.S. [ANATECH (United States); Demma, A. [Electric Power Research Institute - EPRI (United States)

    2011-07-01

    Demonstration of component functionality requires analytical simulations of reactor internals behavior. Towards that aim, EPRI has undertaken the development of irradiated material constitutive model and damage criteria for use in global and local finite-element based functionality analysis methodology. The constitutive behavioral regimes of irradiated stainless steel types 316 and 304 materials included in the model consist of: elastic-plastic material response considering irradiation hardening of the stress-strain curve, irradiation creep, stress relaxation, and void swelling. IASCC and degradation of ductility with irradiation are the primary damage mechanisms considered in the model. The material behavior model development consists of two parts: the first part is a user-material subroutine that can interface with a general-purpose finite element computer program to adapt it to the special-purpose of functionality analysis of reactor internals. The second part is a user utility in the form of Excel Spread sheets that permit users to extract a given property, e.g. the elastic-plastic stress-strain curve, creep curve, or void-swelling curve, as function of the relevant independent variables. The development of the model takes full advantage of the significant work that has been undertaken within EPRI's Material Reliability Program (MRP) to improve the knowledge of the material properties of irradiated stainless steels. Data from EPRI's MRP database have been utilized to develop equations that characterize the yield strength, ultimate tensile strength, uniform elongation, total elongation, reduction in area, void swelling and irradiation creep of stainless steels in a PWR environment. It is noted that, while the development of the model's equations has been statistically faithful to the material database, approximations were introduced in the model to ensure appropriate conservatism in the model's application consistently with accepted

  6. Surgical dissection of the internal carotid artery under flow control by proximal vessel clamping reduces embolic infarcts during carotid endarterectomy.

    Science.gov (United States)

    Yoshida, Kazumichi; Kurosaki, Yoshitaka; Funaki, Takeshi; Kikuchi, Takayuki; Ishii, Akira; Takahashi, Jun C; Takagi, Yasushi; Yamagata, Sen; Miyamoto, Susumu

    2014-01-01

    To evaluate the efficacy of flow control of the internal carotid artery (ICA) by the clamping of the common carotid artery, external carotid artery, and superior thyroid artery during surgical ICA dissection to reduce ischemic complications after carotid endarterectomy (CEA). Sixty-seven patients (59 men; age, 70.5 ± 6.2 years) who underwent CEA by the same surgeon were retrospectively studied. Both conventional CEA (n = 29) and flow-control CEA (n = 38) were performed with the patient under general anesthesia and with the use of somatosensory-evoked potential and near-infrared spectroscopy monitoring as a guide for selective shunting. The number of new postoperative infarcts was assessed with preoperative and postoperative diffusion-weighted images (DWIs) obtained within 3 days of surgery. In addition to surgical technique, the effects of the following factors on new infarcts also were examined: age, side of ICA stenosis, high-grade stenosis, symptoms, and application of shunting. New postoperative DWI lesions were observed in 7 of 67 patients (10.4%), and none of them was symptomatic. With respect to operative technique, the incidence rate of DWI spots was significantly lower in the flow-control group (2.6%) than in the conventional group (20.7%), odds ratio: 0.069; 95% confidence interval: 0.006-0.779; P = 0.031). On multiple logistic regression analysis, age, side of ICA stenosis, high-grade stenosis, symptoms, and the use of internal shunting did not have significant effects on new postoperative DWI lesions, whereas technique did have an effect. The proximal flow-control technique for CEA helps avoid embolic complications during surgical ICA dissection. Copyright © 2014 Elsevier Inc. All rights reserved.

  7. Introduction [to tenth anniversary report of the International Cooperative Group on Cyclic Crack Growth in pressure vessel steels

    International Nuclear Information System (INIS)

    Tomkins, B.

    1989-01-01

    The formation and development of the International Cooperative Group on Cyclic Crack Growth Rate (ICCGR) is outlined. By 1976, it had become apparent that the number of variables that affected the cyclic crack growth rate in reactor water was large and that the rate of data generation was very slow, because it was the low frequency regime that was of major practical importance. A clear need was recognised for a forum to exchange ideas and data, but most important of all to explore collaborative testing to minimise duplication and achieve economies. It was from the outset recognised as a complex and therefore very expensive materials testing area. In response to this situation, it was agreed to set up a group which was formally chartered as the ICCGR in 1978. What had begun as a sharing of views rapidly became an active collaborative group concerned to resolve three issues: 1. development of a consistent method for data reduction; 2. the practice of consistent testing methods; 3. a physical understanding of the underlying mechanisms involved in the process. Three Task Groups were eventually formed to address these issues; Test Methods, Mechanisms, Data Collection and Evaluation. (author)

  8. Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Bischoff, Jeremy; Gandrille, Pascal; Forgeron, Thierry; Brachet, Jean-Christophe; Lorrette, Christophe; Valot, C.; Freyss, M.; Braun, J.; Sauder, C.; Moatti, Marie; Waeckel, Nicolas; Ambard, Antoine; Pasamehmetoglu, Kemal; Johnston, Emma; Bragg-Sitton, Shannon M.; Kurata, M.; Hallstadius, Lars; Ohta, H.; Ogata, T.; Besmann, T.; Chauvin, Nathalie; Cornet, Stephanie; Massara, S.; Kohyama, Akira; Kishimoto, Hirotatsu; Park, Joon Soo; Nakazato, Naofumi; Hayasaka, Daisuke; Asakura, Yuuki; Kanda, Chisato; Kohyama, Akira; Terrani, Kurt; Katoh, Yutai; Yamamoto, Yukinori; Field, Kevin; Snead, Lance; Hu, Xunxiang; Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Pint, Bruce A.; Besmann, T.; Steinbrueck, M.; Grosse, M.; Jianu, A.; Weisenburger, A.; Avincola, V.; Ahmad, S.; Tang, C.; Heuser, Brent J.; Sickafus, Kurt; Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun; Lee, B.O.; Van Nieuwenhove, Rudi; Kim, Young; Rebak, Raul; Dolly, Evan; Dolley, E.J.; Rebak, R.B.; Maloy, Stu; Yang, Jae-Ho; Kim, Dong-Joo; Kim, Keon-Sik; Koo, Yang-Hyun; Lee, Won Jae; Tulenko, James S.; Puide, Mattias; Liu, T.; Gueneau, C.; Gosse, S.; Dupin, N.; Barber, D.; Corcoran, E.; Dumas, J.C.; Hania, R.; Kaye, M.; Turchi, P.

    2015-03-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Task Force 1 (Systems assessment) meeting, 3-4 March 2015: - French evaluation of ATF Concepts (J. Bischoff, AREVA); - Technology Readiness Levels - TRL - for Fuels (K. Pasamehmetoglu, INL); - TRL-definition for advanced fuel concept applied for commercial LWRs in Japan (M. Kurata, JAEA); - Application of TRLs in NNL (E. Johnston, NNL); - Technology Readiness Levels for Advanced Nuclear Fuel and Materials (S. Bragg-Sitton, INL); 1a - Definition of the illustrative scenarios: - AREVA's proposal concerning scenario for Accident Tolerant Fuel studies (P. Gandrille); - A Simplified Accident Scenario (L. Hallstadius); - Accident Scenarios for ATF Performance Evaluation of BWR and PWR in Japan (H. Ohta, CRIEPI); 1b - Related NEA activities: - Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems - WPMM, Expert

  9. Non-gated vessel wall imaging of the internal carotid artery using radial scanning and fast spin echo sequence. Evaluation of vessel signal intensity by flow rate at 3.0 tesla

    International Nuclear Information System (INIS)

    Nakamura, Manami; Makabe, Takeshi; Ichikawa, Masaki; Hatakeyama, Ryohei; Sugimori, Hiroyuki; Sakata, Motomichi

    2013-01-01

    Vessel wall imaging using radial scanning does not use a blood flow suppression pulse with gated acquisition. It has been proposed that there may not be a flow void effect if the flow rate is slow; however, this has yet to be empirically tested. To clarify the relationship between the signal intensity of the vessel lumen and the blood flow rate in a flow phantom, we investigated the usefulness of vessel wall imaging at 3.0 tesla (T). We measured the signal intensity while changing the flow rate in the flow phantom. Radial scanning at 1.5 T showed sufficient flow voids at above medium flow rates. There was no significant difference in lumen signal intensity at the carotid artery flow rate. The signal intensity of the vessel lumen decreased sufficiently using the radial scan method at 3.0 T. We thus obtained sufficient flow void effects at the carotid artery flow rate. We conclude this technique to be useful for evaluating plaque if high contrast can be maintained for fixed tissue (such as plaque) and the vessel lumen. (author)

  10. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  11. Development of PIE techniques for irradiated LWR pressure vessel steels

    International Nuclear Information System (INIS)

    Nishi, Masahiro; Kizaki, Minoru; Sukegawa, Tomohide

    1999-01-01

    For the evaluation of safety and integrity of light water reactors (LWRs), various post irradiation examinations (PIEs) of reactor pressure vessel (RPV) steels and fuel claddings have been carried out in the Research Hot Laboratory (RHL). In recent years, the instrumented Charpy impact testing machine was remodeled aiming at the improvement of accuracy and reliability. By this remodeling, absorbed energy and other useful information on impact properties can be delivered from the force-displacement curve for the evaluation of neutron irradiation embrittlement behavior of LWR-RPV steels at one-time striking. In addition, two advanced PIE technologies are now under development. One is the remote machining of mechanical test pieces from actual irradiated pressure vessel steels. The other is development of low-cycle and high-cycle fatigue test technology in order to clarify the post-irradiation fatigue characteristics of structural and fuel cladding materials. (author)

  12. Components of the primary circuit of LWRs. Design, construction and calculation. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 C). The primary circuit as the pressure containment of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

  13. Containment vessel

    International Nuclear Information System (INIS)

    Zbirohowski-Koscia, K.F.; Roberts, A.C.

    1980-01-01

    A concrete containment vessel for nuclear reactors is disclosed that is spherical and that has prestressing tendons disposed in first, second and third sets, the tendons of each set being all substantially concentric and centred around a respective one of the three orthogonal axes of the sphere; the tendons of the first set being anchored at each end at a first anchor rib running around a circumference of the vessel, the tendons of the second set being anchored at each end at a second anchor rib running around a circumference of the sphere and disposed at 90 0 to the first rib, and the tendons of the third set being anchored some to the first rib and the remainder to the second rib. (author)

  14. Nuclear power plant pressure vessels. Inservice inspections

    International Nuclear Information System (INIS)

    1995-01-01

    The requirements for the planning and reporting of inservice inspections of nuclear power plant pressure vessels are presented. The guide specifically applies to inservice inspections of Safety class 1 and 2 nuclear power plant pressure vessels, piping, pumps and valves plus their supports and reactor pressure vessel internals by non- destructive examination methods (NDE). Inservice inspections according to the Pressure Vessel Degree (549/73) are discussed separately in the guide YVL 3.0. (4 refs.)

  15. Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

    International Nuclear Information System (INIS)

    Beller, D.E.; Sailor, W.C.; Venneri, F.

    1998-01-01

    A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO 2 -fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO 2 fuel cycle

  16. Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

  17. Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Bragg-Sitton, Shannon Michelle [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-09-01

    The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. The Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.

  18. Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

    International Nuclear Information System (INIS)

    Bragg-Sitton, Shannon Michelle

    2016-01-01

    The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force (TF1) includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, and identification of fuel performance and system codes applicable to ATF evaluation. The Cladding and Core Materials (TF2) and Fuel Concepts (TF3) task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment task force is chaired by Shannon Bragg-Sitton (Idaho National Laboratory [INL], U.S.), the Cladding Task Force is chaired by Marie Moatti (Electricite de France [EdF], France), and the Fuels Task Force is chaired by a Masaki Kurata (Japan Atomic Energy Agency [JAEA], Japan). The original Expert Group mandate was established for June 2014 to June 2016. In April 2016 the Expert Group voted to extend the mandate one additional year to June 2017 in order to complete the task force deliverables; this request was subsequently approved by the Nuclear Science Committee. This

  19. Evaluation of alternative partitioning/transmutation scenarios using transmutation in light-water reactors (LWRs)

    International Nuclear Information System (INIS)

    Collins, E.D.; Renier, J.P.; Del Cul, B.; Spencer, B.

    2005-01-01

    Previous Advanced Fuel Cycle Initiative (AFCI) studies were made to assess the effects of the existing accumulation of LWR spent fuel in the United States on the capability to partition/transmute actinides using existing and advanced LWR. The concept of treating the oldest fuel first indicated that significant advantages could be gained in both partitioning, transmutation, and in overall cost reduction. The processing scenarios previously evaluated assumed that (1) 2000 MT/year of spent fuel, irradiated to 45 GWd/MT and decayed for 30 years is processed; (2) recovered plutonium and 90% of the neptunium are transmuted in LWR MOX fuel; and (3) minor actinides, consisting of americium, curium, and 10% of the neptunium are transmuted in burnable poison type targets. Results of the previous study showed that significant benefits could be obtained, including (1) lower costs for partitioning and transmutation and for storage of spent fuel, (2) maintenance of proliferation resistance for the fissile plutonium in spent fuels, and (3) extended lifetime for the repository. The lower costs would be achieved primarily because no capital investment for a special transmuter reactor (fast reactor, accelerator-driven system, etc.) would be required. Instead, only existing and new LWRs would be utilized. Moreover, no new storage capacity could be needed for spent fuels and irradiated targets because the number of spent fuel assemblies would remain the same after the scenario is begun. Even though the total inventory of plutonium would rise during the early cycles, ∼98% of the plutonium would be contained in stored spent fuel and would be protected by high radiation (the Spent Fuel Standard). This is because the spent fuel would be reprocessed and re-irradiated at intervals within which the fission products, 137 Cs and 90 Sr, both with half-lives of ∼ 30 years, exist in significantly high concentrations.The lifetime of the repository would be extended significantly because all of

  20. The Status of Monitoring of Cabon-14 Emissions from LWRs in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Il; Jung, Hai Yong; Yook, Dae Sik; Jeong, Chan Woo; Lee, Byung Soo [KINS, Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear reactions that produce {sup 14}C in reactors with thermal neutrons are {sup 17}O(n,α){sup 14}C, {sup 14}N(n,p){sup 14}C. When it is released in the form of {sup 14}CO{sub 2} through the exhaust port of NPPs, the substance is retained in the bodies of plants and animals through plants' photosynthesis and the ingestion of plants. When plants containing {sup 14}C are ingested as food, the substance is introduced into the human body and causes radiation exposure. Consequently, regulatory agencies require NPP operators to evaluate and to monitor the amount of {sup 14}C emitted into the atmosphere through the exhaust port of NPPs. The present study examines the current state of regulation of {sup 14}C in Korea and the United States, its effect on {sup 14}C emissions, and the off-site dose calculation of local resident based on evaluations recently conducted at LWRs domestically. In the present study, Korean and U. S. standards for {sup 14}C emission monitoring were compared and the current state of emissions in Korea was examined. While most U. S. NPPs evaluate {sup 14}C emissions based on theoretical calculations, Korean NPPs do so by directly collecting and measuring samples. When the results of the evaluation of {sup 14}C emissions generated from the OPR-1000 NPP in Korea were recently reviewed, the limit stipulated in Article 6 (Effluent Concentration Limit) and Article 16 (Prevention of Hazards to Environment) of NSSC Notice No. 2013-49 'Standards on Radiation Protection, etc.' was amply satisfied. However, because 90% or more of the effect on the residents' radiation exposure doses was demonstrated to consist of gaseous {sup 14}C effluents, to minimize the emission of {sup 14}C gas, NPP operators must actively study and practice ways of reducing the release of {sup 14}C gas, such as periodically conducting gas stripper operation and the release of {sup 14}C in a liquid form.

  1. Association between proximal internal carotid artery steno-occlusive disease and diffuse wall thickening in its petrous segment: a magnetic resonance vessel wall imaging study

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Xiaoyi; Li, Dongye [Capital Medical University and Beijing Institute for Brain Disorders, Center for Brain Disorders Research, Beijing (China); Tsinghua University School of Medicine, Center for Biomedical Imaging Research, Department of Biomedical Engineering, Beijing (China); Zhao, Huilin [Shanghai Jiao Tong University, Department of Radiology, Renji Hospital, School of Medicine, Shanghai (China); Chen, Zhensen; Qiao, Huiyu; He, Le; Li, Rui [Tsinghua University School of Medicine, Center for Biomedical Imaging Research, Department of Biomedical Engineering, Beijing (China); Cui, Yuanyuan [PLA General Hospital, Department of Radiology, Beijing (China); Zhou, Zechen [Philips Research China, Healthcare Department, Beijing (China); Yuan, Chun [Tsinghua University School of Medicine, Center for Biomedical Imaging Research, Department of Biomedical Engineering, Beijing (China); University of Washington, Department of Radiology, Seattle, WA (United States); Zhao, Xihai [Tsinghua University School of Medicine, Center for Biomedical Imaging Research, Department of Biomedical Engineering, Beijing (China); Beijing Institute for Brain Disorders, Center for Stroke, Beijing (China)

    2017-05-15

    Significant stenosis or occlusion in carotid arteries may lead to diffuse wall thickening (DWT) in the arterial wall of downstream. This study aimed to investigate the correlation between proximal internal carotid artery (ICA) steno-occlusive disease and DWT in ipsilateral petrous ICA. Symptomatic patients with atherosclerotic stenosis (>0%) in proximal ICA were recruited and underwent carotid MR vessel wall imaging. The 3D motion sensitized-driven equilibrium prepared rapid gradient-echo (3D-MERGE) was acquired for characterizing the wall thickness and longitudinal extent of the lesions in petrous ICA and the distance from proximal lesion to the petrous ICA. The stenosis degree in proximal ICA was measured on the time-of-flight (TOF) images. In total, 166 carotid arteries from 125 patients (mean age 61.0 ± 10.5 years, 99 males) were eligible for final analysis and 64 showed DWT in petrous ICAs. The prevalence of severe DWT in petrous ICA was 1.4%, 5.3%, 5.9%, and 80.4% in ipsilateral proximal ICAs with stenosis category of 1%-49%, 50%-69%, 70%-99%, and total occlusion, respectively. Proximal ICA stenosis was significantly correlated with the wall thickness in petrous ICA (r = 0.767, P < 0.001). Logistic regression analysis showed that proximal ICA stenosis was independently associated with DWT in ipsilateral petrous ICA (odds ratio (OR) = 2.459, 95% confidence interval (CI) 1.896-3.189, P < 0.001). Proximal ICA steno-occlusive disease is independently associated with DWT in ipsilateral petrous ICA. (orig.)

  2. Association between proximal internal carotid artery steno-occlusive disease and diffuse wall thickening in its petrous segment: a magnetic resonance vessel wall imaging study

    International Nuclear Information System (INIS)

    Chen, Xiaoyi; Li, Dongye; Zhao, Huilin; Chen, Zhensen; Qiao, Huiyu; He, Le; Li, Rui; Cui, Yuanyuan; Zhou, Zechen; Yuan, Chun; Zhao, Xihai

    2017-01-01

    Significant stenosis or occlusion in carotid arteries may lead to diffuse wall thickening (DWT) in the arterial wall of downstream. This study aimed to investigate the correlation between proximal internal carotid artery (ICA) steno-occlusive disease and DWT in ipsilateral petrous ICA. Symptomatic patients with atherosclerotic stenosis (>0%) in proximal ICA were recruited and underwent carotid MR vessel wall imaging. The 3D motion sensitized-driven equilibrium prepared rapid gradient-echo (3D-MERGE) was acquired for characterizing the wall thickness and longitudinal extent of the lesions in petrous ICA and the distance from proximal lesion to the petrous ICA. The stenosis degree in proximal ICA was measured on the time-of-flight (TOF) images. In total, 166 carotid arteries from 125 patients (mean age 61.0 ± 10.5 years, 99 males) were eligible for final analysis and 64 showed DWT in petrous ICAs. The prevalence of severe DWT in petrous ICA was 1.4%, 5.3%, 5.9%, and 80.4% in ipsilateral proximal ICAs with stenosis category of 1%-49%, 50%-69%, 70%-99%, and total occlusion, respectively. Proximal ICA stenosis was significantly correlated with the wall thickness in petrous ICA (r = 0.767, P < 0.001). Logistic regression analysis showed that proximal ICA stenosis was independently associated with DWT in ipsilateral petrous ICA (odds ratio (OR) = 2.459, 95% confidence interval (CI) 1.896-3.189, P < 0.001). Proximal ICA steno-occlusive disease is independently associated with DWT in ipsilateral petrous ICA. (orig.)

  3. Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

  4. Resonance self-shielding method using resonance interference factor library for practical lattice physics computations of LWRs

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

    2016-01-01

    This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)

  5. Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

    International Nuclear Information System (INIS)

    Pope, Michael A.; Sen, R. Sonat; Boer, Brian; Ougouag, Abderrafi M.; Youinou, Gilles

    2011-01-01

    The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

  6. 75 FR 56015 - Vessel Inspection Alternatives

    Science.gov (United States)

    2010-09-15

    ... DEPARTMENT OF HOMELAND SECURITY Coast Guard 46 CFR Part 8 Vessel Inspection Alternatives CFR... Certificate; (ii) International Tonnage Certificate; (iii) Cargo Ship Safety Construction Certificate; (iv) Cargo Ship Safety Equipment Certificate; and (v) International Oil Pollution Prevention Certificate; and...

  7. Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II; Desgaste placas tubos guia barras de control interno superior vasija del reactor C.N. Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-07-01

    The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

  8. Application of Cherenkov light observation to reactor measurements (3). Evaluation of spent fuel elements of LWRs with Cherenkov light estimation system

    International Nuclear Information System (INIS)

    Yamamoto, Keiichi; Takeuchi, Tomoaki; Tsuchiya, Kunihiko; Hayashi, Takayasu; Kosuge, Fumiaki

    2016-11-01

    Development of the reactor measurement system has been carried out to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. This report summarized the modification of Cherenkov light estimation system described JAEA-Testing 2015-001 and the result of the burn-up evaluation by Cherenkov light image emitted from spent fuel elements of LWRs with the modified system. (author)

  9. International

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    This rubric reports on 10 short notes about international economical facts about nuclear power: Electricite de France (EdF) and its assistance and management contracts with Eastern Europe countries (Poland, Hungary, Bulgaria); Transnuclear Inc. company (a 100% Cogema daughter company) acquired the US Vectra Technologies company; the construction of the Khumo nuclear power plant in Northern Korea plays in favour of the reconciliation between Northern and Southern Korea; the delivery of two VVER 1000 Russian reactors to China; the enforcement of the cooperation agreement between Euratom and Argentina; Japan requested for the financing of a Russian fast breeder reactor; Russia has planned to sell a floating barge-type nuclear power plant to Indonesia; the control of the Swedish reactor vessels of Sydkraft AB company committed to Tractebel (Belgium); the renewal of the nuclear cooperation agreement between Swiss and USA; the call for bids from the Turkish TEAS electric power company for the building of the Akkuyu nuclear power plant answered by three candidates: Atomic Energy of Canada Limited (AECL), Westinghouse (US) and the French-German NPI company. (J.S.)

  10. 14. Internal symposium on secular change of structural materials for nuclear energy. Secular change mechanism in light water reactor environment

    International Nuclear Information System (INIS)

    1993-01-01

    At this symposium, lectures were given on the embrittlement by neutron irradiation of LWR pressure vessel steel, the effect that neutron irradiation exerts to austenitic stainless steel becoming sensitive, the mechanism of the occurrence and development of stress corrosion cracking in the water environment of LWRs, the effect that the water environment of LWRs exerts to fatigue life, and the environment-promoted cracking in LWR environment and its forecast. Thereafter, panel discussion was held by the lecturers. In this book, the summaries of the lectures are collected. (K.I.)

  11. Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

    2018-01-01

    Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

  12. Vessel Operator System

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Operator cards are required for any operator of a charter/party boat and or a commercial vessel (including carrier and processor vessels) issued a vessel permit from...

  13. Studies on the feasibility of the LWRs waste-thorium in-core fuel cycle in the Gas Turbine-Modular Helium Reactor

    International Nuclear Information System (INIS)

    Talamo, Alberto

    2006-01-01

    The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233 U by fertile thorium. Since a mixture of pure 239 Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29eV of the fission and capture microscopic cross sections of 239 Pu diminishes also the absorption rate of fertile 232 Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232 Th is about 80-100 times smaller than the fission one of 239 Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233 U represents only a small fraction of the depleted 239 Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one. (author)

  14. Life extension of the BR2 aluminium vessel

    International Nuclear Information System (INIS)

    Koonen, E.; Fabry, A.; Chaouadi, R.; Verwerft, M.; Raedt, C. de; Winckel, S. van; Wacquier, W.; Dadoumont, J.; Verwimp, A.

    2000-01-01

    The BR2 reactor has recently undergone a major refurbishment comprising the replacement of all vessel internals. The vessel itself however was not replaced. An important requalification programme has been executed to prove that the vessel would remain fit during the contemplated life extension period of BR2. Representative material samples could be obtained from the shroud surrounding the vessel. A comprehensive in-service inspection was carried out and a vessel surveillance programme has been established. (author)

  15. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10 6 R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  16. High gamma-rays irradiation tests of critical components for ITER (International Thermonuclear Experimental Reactor) in-vessel remote handling system

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi [Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, Naka, Ibaraki (Japan)] [and others

    1999-02-01

    In ITER, the in-vessel remote handling is inevitably required to assemble and maintain the activated in-vessel components due to deuterium and tritium operation. Since the in-vessel remote handling system has to be operated under the intense of gamma ray irradiation, the components of the remote handling system are required to have radiation hardness so as to allow maintenance operation for a sufficient length of time under the ITER in-vessel environments. For this, the Japan, European and Russian Home Teams have extensively conducted gamma ray irradiation tests and quality improvements including optimization of material composition through ITER R and D program in order to develop radiation hard components which satisfy the doses from 10 MGy to 100 MGy at a dose rate of 1 x 10{sup 6} R/h (ITER R and D Task: T252). This report describes the latest status of radiation hard component development which has been conducted by the Japan Home Team in the ITER R and D program. The number of remote handling components tested is about seventy and these are categorized into robotics (Subtask 1), viewing system (Subtask 2) and common components (Subtask 3). The irradiation tests, including commercial base products for screening, modified products and newly developed products to improve the radiation hardness, were carried out using the gamma ray irradiation cells in Takasaki Establishment, JAERI. As a result, the development of the radiation hard components which can be tolerable for high temperature and gamma radiation has been well progressed, and many components, such as AC servo motor with ceramics insulated wire, optical periscope and CCD camera, have been newly developed. (author)

  17. The TPX vacuum vessel and in-vessel components

    International Nuclear Information System (INIS)

    Heitzenroeder, P.; Bialek, J.; Ellis, R.; Kessel, C.; Liew, S.

    1994-01-01

    The Tokamak Physics Experiment (TPX) is a superconducting tokamak with double-null diverters. TPX is designed for 1,000-second discharges with the capability of being upgraded to steady state operation. High neutron yields resulting from the long duration discharges require that special consideration be given to materials and maintainability. A unique feature of the TPX is the use of a low activation, titanium alloy vacuum vessel. Double-wall vessel construction is used since it offers an efficient solution for shielding, bakeout and cooling. Contained within the vacuum vessel are the passive coil system, Plasma Facing Components (PFCs), magnetic diagnostics, and the internal control coils. All PFCs utilize carbon-carbon composites for exposed surfaces

  18. Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

    International Nuclear Information System (INIS)

    Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

    2013-01-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

  19. Multiple shell pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.

    1988-01-01

    A method is described of fabricating a pressure vessel comprising the steps of: attaching a first inner pressure vessel having means defining inlet and outlet openings to a top flange, placing a second inner pressure vessel, having means defining inlet and outlet opening, concentric with and spaced about the first inner pressure vessel and attaching the second inner pressure vessel to the top flange, placing an outer pressure vessel, having inlet and outlet openings, concentric with and spaced apart about the second inner pressure vessel and attaching the outer pressure vessel to the top flange, attaching a generally cylindrical inner inlet conduit and a generally cylindrical inner outlet conduit respectively to the inlet and outlet openings in the first inner pressure vessel, attaching a generally cylindrical outer inlet conduit and a generally cylindrical outer outlet conduit respectively to the inlet and outlet opening in the second inner pressure vessel, heating the assembled pressure vessel to a temperature above the melting point of a material selected from the group, lead, tin, antimony, bismuth, potassium, sodium, boron and mixtures thereof, filling the space between the first inner pressure vessel and the second inner pressure vessel with material selected from the group, filling the space between the second inner pressure vessel and the outer pressure vessel with material selected from the group, and pressurizing the material filling the spaces between the pressure vessels to a predetermined pressure, the step comprising: pressurizing the spaces to a pressure whereby the wall of the first inner pressure vessel is maintained in compression during steady state operation of the pressure vessel

  20. Probabilistic retinal vessel segmentation

    Science.gov (United States)

    Wu, Chang-Hua; Agam, Gady

    2007-03-01

    Optic fundus assessment is widely used for diagnosing vascular and non-vascular pathology. Inspection of the retinal vasculature may reveal hypertension, diabetes, arteriosclerosis, cardiovascular disease and stroke. Due to various imaging conditions retinal images may be degraded. Consequently, the enhancement of such images and vessels in them is an important task with direct clinical applications. We propose a novel technique for vessel enhancement in retinal images that is capable of enhancing vessel junctions in addition to linear vessel segments. This is an extension of vessel filters we have previously developed for vessel enhancement in thoracic CT scans. The proposed approach is based on probabilistic models which can discern vessels and junctions. Evaluation shows the proposed filter is better than several known techniques and is comparable to the state of the art when evaluated on a standard dataset. A ridge-based vessel tracking process is applied on the enhanced image to demonstrate the effectiveness of the enhancement filter.

  1. Improvement to reactor vessel

    International Nuclear Information System (INIS)

    1974-01-01

    The vessel described includes a prestressed concrete vessel containing a chamber and a removable cover closing this chamber. The cover is in concrete and is kept in its closed position by main and auxiliary retainers, comprising fittings integral with the concrete of the vessel. The auxiliary retainers pass through the concrete of the cover. This improvement may be applied to BWR, PWR and LMFBR type reactor vessel [fr

  2. ALICE HMPID Radiator Vessel

    CERN Document Server

    2003-01-01

    View of the radiator vessels of the ALICE/HMPID mounted on the support frame. Each HMPID module is equipped with 3 indipendent radiator vessels made out of neoceram and fused silica (quartz) windows glued together. The spacers inside the vessel are needed to stand the hydrostatic pressure. http://alice-hmpid.web.cern.ch/alice-hmpid

  3. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    International Nuclear Information System (INIS)

    Ragusa, Jean; Vierow, Karen

    2011-01-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  4. Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Ragusa, Jean; Vierow, Karen

    2011-09-01

    The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

  5. The LWRs Fuel Cycle/What Could Change in the 90's

    International Nuclear Information System (INIS)

    Lamorlette, G.

    1990-01-01

    One of the main characteristics of the nuclear power industry is the size of its investments and the time it takes to build its facilities. They must operate long enough to yield a return on the sums invested while producing electricity at a competitive price. A recent study by OECD/N/A and the International Energy Agency shows that in most countries committed to long-term nuclear programmes, the nuclear kWh has a significant economic edge over the oil-generated kWh. Most of these countries are poor in natural resources, and this inspired their commitment. After one or two decades, nuclear power is proving to be a good choice. The nuclear fuel cycle industry has the same characteristics. From the discovery of a Uranium deposit to the opening of mine takes ten years, another the years to build a reprocessing plant. But a closer analysis of products and services are well known, the way in which producers will be able to meet them, and the utilities' procurement strategy, are today still hard to predict. We shall here attempt to identify, in the light of the experience of the 1980s, a few events that might lead to major changes in trade in products and services in three fields with which Cogema is quite familiar : Uranium, enrichment and the spent fuels management

  6. Benchmarks for Uncertainty Analysis in Modelling (UAM) for the Design, Operation and Safety Analysis of LWRs - Volume I: Specification and Support Data for Neutronics Cases (Phase I)

    International Nuclear Information System (INIS)

    Ivanov, K.; Avramova, M.; Kamerow, S.; Kodeli, I.; Sartori, E.; Ivanov, E.; Cabellos, O.

    2013-01-01

    The objective of the OECD LWR UAM activity is to establish an internationally accepted benchmark framework to compare, assess and further develop different uncertainty analysis methods associated with the design, operation and safety of LWRs. As a result, the LWR UAM benchmark will help to address current nuclear power generation industry and regulation needs and issues related to practical implementation of risk-informed regulation. The realistic evaluation of consequences must be made with best-estimate coupled codes, but to be meaningful, such results should be supplemented by an uncertainty analysis. The use of coupled codes allows us to avoid unnecessary penalties due to incoherent approximations in the traditional decoupled calculations, and to obtain more accurate evaluation of margins regarding licensing limit. This becomes important for licensing power upgrades, improved fuel assembly and control rod designs, higher burn-up and others issues related to operating LWRs as well as to the new Generation 3+ designs being licensed now (ESBWR, AP-1 000, EPR-1 600, etc.). Establishing an internationally accepted LWR UAM benchmark framework offers the possibility to accelerate the licensing process when using best estimate methods. The proposed technical approach is to establish a benchmark for uncertainty analysis in best-estimate modelling and coupled multi-physics and multi-scale LWR analysis, using as bases a series of well-defined problems with complete sets of input specifications and reference experimental data. The objective is to determine the uncertainty in LWR system calculations at all stages of coupled reactor physics/thermal hydraulics calculations. The full chain of uncertainty propagation from basic data, engineering uncertainties, across different scales (multi-scale), and physics phenomena (multi-physics) will be tested on a number of benchmark exercises for which experimental data are available and for which the power plant details have been

  7. European Union research in safety of LWRs with emphasis on accident management measures

    International Nuclear Information System (INIS)

    Bermejo, J.M.; Van Goethem, G.

    1998-01-01

    On April 26th 1994 the European Union (EU) adopted via a Council Decision a multiannual programme for community activities in the field of nuclear research and training for the period 1994 to 1998. This programme continued the EU research activities of the 1992-1995 Reactor Safety Programme which was carried out as a Reinforced Concerted Action (RCA), and which covered mainly research activities in the area of severe accident phenomena, both for the existing and next-generation light water reactors. The 1994-1998 Framework programme includes activities regarding Research and Technological Development (R and TD), such as demonstration projects, international cooperation, dissemination and optimization of results, as well as training, in a wide range of scientific fields, including nuclear fission safety and controlled thermonuclear fusion. The 1994-1998 specific programme for nuclear fission safety has five main activity areas: (i) Exploring Innovative Approaches, (ii) Reactor Safety, (iii) Radioactive Waste Management, Disposal, and Decommissioning, (iv) Radiological Impact on Man and Environment, and (v) Mastering Events of the past. The specific topics included in this work programme were chosen in consultation with the EU Joint Research Centres (JRC), and with experts in the different fields taking into account the needs of the end users of the Community research, i.e. vendors, utilities and licensing and regulators authorities. This paper briefly discusses the objectives and achievements of the 1992-1995 RCA and also describes the projects being (or to be) implemented as part of the 1994-1995 programme in the areas of R eactor Safety/Severe Accidents , particularly those related to Accident Management (AM) Measures. In addition to this, some relevant projects related to AM which have been funded via independent PHARE/TACIS assistance programmes will also be mentioned

  8. CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

    International Nuclear Information System (INIS)

    Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

    2015-01-01

    This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

  9. Study on prestressed concrete reactor vessel structures. II-5: Crack analysis by three dimensional finite elements method of 1/20 multicavity type PCRV subjected to internal pressure

    Science.gov (United States)

    1978-01-01

    A three-dimensional finite elements analysis is reported of the nonlinear behavior of PCRV subjected to internal pressure by comparing calculated results with test results. As the first stage, an analysis considering the nonlinearity of cracking in concrete was attempted. As a result, it is found possible to make an analysis up to three times the design pressure (50 kg/sqcm), and calculated results agree well with test results.

  10. Components of the primary circuit of LWRs. Design, construction and calculation. Draft. Komponenten des Primaerkreises von Leichtwasserreaktoren. Auslegung, Konstruktion und Berechnung. Entwurf

    Energy Technology Data Exchange (ETDEWEB)

    1979-10-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673/sup 0/K (400/sup 0/C). The primary circuit as the pressure continement of the reactor coolant comprises: Reactor pressure vessel (without internals), steam generator (primary loop), pressurizer, reactor coolant pump housing, interconnecting pipings between the components mentioned above and appropriate various valve and instrument casings, pipings branding off from the above components and interconnecting pipings, including the appropriate instrument casings, up to and including the first isolating valve, pressure shielding of control rod drives.

  11. Structural Analysis of the NCSX Vacuum Vessel

    International Nuclear Information System (INIS)

    Fred Dahlgren; Art Brooks; Paul Goranson; Mike Cole; Peter Titus

    2004-01-01

    The NCSX (National Compact Stellarator Experiment) vacuum vessel has a rather unique shape being very closely coupled topologically to the three-fold stellarator symmetry of the plasma it contains. This shape does not permit the use of the common forms of pressure vessel analysis and necessitates the reliance on finite element analysis. The current paper describes the NCSX vacuum vessel stress analysis including external pressure, thermal, and electro-magnetic loading from internal plasma disruptions and bakeout temperatures of up to 400 degrees centigrade. Buckling and dynamic loading conditions are also considered

  12. Design concept for vessels and heat exchangers

    International Nuclear Information System (INIS)

    Elfmann, W.; Ferrari, L.D.B.

    1981-01-01

    A design concept for vessels and heat exchangers against internal and external loads resulting from normal operation and accident is shown. A definition and explanation of the operating conditions and stress levels are given. A description of the type of analysis (stress, fatigue, deformation, stability, earthquake and vibration) is presented in detail, also including technical guidelines which are used for the vessels and heat exchangers and their individual structure parts. (Author) [pt

  13. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

    Pressure vessels are closed containers designed to hold gases or liquids at a pressure substantially different from the ambient pressure. They have a variety of applications in industry, including in oil refineries, nuclear reactors, vehicle airbrake reservoirs, and more. The pressure differential with such vessels is dangerous, and due to the risk of accident and fatality around their use, the design, manufacture, operation and inspection of pressure vessels is regulated by engineering authorities and guided by legal codes and standards. Pressure Vessel Design Manual is a solutions-focused guide to the many problems and technical challenges involved in the design of pressure vessels to match stringent standards and codes. It brings together otherwise scattered information and explanations into one easy-to-use resource to minimize research and take readers from problem to solution in the most direct manner possible. * Covers almost all problems that a working pressure vessel designer can expect to face, with ...

  14. Tumor Blood Vessel Dynamics

    Science.gov (United States)

    Munn, Lance

    2009-11-01

    ``Normalization'' of tumor blood vessels has shown promise to improve the efficacy of chemotherapeutics. In theory, anti-angiogenic drugs targeting endothelial VEGF signaling can improve vessel network structure and function, enhancing the transport of subsequent cytotoxic drugs to cancer cells. In practice, the effects are unpredictable, with varying levels of success. The predominant effects of anti-VEGF therapies are decreased vessel leakiness (hydraulic conductivity), decreased vessel diameters and pruning of the immature vessel network. It is thought that each of these can influence perfusion of the vessel network, inducing flow in regions that were previously sluggish or stagnant. Unfortunately, when anti-VEGF therapies affect vessel structure and function, the changes are dynamic and overlapping in time, and it has been difficult to identify a consistent and predictable normalization ``window'' during which perfusion and subsequent drug delivery is optimal. This is largely due to the non-linearity in the system, and the inability to distinguish the effects of decreased vessel leakiness from those due to network structural changes in clinical trials or animal studies. We have developed a mathematical model to calculate blood flow in complex tumor networks imaged by two-photon microscopy. The model incorporates the necessary and sufficient components for addressing the problem of normalization of tumor vasculature: i) lattice-Boltzmann calculations of the full flow field within the vasculature and within the tissue, ii) diffusion and convection of soluble species such as oxygen or drugs within vessels and the tissue domain, iii) distinct and spatially-resolved vessel hydraulic conductivities and permeabilities for each species, iv) erythrocyte particles advecting in the flow and delivering oxygen with real oxygen release kinetics, v) shear stress-mediated vascular remodeling. This model, guided by multi-parameter intravital imaging of tumor vessel structure

  15. Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals

    Energy Technology Data Exchange (ETDEWEB)

    Xie, Linjun [Zhejiang Univ. of Technology, Hangzhou (China). College of Mechanical Engineering; Xue, Guohong; Zhang, Ming [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China)

    2017-11-15

    The friction force between the contact surfaces of a reactor internal hold-down spring (HDS) and core barrel flanges can directly influence the axial stiffness of an HDS. However, friction coefficient cannot be obtained through theoretical analysis. This study performs a mathematical deduction of the physical model of an HDS. Moreover, a mathematical model of axial load P, displacement δ, and flexibility coefficient is established, and a set of test apparatuses is designed to simulate the preloading process of the HDS. According to the experimental research and theoretical analysis, P-δ curves and the flexibility coefficient λ are obtained in the loading processes of the HDS. The friction coefficient f of the M1000 HDS is further calculated as 0.224. The displacement limit load value (4,638 kN) can be obtained through a displacement limit experiment. With the friction coefficient considered, the theoretical load is 4,271 kN, which is relatively close to the experimental result. Thus, the friction coefficient exerts an influence on the displacement limit load P. The friction coefficient should be considered in the design analysis for HDS.

  16. Friction coefficient and limiter load test analysis by flexibility coefficient model of Hold-Down Spring of nuclear reactor vessel internals

    International Nuclear Information System (INIS)

    Xie, Linjun

    2017-01-01

    The friction force between the contact surfaces of a reactor internal hold-down spring (HDS) and core barrel flanges can directly influence the axial stiffness of an HDS. However, friction coefficient cannot be obtained through theoretical analysis. This study performs a mathematical deduction of the physical model of an HDS. Moreover, a mathematical model of axial load P, displacement δ, and flexibility coefficient is established, and a set of test apparatuses is designed to simulate the preloading process of the HDS. According to the experimental research and theoretical analysis, P-δ curves and the flexibility coefficient λ are obtained in the loading processes of the HDS. The friction coefficient f of the M1000 HDS is further calculated as 0.224. The displacement limit load value (4,638 kN) can be obtained through a displacement limit experiment. With the friction coefficient considered, the theoretical load is 4,271 kN, which is relatively close to the experimental result. Thus, the friction coefficient exerts an influence on the displacement limit load P. The friction coefficient should be considered in the design analysis for HDS.

  17. International nuclear technology transfer

    International Nuclear Information System (INIS)

    Cartwright, P.; Rocchio, J.P.

    1978-01-01

    Light water reactors (LWRs), originally developed in the United States, became the nuclear workhorses for utilities in Europe and Japan largely because the U.S. industry was willing and able to transfer its nuclear know-how abroad. In this international effort, the industry had the encouragement and support of the U.S. governement. In the case of the boiling water reactor (BWR) the program for technology transfer was developed in response to overseas customer demands for support in building local designs and manufacturing capabilities. The principal vehicles have been technology exchange agreements through which complete engineering and manufacturing information is furnished covering BWR systems and fuel. Agreements are held with companies in Germany, Japan, Italy, and Sweden. In recent years, a comprehensive program of joint technology development with overseas manufacturers has begun. The rapidly escalating cost of nuclear research and development make it desirable to minimize duplication of effort. These joint programs provide a mechanism for two or more parties jointly to plan a development program, assign work tasks among themselves, and exchange test results. Despite a slower-than-hoped-for start, nuclear power today is playing a significant role in the economic growth of some developing countries, and can continue to do so. Roughly half of the 23 free world nations that have adopted LWRs are developing countries

  18. Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

    International Nuclear Information System (INIS)

    Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

    2016-01-01

    Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

  19. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  20. 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.

  1. Proceedings of the CSNI workshop on International Standard Problem 48 - Analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    International Nuclear Information System (INIS)

    2005-01-01

    At the CSNI meeting in June 2002, the proposal for an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test was approved. Objectives were to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. The ISP was sponsored by the USNRC, and results had been made available thanks to NUPEC and to the USNRC. Sandia National Laboratory was contracted to manage the technical aspects of the ISP. At the end of the ISP48, a workshop was organized in Lyon, France on April 6-7, 2005 hosted by Electricite de France. Its overall objective was to present results obtained by participants in the ISP 48 and to assess the current practices and the state of the art with respect to the calculation of concrete structures under severe accident conditions. Experience from other areas in civil engineering related to the modelling of complex structures was greatly beneficial to all. Information obtained as a result of this assessment were utilized to develop a consensus on these calculations and identify issues or 'gaps' in the present knowledge for the primary purpose of formulating and prioritizing research needs on this topic. The ISP48 exercise was published in the report referenced NEA/CSNI/R(2005)5 in 3 volumes. Volume 1 contains the synthesis of the exercise; Volumes 2 and 3 contain individual contributions of participating organizations. The CSNI Working Group on the Integrity and Ageing and in particular its sub-group on the behaviour of concrete structures has produced extensive material over the last few years. The complete list of references is given in this document. These proceedings gather the papers and presentations given by the participants at the Lyon workshop

  2. TMI-2 Vessel Investigation Project integration report

    Energy Technology Data Exchange (ETDEWEB)

    Wolf, J. R.; Rempe, J. L.; Stickler, L. A.; Korth, G. E.; Diercks, D. R.; Neimark, L. A.; Akers, D W; Schuetz, B. K.; Shearer, T L; Chavez, S. A.; Thinnes, G. L.; Witt, R. J.; Corradini, M L; Kos, J. A. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel`s condition after the accident.

  3. NCSX Vacuum Vessel Fabrication

    International Nuclear Information System (INIS)

    Viola ME; Brown T; Heitzenroeder P; Malinowski F; Reiersen W; Sutton L; Goranson P; Nelson B; Cole M; Manuel M; McCorkle D.

    2005-01-01

    The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in conjunction with the Oak Ridge National Laboratory (ORNL). The goal of this experiment is to develop a device which has the steady state properties of a traditional stellarator along with the high performance characteristics of a tokamak. A key element of this device is its highly shaped Inconel 625 vacuum vessel. This paper describes the manufacturing of the vessel. The vessel is being fabricated by Major Tool and Machine, Inc. (MTM) in three identical 120 o vessel segments, corresponding to the three NCSX field periods, in order to accommodate assembly of the device. The port extensions are welded on, leak checked, cut off within 1-inch of the vessel surface at MTM and then reattached at PPPL, to accommodate assembly of the close-fitting modular coils that surround the vessel. The 120 o vessel segments are formed by welding two 60 o segments together. Each 60 o segment is fabricated by welding ten press-formed panels together over a collapsible welding fixture which is needed to precisely position the panels. The vessel is joined at assembly by welding via custom machined 8-inch (20.3 cm) wide spacer ''spool pieces''. The vessel must have a total leak rate less than 5 X 10 -6 t-l/s, magnetic permeability less than 1.02(micro), and its contours must be within 0.188-inch (4.76 mm). It is scheduled for completion in January 2006

  4. Radioactive waste processing vessel

    International Nuclear Information System (INIS)

    Hayashi, Masaru; Suzuki, Osamu; Ishizaki, Kanjiro.

    1987-01-01

    Purpose: To obtain a vessel of a reduced weight and with no external leaching of radioactive materials. Constitution: The vessel main body is constituted, for example, with light weight concretes or foamed concretes, particularly, foamed concretes containing fine closed bubbles in the inside. Then, layers having dense texture made of synthetic resin such as polystylene, vinylchloride resin, etc. or metal plate such as stainless plate are integrally disposed to the inner surface of the vessel main body. The cover member also has the same structure. (Sekiya, K.)

  5. Tempest in a vessel

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-01-01

    As the ASN made some statements about anomalies of carbon content in the EPR vessel bottom and top, the author recalls and comments some technical issues to better understand the information published on this topic. He notably addresses the role of the vessel, briefly indicates its operating conditions, shape and structure, and mechanical components for the top, its material and mechanical properties, and test samples used to assess mechanical properties. He also comments the phenomenon of radio-induced embrittlement, the vessel manufacturing process, and evokes the applicable regulations. He quotes and comments statements made by the ASN and Areva which evoke further assessments of the concerned components

  6. A thermal insulation system intended for a prestressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    The description is given of a thermal insulation system withstanding the pressure of a vaporisable fluid for a prestressed concrete vessel, particularly the vessel of a boiling water nuclear reactor. The ring in the lower part of the vessel has, between the fluid inlet pipes and the bottom of the vessel, an annular opening of which the bottom edge is integral with an annular part rising inside the ring and parallel to it. This ring is hermetically connected to the bottom of the vessel and is coated with a metal lagging, at least facing the annular opening. This annular opening is made in the ring half-way up between the fluid inlet pipes and the bottom of the vessel. It is connected to the bottom of the vessel through the internal structure enveloping the reactor core [fr

  7. Cheboygan Vessel Base

    Data.gov (United States)

    Federal Laboratory Consortium — Cheboygan Vessel Base (CVB), located in Cheboygan, Michigan, is a field station of the USGS Great Lakes Science Center (GLSC). CVB was established by congressional...

  8. High Performance Marine Vessels

    CERN Document Server

    Yun, Liang

    2012-01-01

    High Performance Marine Vessels (HPMVs) range from the Fast Ferries to the latest high speed Navy Craft, including competition power boats and hydroplanes, hydrofoils, hovercraft, catamarans and other multi-hull craft. High Performance Marine Vessels covers the main concepts of HPMVs and discusses historical background, design features, services that have been successful and not so successful, and some sample data of the range of HPMVs to date. Included is a comparison of all HPMVs craft and the differences between them and descriptions of performance (hydrodynamics and aerodynamics). Readers will find a comprehensive overview of the design, development and building of HPMVs. In summary, this book: Focuses on technology at the aero-marine interface Covers the full range of high performance marine vessel concepts Explains the historical development of various HPMVs Discusses ferries, racing and pleasure craft, as well as utility and military missions High Performance Marine Vessels is an ideal book for student...

  9. 2011 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  10. 2011 Fishing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  11. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

    This report summarizes testing of an alternate ''pressurized vessel slurry pumping'' apparatus. The principle is similar to rural domestic water systems and ''acid eggs'' used in chemical laboratories in that material is extruded by displacement with compressed air

  12. 2013 Tanker Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  13. Coastal Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains catch (landed catch) and effort for fishing trips made by vessels that have been issued a Federal permit for the Gulf of Mexico reef fish,...

  14. In-vessel tritium

    International Nuclear Information System (INIS)

    Ueda, Yoshio; Ohya, Kaoru; Ashikawa, Naoko; Ito, Atsushi M.; Kato, Daiji; Kawamura, Gakushi; Takayama, Arimichi; Tomita, Yukihiro; Nakamura, Hiroaki; Ono, Tadayoshi; Kawashima, Hisato; Shimizu, Katsuhiro; Takizuka, Tomonori; Nakano, Tomohide; Nakamura, Makoto; Hoshino, Kazuo; Kenmotsu, Takahiro; Wada, Motoi; Saito, Seiki; Takagi, Ikuji; Tanaka, Yasunori; Tanabe, Tetsuo; Yoshida, Masafumi; Toma, Mitsunori; Hatayama, Akiyoshi; Homma, Yuki; Tolstikhina, Inga Yu.

    2012-01-01

    The in-vessel tritium research is closely related to the plasma-materials interaction. It deals with the edge-plasma-wall interaction, the wall erosion, transport and re-deposition of neutral particles and the effect of neutral particles on the fuel recycling. Since the in-vessel tritium shows a complex nonlinear behavior, there remain many unsolved problems. So far, behaviors of in-vessel tritium have been investigated by two groups A01 and A02. The A01 group performed experiments on accumulation and recovery of tritium in thermonuclear fusion reactors and the A02 group studied theory and simulation on the in-vessel tritium behavior. In the present article, outcomes of the research are reviewed. (author)

  15. Reactor pressure vessel support

    International Nuclear Information System (INIS)

    Butti, J.P.

    1977-01-01

    A link and pin support system provides the primary vertical and lateral support for a nuclear reactor pressure vessel without restricting thermally induced radial and vertical expansion and contraction. (Auth.)

  16. 2013 Cargo Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  17. 2013 Fishing Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  18. 2013 Vessel Density

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  19. Ocean Station Vessel

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Ocean Station Vessels (OSV) or Weather Ships captured atmospheric conditions while being stationed continuously in a single location. While While most of the...

  20. Vessel Sewage Discharges

    Science.gov (United States)

    Vessel sewage discharges are regulated under Section 312 of the Clean Water Act, which is jointly implemented by the EPA and Coast Guard. This homepage links to information on marine sanitation devices and no discharge zones.

  1. Reactor pressure vessel design

    International Nuclear Information System (INIS)

    Foehl, J.

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 2, the general principles of reactor pressure vessel design are elaborated. Crack and fracture initiation and propagation are treated in some detail

  2. Graywater Discharges from Vessels

    Science.gov (United States)

    2011-11-01

    metals (e.g., cadmium, chromium, lead, copper , zinc, silver, nickel, and mercury), solids, and nutrients (USEPA, 2008b; USEPA 2010). Wastewater from... flotation ), and disinfection (using ultraviolet light) as compared to traditional Type II MSDs that use either simple maceration and chlorination, or...Coliform Naval Vessels Oceanographic Vessels Small Cruise Ships 25a Vendor 2 Hamann AG Biological Treatment with Dissolved Air Flotation and

  3. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

    Webber, Nels W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-11-25

    Los Alamos National Laboratory in J-1 DARHT Operations Group uses 6ft spherical vessels to contain hazardous materials produced in a hydrodynamic experiment. These contaminated vessels must be analyzed by means of a worker entering the vessel to locate, measure, and document every penetration mark on the vessel. If the worker can be replaced by a highly automated robotic system with a high precision scanner, it will eliminate the risks to the worker and provide management with an accurate 3D model of the vessel presenting the existing damage with the flexibility to manipulate the model for better and more in-depth assessment.The project was successful in meeting the primary goal of installing an automated system which scanned a 6ft vessel with an elapsed time of 45 minutes. This robotic system reduces the total time for the original scope of work by 75 minutes and results in excellent data accumulation and transmission to the 3D model imaging program.

  4. Radiology trainer. Torso, internal organs and vessels

    International Nuclear Information System (INIS)

    Staebler, A.; Ertl-Wagner, B.

    2006-01-01

    This book enables students to simulate examinations. The Radiology Trainer series comprises the whole knowledge of radiology in the form of case studies for self-testing. It is based on the best-sorted German-language collection of radiological examinations of all organ regions. Step by step, radiological knowledge is trained in order to make diagnoses more efficient. The book series ensures optimal preparation for the final medical examinations and is also a valuable tool for practical training. (orig.)

  5. 46 CFR 2.10-101 - Annual vessel inspection fee.

    Science.gov (United States)

    2010-10-01

    ... inspections, damage surveys, repair and modification inspections, change in vessel service inspections, permit... of international certificates. (d) Entitlement to inspection services for the current year remains... MODUs 4,695 Semi-submersible MODUs 8,050 Nautical School Vessels: Length not greater than 100 feet 835...

  6. Adsorption and revaporisation studies of thin iodine oxide and CsI aerosol deposits from containment surface materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Tietze, S.; Foreman, M.; Ekberg, C. [Chalmers Univ. of Technology, Goeteborg (Sweden); Kaerkelae, T.; Auvinen, A.; Tapper, U.; Jokiniemi, J. [VTT Technical Research Centre of Finland, Espoo (Finland)

    2013-07-15

    During a severe nuclear accident released fission and radiolysis products can react with each other to form new species which might contribute to the volatile source term. Iodine will be released from UO2 fuel mainly in form as CsI aerosol particles and elemental iodine. Elemental iodine can react in gaseous phase with ozone to form solid iodine oxide aerosol particles (IOx). Within the AIAS-2 (Adsorption of Iodine Aerosols on Surfaces) project the interactions of IOx and CsI aerosols with common containment surface materials was investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS. Non-radioactive and {sup 131}I labelled aerosols were produced from a KI solution and ozone with a new facility designed and built at VTT Technical Research Centre of Finland. CsI aerosols were produced from a CsI solution with the same facility. A monolayer of the aerosols was deposited on the surfaces. The deposits were analysed with microscopic and spectroscopic measurement techniques to identify the chemical form of the deposits on the surfaces to identify if a chemical conversion on the different surface materials had occured. The revaporisation behaviour of the deposited aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 with a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The decomposition effect of the radiolysis product carbon monoxide was tested on IOx aerosols deposited on a glass fibre filter. Iodine oxide particles were produced at 50 deg. C, 100 deg. C and 120 deg. C and deposited on filter samples in order to study the chemical

  7. Adsorption and revaporisation studies of thin iodine oxide and CsI aerosol deposits from containment surface materials in LWRs

    International Nuclear Information System (INIS)

    Tietze, S.; Foreman, M.; Ekberg, C.; Kaerkelae, T.; Auvinen, A.; Tapper, U.; Jokiniemi, J.

    2013-07-01

    During a severe nuclear accident released fission and radiolysis products can react with each other to form new species which might contribute to the volatile source term. Iodine will be released from UO2 fuel mainly in form as CsI aerosol particles and elemental iodine. Elemental iodine can react in gaseous phase with ozone to form solid iodine oxide aerosol particles (IOx). Within the AIAS-2 (Adsorption of Iodine Aerosols on Surfaces) project the interactions of IOx and CsI aerosols with common containment surface materials was investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS. Non-radioactive and 131 I labelled aerosols were produced from a KI solution and ozone with a new facility designed and built at VTT Technical Research Centre of Finland. CsI aerosols were produced from a CsI solution with the same facility. A monolayer of the aerosols was deposited on the surfaces. The deposits were analysed with microscopic and spectroscopic measurement techniques to identify the chemical form of the deposits on the surfaces to identify if a chemical conversion on the different surface materials had occured. The revaporisation behaviour of the deposited aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 with a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The decomposition effect of the radiolysis product carbon monoxide was tested on IOx aerosols deposited on a glass fibre filter. Iodine oxide particles were produced at 50 deg. C, 100 deg. C and 120 deg. C and deposited on filter samples in order to study the chemical speciation of

  8. Acrylic vessel cleaning tests

    International Nuclear Information System (INIS)

    Earle, D.; Hahn, R.L.; Boger, J.; Bonvin, E.

    1997-01-01

    The acrylic vessel as constructed is dirty. The dirt includes blue tape, Al tape, grease pencil, gemak, the glue or residue form these tapes, finger prints and dust of an unknown composition but probably mostly acrylic dust. This dirt has to be removed and once removed, the vessel has to be kept clean or at least to be easily cleanable at some future stage when access becomes much more difficult. The authors report on the results of a series of tests designed: (a) to prepare typical dirty samples of acrylic; (b) to remove dirt stuck to the acrylic surface; and (c) to measure the optical quality and Th concentration after cleaning. Specifications of the vessel call for very low levels of Th which could come from tape residues, the grease pencil, or other sources of dirt. This report does not address the concerns of how to keep the vessel clean after an initial cleaning and during the removal of the scaffolding. Alconox is recommended as the cleaner of choice. This acrylic vessel will be used in the Sudbury Neutrino Observatory

  9. Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

    Energy Technology Data Exchange (ETDEWEB)

    Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

    2012-09-01

    The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

  10. Evaluation of fatigue damage of pressure vessel materials by observation of microstructures

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1994-01-01

    As the important factor as the secular change mode of pressure vessel materials, there is fatigue damage. In USA, there is the move to use LWRs by extending their life, and it becomes necessary to show the soundness of the structures of machinery and equipment for long period. For exactly evaluating the soundness of the structures of machinery and equipment, it is important to clarify the degree of secular deterioration of the materials. In this report, by limiting to the fatigue damage of LWR pressure vessel steel, the method of grasping the change of microstructure and the method of estimating the degree of fatigue damage from the change of microstructure are shown. The change of microstructure arising in materials due to fatigue advances in the following steps, namely, the multiplication of dislocations, the tangling of dislocations, the formation of cell structure, the turning of cells, the formation of microcracks, the growth of cracks and fracture. In the case of pressure vessel steel, due to the quenching and tempering, the cell structure is formed from the beginning, and the advance of fatigue is recognized as the increase of the turning angle of cell structures. The detection of fatigue damage by microstructure is reported. (K.I.)

  11. Conformable pressure vessel for high pressure gas storage

    Science.gov (United States)

    Simmons, Kevin L.; Johnson, Kenneth I.; Lavender, Curt A.; Newhouse, Norman L.; Yeggy, Brian C.

    2016-01-12

    A non-cylindrical pressure vessel storage tank is disclosed. The storage tank includes an internal structure. The internal structure is coupled to at least one wall of the storage tank. The internal structure shapes and internally supports the storage tank. The pressure vessel storage tank has a conformability of about 0.8 to about 1.0. The internal structure can be, but is not limited to, a Schwarz-P structure, an egg-crate shaped structure, or carbon fiber ligament structure.

  12. Radioactive liquid containing vessel

    International Nuclear Information System (INIS)

    Sakurada, Tetsuo; Kawamura, Hironobu.

    1993-01-01

    Cooling jackets are coiled around the outer circumference of a container vessel, and the outer circumference thereof is covered with a surrounding plate. A liquid of good conductivity (for example, water) is filled between the cooling jackets and the surrounding plate. A radioactive liquid is supplied to the container vessel passing through a supply pipe and discharged passing through a discharge pipe. Cooling water at high pressure is passed through the cooling water jackets in order to remove the heat generated from the radioactive liquid. Since cooling water at high pressure is thus passed through the coiled pipes, the wall thickness of the container vessel and the cooling water jackets can be reduced, thereby enabling to reduce the cost. Further, even if the radioactive liquid is leaked, there is no worry of contaminating cooling water, to prevent contamination. (I.N.)

  13. ISO and EIGA standards for cryogenic vessels and accessories

    CERN Multimedia

    CERN. Geneva

    2016-01-01

    The EIGA/WG 6’s scope is cryogenic vessels and accessories, including their design, material compatibility, operational requirements and periodical inspection. The specific responsibilities include monitoring international standardization (ISO, CEN) and regulations (UN, TPED, PED...

  14. Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

    International Nuclear Information System (INIS)

    Park, J.W.; Bae, J.H.; Seol, W.C.

    2015-01-01

    An integrated conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-layered stratified corium (metallic pool, oxide pool and heavy metal) and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel. The 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

  15. TMI-2 Vessel Investigation Project integration report

    International Nuclear Information System (INIS)

    Wolf, J.R.; Rempe, J.L.; Stickler, L.A.; Korth, G.E.; Diercks, D.R.; Neimark, L.A.; Akers, D.W.; Schuetz, B.K.; Shearer, T.L.; Chavez, S.A.; Thinnes, G.L.; Witt, R.J.; Corradini, M.L.; Kos, J.A.

    1994-03-01

    The Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) was an international effort that was sponsored by the Nuclear Energy Agency of the Organization for Economic Cooperation and Development. The primary objectives of the VIP were to extract and examine samples from the lower head and to evaluate the potential modes of failure and the margin of structural integrity that remained in the TMI-2 reactor vessel during the accident. This report presents a summary of the major findings and conclusions that were developed from research during the VIP. Results from the various elements of the project are integrated to form a cohesive understanding of the vessel's condition after the accident

  16. Crashworthy sealed pressure vessel for plutonium transport

    International Nuclear Information System (INIS)

    Andersen, J.A.

    1980-01-01

    A rugged transportation package for the air shipment of radioisotopic materials was recently developed. This package includes a tough, sealed, stainless steel inner containment vessel of 1460 cc capacity. This vessel, intended for a mass load of up to 2 Kg PuO 2 in various isotopic forms (not to exceed 25 watts thermal activity), has a positive closure design consisting of a recessed, shouldered lid fastened to the vessel body by twelve stainless-steel bolts; sealing is accomplished by a ductile copper gasket in conjunction with knife-edge sealing beads on both the body and lid. Follow-on applications of this seal in newer, smaller packages for international air shipments of plutonium safeguards samples, and in newer, more optimized packages for greater payload and improved efficiency and utility, are briefly presented

  17. Pressure vessel integrity 1991

    International Nuclear Information System (INIS)

    Bhandari, S.; Doney, R.O.; McDonald, M.S.; Jones, D.P.; Wilson, W.K.; Pennell, W.E.

    1991-01-01

    This volume contains papers relating to the structural integrity assessment of pressure vessels and piping, with special emphasis on nuclear industry applications. The papers were prepared for technical sessions developed under the sponsorship of the ASME Pressure Vessels and Piping Division Committees for Codes and Standards, Computer Technology, Design and Analysis, and Materials Fabrication. They were presented at the 1991 Pressure Vessels and Piping Division Conference in San Diego, California, June 23-27. The primary objective of the sponsoring organization is to provide a forum for the dissemination and discussion of information on development and application of technology for the structural integrity assessment of pressure vessels and piping. This publication includes contributions from authors from Australia, France, Japan, Sweden, Switzerland, the United Kingdom, and the United States. The papers here are organized in six sections, each with a particular emphasis as indicated in the following section titles: Fracture Technology Status and Application Experience; Crack Initiation, Propagation and Arrest; Ductile Tearing; Constraint, Stress State, and Local-Brittle-Zones Effects; Computational Techniques for Fracture and Corrosion Fatigue; and Codes and Standards for Fatigue, Fracture and Erosion/Corrosion

  18. The reactor vessel steels

    International Nuclear Information System (INIS)

    Bilous, W.; Hajewska, E.; Szteke, W.; Przyborska, M.; Wasiak, J.; Wieczorkowski, M.

    2005-01-01

    In the paper the fundamental steels using in the construction of pressure vessel water reactor are discussed. The properties of these steels as well as the influence of neutron irradiation on its degradation in the time of exploitation are also done. (authors)

  19. Vacuum distilling vessel

    Energy Technology Data Exchange (ETDEWEB)

    Reik, H

    1928-12-27

    Vacuum distilling vessel for mineral oil and the like, characterized by the ring-form or polyconal stiffeners arranged inside, suitably eccentric to the casing, being held at a distance from the casing by connecting members of such a height that in the resulting space if necessary can be arranged vapor-distributing pipes and a complete removal of the residue is possible.

  20. Visualization of vessel traffic

    NARCIS (Netherlands)

    Willems, C.M.E.

    2011-01-01

    Moving objects are captured in multivariate trajectories, often large data with multiple attributes. We focus on vessel traffic as a source of such data. Patterns appearing from visually analyzing attributes are used to explain why certain movements have occurred. In this research, we have developed

  1. GOLD PRESSURE VESSEL SEAL

    Science.gov (United States)

    Smith, A.E.

    1963-11-26

    An improved seal between the piston and die member of a piston-cylinder type pressure vessel is presented. A layer of gold, of sufficient thickness to provide an interference fit between the piston and die member, is plated on the contacting surface of at least one of the members. (AEC)

  2. Reactor vessel stud tensioner

    International Nuclear Information System (INIS)

    Malandra, L.J.; Beer, R.W.; Salton, R.B.; Spiegelman, S.R.; Cognevich, M.L.

    1982-01-01

    A quick-acting stud tensioner, for facilitating the loosening or tightening of a stud nut on a reactor vessel stud, has gripper jaws which when the tensioner is lowered into engagement with the upper end of the stud are moved inwards to grip the upper end and which when the tensioner is lifted move outward to release the upper end. (author)

  3. TMI-2 Vessel Investigation Project Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-01-01

    The TMI-2 [Three Mile Island unit 2] Vessel Investigation Project Metallurgical Program at Argonne National Laboratory is a part of the international TMI-2 Vessel Investigation Project being conducted jointly by the U.S. Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which accounts for a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  4. Eddy current testing of composite pressure vessels

    Science.gov (United States)

    Casperson, R.; Pohl, R.; Munzke, D.; Becker, B.; Pelkner, M.

    2018-04-01

    The use of composite pressure vessels instead of conventional vessels made of steel or aluminum grew strongly over the last decade. The reason for this trend is the tremendous weight saving in the case of composite vessels. However, the long-time behavior is not fully understood for filling and discharging cycles and creep strength and their influence on the CFRP coating (carbon fiber reinforced plastics) and the internal liner (steel, aluminum, or plastics). The CFRP ensures the pressure resistance while the inner liner is used as a container for liquid or gas. To overcome the missing knowledge of aging, BAM started an internal project to investigate degradation of these material systems. Therefore, applicable testing methods like eddy current testing are needed. Normally, high-frequency eddy current testing (HF-ET, f > 10 MHz) is deployed for CFRP due to its low conductivity of the fiber, which is in the order of 0.01 MS/s, and the capacitive coupling between the fibers. Nevertheless, in some cases conventional ET can be applied. We show a concise summary of studies on the application of conventional ET of composite pressure vessels.

  5. Residual Stress Estimation and Fatigue Life Prediction of an Autofrettaged Pressure Vessel

    Energy Technology Data Exchange (ETDEWEB)

    Song, Kyung Jin; Kim, Eun Kyum; Koh, Seung Kee [Kunsan Nat’l Univ., Kunsan (Korea, Republic of)

    2017-09-15

    Fatigue failure of an autofrettaged pressure vessel with a groove at the outside surface occurs owing to the fatigue crack initiation and propagation at the groove root. In order to predict the fatigue life of the autofrettaged pressure vessel, residual stresses in the autofrettaged pressure vessel were evaluated using the finite element method, and the fatigue properties of the pressure vessel steel were obtained from the fatigue tests. Fatigue life of a pressure vessel obtained through summation of the crack initiation and propagation lives was calculated to be 2,598 cycles for an 80% autofrettaged pressure vessel subjected to a pulsating internal pressure of 424 MPa.

  6. Vessels in Transit - Web Tool

    Data.gov (United States)

    Department of Transportation — A web tool that provides real-time information on vessels transiting the Saint Lawrence Seaway. Visitors may sort by order of turn, vessel name, or last location in...

  7. TPX vacuum vessel transient thermal and stress conditions

    International Nuclear Information System (INIS)

    Feldshteyn, Y.; Dinkevich, S.; Feng, T.; Majumder, D.

    1995-01-01

    The TPX vacuum vessel provides the vacuum boundary for the plasma and the mechanical support for the internal components. Another function of the vacuum vessel is to contain neutron shielding water in the double wall space during normal operation. This double wall space serves as a heat reservoir for the entire vacuum vessel during bakeout. The vacuum vessel and the internal components are subjected to thermal stresses induced by a nonuniform temperature distribution within the structure during bakeout. A successful Conceptual Design Review in March 1993 has established superheated steam as the heating source of the vacuum vessel. A transient bakeout mode of the vacuum vessel and in-vessel components has been analyzed to evaluate transient period duration, proper temperature level, actual thermal stresses and performance of the steam equipment. Thermally, the vacuum vessel structure may be considered as an adiabatic system because it is perfectly insulated by the strong surrounding vacuum and multiple layers of superinsulation. Important aspects of the analysis are described herein

  8. In service inspection of SUPERPHENIX 1 vessels: MIR

    International Nuclear Information System (INIS)

    Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

    1985-01-01

    Although no in-service inspection constraints were imposed on the Phenix vessels, the Safety Authorities asked that the design of SUPERPHENIX 1 makes it possible to monitor throughout the lifetime of the reactor, surface and internal defects on the main vessel. A pool design and the presence of heat baffles inside the main vessel make access from the inside of the vessel impossible. Thus, an inspection can only be performed from the outside of the main vessel: the distance between the walls of the main and safety vessels is such that an inspection device can be introduced into the corresponding space. As the design of the reactor precludes radiographic inspection, the method which was selected for monitoring internal defects in the main vessel is ultrasonics. However, the anisotropic structure of austenitic stainless steel welds limits the performance of this technique. The authors present the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of SUPERPHENIX 1 vessels

  9. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    An apparatus is described for sealing a cold leg nozzle of a nuclear reactor pressure vessel from a remote location comprising: at least one sealing plug for mechanically sealing the nozzle from the inside of the reactor pressure vessel. The sealing plug includes a plate and a cone assembly having an end part receptive in the nozzle, the plate being axially moveable relative to the cone assembly. The plate and cone assembly have confronting bevelled edges defining an opening therebetween. A primary O-ring is disposed about the opening and is supported on the bevelled edges, the plate being guidably mounted to the cone assembly for movement toward the cone assembly to radially expand the primary O-ring into sealing engagement with the nozzle. A means is included for providing relative movement between the outer plate and the cone assembly

  10. Nuclear reactor vessel inspection apparatus

    International Nuclear Information System (INIS)

    Blackstone, E.G.; Lofy, R.A.; Williams, L.P.

    1979-01-01

    Apparatus for the in situ inspection of a nuclear reactor vessel to detect the location and character of flaws in the walls of the vessel, in the welds joining the various sections of the vessel, in the welds joining attachments such as nozzles, elbows and the like to the reactor vessel and in such attachments wherein an inspection head carrying one or more ultrasonic transducers follows predetermined paths in scanning the various reactor sections, welds and attachments

  11. Reactor vessel stud closure system

    International Nuclear Information System (INIS)

    Spiegelman, S.R.; Salton, R.B.; Beer, R.W.; Malandra, L.J.; Cognevich, M.L.

    1982-01-01

    A quick-acting stud tensioner apparatus for enabling the loosening or tightening of a stud nut on a reactor vessel stud. The apparatus is adapted to engage the vessel stud by closing a gripper around an upper end of the vessel stud when the apparatus is seated on the stud. Upon lifting the apparatus, the gripper releases the vessel stud so that the apparatus can be removed

  12. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

    Van De Velde, J.; Fabry, A.; Van Walle, E.; Chaouuadi, R.

    1998-01-01

    Research and development activities related to reactor pressure vessel steels during 1997 are reported. The objectives of activities of the Belgian Nuclear Research Centre SCK/CEN in this domain are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate a methodology on a broad database; (3) to achieve regulatory acceptance and industrial use

  13. Reactor Pressure Vessel Steels

    Energy Technology Data Exchange (ETDEWEB)

    Van de Velde, J.; Fabry, A.; Van Walle, E.; Chaoudi, R

    1998-07-01

    SCK-CEN's R and D programme on Reactor Pressure Vessel (RPV) Steels in performed in support of the RVP integrity assessment. Its main objectives are: (1) to develop enhanced surveillance concepts by applying micromechanics and fracture-toughness tests to small specimens, and by performing damage modelling and microstructure characterization; (2) to demonstrate the applied methodology on a broad database; (3) to achieve regulatory acceptance and industrial use. Progress and achievements in 1999 are reported.

  14. Vessel generator noise as a settlement cue for marine biofouling species.

    Science.gov (United States)

    McDonald, J I; Wilkens, S L; Stanley, J A; Jeffs, A G

    2014-01-01

    Underwater noise is increasing globally, largely due to increased vessel numbers and international ocean trade. Vessels are also a major vector for translocation of non-indigenous marine species which can have serious implications for biosecurity. The possibility that underwater noise from fishing vessels may promote settlement of biofouling on hulls was investigated for the ascidian Ciona intestinalis. Spatial differences in biofouling appear to be correlated with spatial differences in the intensity and frequency of the noise emitted by the vessel's generator. This correlation was confirmed in laboratory experiments where C. intestinalis larvae showed significantly faster settlement and metamorphosis when exposed to the underwater noise produced by the vessel generator. Larval survival rates were also significantly higher in treatments exposed to vessel generator noise. Enhanced settlement attributable to vessel generator noise may indicate that vessels not only provide a suitable fouling substratum, but vessels running generators may be attracting larvae and enhancing their survival and growth.

  15. Phenomenological vessel burst investigations

    International Nuclear Information System (INIS)

    Hippelein, K.W.; Julisch, P.; Muz, J.; Schiedermaier, J.

    1985-07-01

    Fourteen burst experiments have been carried out using vessels with circumferential and longitudinal flaws, for investigation of the fracture behaviour, i.e. the time-related fracture opening. The vessels had dimensions (outer diameter x wall thickness = 800 x 47 mm) which correspond to the dimensions of the main coolant piping of a 1300 MW e PWR. The test specimens had been made of the base-safe material 20 MnMoNi 55 and of a special, 22 NiMoCr 37 base alloy. The experimental conditions with regard to pressure and temperature have been chosen so as to correspond to normal operating conditions of a PWR (p∝17.5 MPa, T∝300 0 C), i.e. the flaws have been so dimensioned that failure was to be expected at a pressure of p∝17.5 MPa. As a rule, water has been used as the pressure medium, or in some cases air, in order to influence the time-dependent pressure decrease. Fluid and structural dynamics calculations have also been made. In order to determine the impact of a fast propagating crack on the leak-to-fracture curve, which normally is defined by quasistationary experiments, suitable tests have been made with large-volume, cylindrical vessels (outer diameter x wall thickness x length = 3000 x 21 x 14000 mm) made of the material WSt E 43. The leak-before-fracture criterion has been confirmed. (orig./HP) [de

  16. Blood Vessels in Allotransplantation.

    Science.gov (United States)

    Abrahimi, P; Liu, R; Pober, J S

    2015-07-01

    Human vascularized allografts are perfused through blood vessels composed of cells (endothelium, pericytes, and smooth muscle cells) that remain largely of graft origin and are thus subject to host alloimmune responses. Graft vessels must be healthy to maintain homeostatic functions including control of perfusion, maintenance of permselectivity, prevention of thrombosis, and participation in immune surveillance. Vascular cell injury can cause dysfunction that interferes with these processes. Graft vascular cells can be activated by mediators of innate and adaptive immunity to participate in graft inflammation contributing to both ischemia/reperfusion injury and allograft rejection. Different forms of rejection may affect graft vessels in different ways, ranging from thrombosis and neutrophilic inflammation in hyperacute rejection, to endothelialitis/intimal arteritis and fibrinoid necrosis in acute cell-mediated or antibody-mediated rejection, respectively, and to diffuse luminal stenosis in chronic rejection. While some current therapies targeting the host immune system do affect graft vascular cells, direct targeting of the graft vasculature may create new opportunities for preventing allograft injury and loss. © Copyright 2015 The American Society of Transplantation and the American Society of Transplant Surgeons.

  17. Ionizing radiations and blood vessels

    International Nuclear Information System (INIS)

    Vorob'ev, E.I.; Stepanov, R.P.

    1985-01-01

    Data on phenomenology of radiation-induced changes in blood vessels are systematized and authors' experience is generalized. Modern concepts about processes leading to vessel structure injury after irradiation is critically analyzed. Special attention is paid to reparation and compensation of X-ray vessel injury, consideration of which is not yet sufficiently elucidated in literature

  18. Ionizing radiations and blood vessels

    International Nuclear Information System (INIS)

    Vorob'ev, E.I.; Stepanov, R.P.

    1985-01-01

    Data on phenomeology of radiation changes of blood vessels are systemized and the authors' experience is generalyzed. A critical analysis of modern conceptions on processes resulting in vessel structure damage after irradiation, is given. Special attention is paid to reparation and compensation of radiation injury of vessels

  19. Burning of MOX fuels in LWRs; fuel history effects on thermal properties of hull and end piece wastes and the repository performance

    International Nuclear Information System (INIS)

    Hirano, Fumio; Sato, Seichi; Kozaki, Tamotsu

    2012-01-01

    The thermal impacts of hull and end piece wastes from the reprocessing of MOX spent fuels burned in LWRs on repository performance were investigated. The heat generation rates in MOX spent fuels and the resulting heat generation rates in hull and end piece wastes change depending on the history of MOX fuels. This history includes the burn-up of UO 2 spent fuels from which the Pu is obtained, the cooling period before reprocessing, the storage period of fresh MOX fuels before being loaded into an LWR, as well as the burn-up of the MOX fuels. The heat generation rates in hull and end piece wastes from the reprocessing of MOX spent fuels with any of those histories are significantly larger than those from UO 2 spent fuels with burn-ups of 45 GWd/THM. If a temperature below 80degC is specified for cement-based materials used in waste packages after disposal, the allowable number of canisters containing compacted hull and end pieces in a package for 45 and 70 GWd-MOX needs to be limited to a value of 0.4-1.6, which is significantly lower than 4.0 for 45 GWd-UO 2 . (author)

  20. Reliability analysis of reactor pressure vessel intensity

    International Nuclear Information System (INIS)

    Zheng Liangang; Lu Yongbo

    2012-01-01

    This paper performs the reliability analysis of reactor pressure vessel (RPV) with ANSYS. The analysis method include direct Monte Carlo Simulation method, Latin Hypercube Sampling, central composite design and Box-Behnken Matrix design. The RPV integrity reliability under given input condition is proposed. The result shows that the effects on the RPV base material reliability are internal press, allowable basic stress and elasticity modulus of base material in descending order, and the effects on the bolt reliability are allowable basic stress of bolt material, preload of bolt and internal press in descending order. (authors)

  1. Use of miniature and standard specimens to evaluate effects of irradiation temperature on pressure vessel steels

    International Nuclear Information System (INIS)

    Haggag, F.M.; Nanstad, R.K.; Byrne, S.T.

    1991-01-01

    The effects of neutron irradiation on the steel reactor vessel for the modular high-temperature gas-cooled reactor (MHTGR) are being investigated, primarily because the operating temperatures are low [121 to 210 degrees C (250--410 degrees F)] compared to those for commercial light-water reactors (LWRs) [∼288 degrees C (550 degrees F)]. The need for design data on the reference temperature shift necessitated the irradiation at different temperatures of A 533 grade B class 1 plate. A 508 class 3 forging, and welds used for the vessel shell, vessel closure head, the vessel flange. This paper presents results from the first four irradiation capsules of this program. The four capsules were irradiated in the University of Buffalo Reactor to an effective fast fluence of 1 x10 18 neutron/cm 2 [0.68 x 10 18 neutron/cm 2 (>1 MeV)] at temperatures of 288, 204, 163, and 121 degrees C (550, 400, 325, and 250 degrees F), respectively. The yield and ultimate strengths of both steel plate materials of the MHTGR Program increased with decreasing irradiation temperature. Similarly, the 41-J Charpy V-notch (CVN) transition temperature shift increased with decreasing irradiation temperature (in agreement with the increase in yield strength). The miniature tensile and automated ball indentation (ABI) test results (yield strength and flow properties) were in good agreement with those from standard tensile specimens. The miniature tensile and ABI test results were also used in a model that utilizes the changes in yield strength to estimate the CVN ductile-to-brittle transition temperature shift due to irradiation. The model predictions were compared with CVN test results obtained here and in earlier work. 5 refs., 11 figs., 6 tabs

  2. Pressure vessel lid

    International Nuclear Information System (INIS)

    Schoening, J.; Elter, C.; Becker, G.; Pertiller, S.

    1986-01-01

    The invention concerns a lid for closing openings in reactor pressure vessels containing helium, which is made as a circular casting with hollow spaces and a flat floor and is set on the opening and kept down. It consists of helium-tight metal cast material with sufficient temperature resistance. There are at least two concentric heat resistant seals let into the bottom of the lid. The bottom is in immediate contact with the container atmosphere and has hollow spaces in its inside in the area opposite to the opening. (orig./HP) [de

  3. [Small vessel cerebrovascular disease].

    Science.gov (United States)

    Cardona Portela, P; Escrig Avellaneda, A

    2018-05-09

    Small vessel vascular disease is a spectrum of different conditions that includes lacunar infarction, alteration of deep white matter, or microbleeds. Hypertension is the main risk factor, although the atherothrombotic lesion may be present, particularly in large-sized lacunar infarctions along with other vascular risk factors. MRI findings are characteristic and the lesions authentic biomarkers that allow differentiating the value of risk factors and defining their prognostic value. Copyright © 2018 SEH-LELHA. Publicado por Elsevier España, S.L.U. All rights reserved.

  4. Stress criteria for nuclear vessel concrete

    International Nuclear Information System (INIS)

    Costes, D.

    1975-01-01

    Concrete nuclear vessels are submitted to prestressing forces which limit tensile stresses in concrete when the vessel is under pressure with thermal gradients. Hence, the most severe conditions for concrete appear when the vessel is prestressed and not submitted to internal pressure. The triaxial states of stress in the concrete may be computed postulating elastic or other behavior and compared with safe limits obtained from rupture tests and fatigue tests. The first part of the paper, recalls experimental rupture results and the acceptability procedures currently used. Criteria founded on the lemniscoid surfaces are proposed, parameters for which are obtained by various tests and safety considerations. In the second part, rupture tests are reported on small, thick, cylindrical vessels submitted to external hydraulic pressure simulating prestressing forces. Materials used are plain concrete, microconcrete, marble and graphite. The strengths obtained are much higher than those which could be elastically computed, triaxial rupture states being provided by previous experiments. Such results may be due to a plastic stress redistribution before fracture and to stabilizing effects of stress gradients around the more stressed areas. Fatigue tests by external hydraulic loading are reported [fr

  5. PWR pressure vessel integrity during overcooling accidents

    International Nuclear Information System (INIS)

    Cheverton, R.D.

    1981-01-01

    Pressurized water reactors are susceptible to certain types of hypothetical accidents that under some circumstances, including operation of the reactor beyond a critical time in its life, could result in failure of the pressure vessel as a result of propagation of crack-like defects in the vessel wall. The accidents of concern are those that result in thermal shock to the vessel while the vessel is subjected to internal pressure. Such accidents, referred to as pressurized thermal shock or overcooling accidents (OCA), include a steamline break, small-break LOCA, turbine trip followed by stuck-open bypass valves, the 1978 Rancho Seco and the TMI accidents and many other postulated and actual accidents. The source of cold water for the thermal shock is either emergency core coolant or the normal primary-system coolant. ORNL performed fracture-mechanics calculations for a steamline break in 1978 and for a turbine-trip case in 1980 and concluded on the basis of the results that many more such calculations would be required. To meet the expected demand in a realistic way a computer code, OCA-I, was developed that accepts primary-system temperature and pressure transients as input and then performs one-dimensional thermal and stress analyses for the wall and a corresponding fracture-mechanics analysis for a long axial flaw. The code is briefly described, and its use in both generic and specific plant analyses is discussed

  6. An automated vessel segmentation of retinal images using multiscale vesselness

    International Nuclear Information System (INIS)

    Ben Abdallah, M.; Malek, J.; Tourki, R.; Krissian, K.

    2011-01-01

    The ocular fundus image can provide information on pathological changes caused by local ocular diseases and early signs of certain systemic diseases, such as diabetes and hypertension. Automated analysis and interpretation of fundus images has become a necessary and important diagnostic procedure in ophthalmology. The extraction of blood vessels from retinal images is an important and challenging task in medical analysis and diagnosis. In this paper, we introduce an implementation of the anisotropic diffusion which allows reducing the noise and better preserving small structures like vessels in 2D images. A vessel detection filter, based on a multi-scale vesselness function, is then applied to enhance vascular structures.

  7. Design characteristics of pantograph type in vessel fuel handling system in SFR

    International Nuclear Information System (INIS)

    Kim, S. H.; Koo, G. H.

    2012-01-01

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied

  8. Design characteristics of pantograph type in vessel fuel handling system in SFR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. H.; Koo, G. H. [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    The pantograph type in vessel fuel handling system in a sodium cooled fast reactor (SFR), which requires installation space for the slot in the upper internal structure attached under the rotating plug, is composed of an in vessel transfer machine (IVTM), a single rotating plug, in vessel storage, and a fuel transfer port (FTP). The pantograph type IVTM can exchange fuel assemblies through a slot, the design requirement of which should be essentially considered in the design of the in vessel fuel handling system. In addition, the spent fuel assemblies temporarily stored in the in vessel storage of the reactor vessel are removed to the outside of the reactor vessel through the FTP. The fuel transfer basket is then provided in the FTP, and a fuel transfer is performed by using it. In this study, the design characteristics for a pantograph type in vessel fuel handling system are reviewed, and the preconceptual designs are studied.

  9. Targeting Therapy Resistant Tumor Vessels

    Science.gov (United States)

    2008-08-01

    Morris LS. Hysterectomy vs. resectoscopic endometrial ablation for the control of abnormal uterine bleeding . A cost-comparative study. J Reprod Med 1994;39...after the antibody treatment contain a pericyte coat, vessel architecture is normal, the diameter of the vessels is smaller (dilated, abnormal vessels...involvement of proteases from inflammatory mast cells and functionally abnormal (Carmeliet and Jain, 2000; Pasqualini (Coussens et al., 1999) and other bone

  10. The vessel fluence; Fluence cuve

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This book presents the proceedings of the technical meeting on the reactors vessels fluence. They are grouped in eight sessions: the industrial context and the stakes of the vessels control; the organization and the methodology for the fluence computation; the concerned physical properties; the reference computation methods; the fluence monitoring in an industrial context; vessels monitoring under irradiation; others methods in the world; the research and development programs. (A.L.B.)

  11. Americium behaviour in plastic vessels

    International Nuclear Information System (INIS)

    Legarda, F.; Herranz, M.; Idoeta, R.; Abelairas, A.

    2010-01-01

    The adsorption of 241 Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of 241 Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of 241 Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

  12. Americium behaviour in plastic vessels

    Energy Technology Data Exchange (ETDEWEB)

    Legarda, F.; Herranz, M. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Idoeta, R., E-mail: raquel.idoeta@ehu.e [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain); Abelairas, A. [Departamento de Ingenieria Nuclear y Mecanica de Fluidos, Escuela Tecnica Superior de Ingenieria de Bilbao, Universidad del Pais Vasco (UPV/EHU), Alameda de Urquijo s/n, 48013 Bilbao (Spain)

    2010-07-15

    The adsorption of {sup 241}Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of {sup 241}Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of {sup 241}Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification.

  13. Americium behaviour in plastic vessels.

    Science.gov (United States)

    Legarda, F; Herranz, M; Idoeta, R; Abelairas, A

    2010-01-01

    The adsorption of (241)Am dissolved in water in different plastic storage vessels was determined. Three different plastics were investigated with natural and distilled waters and the retention of (241)Am by these plastics was studied. The same was done by varying vessel agitation time, vessel agitation speed, surface/volume ratio of water in the vessels and water pH. Adsorptions were measured to be between 0% and 70%. The adsorption of (241)Am is minimized with no water agitation, with PET or PVC plastics, and by water acidification. Copyright 2009 Elsevier Ltd. All rights reserved.

  14. [Large vessel vasculitides].

    Science.gov (United States)

    Morović-Vergles, Jadranka; Puksić, Silva; Gracanin, Ana Gudelj

    2013-01-01

    Large vessel vasculitis includes Giant cell arteritis and Takayasu arteritis. Giant cell arteritis is the most common form of vasculitis affect patients aged 50 years or over. The diagnosis should be considered in older patients who present with new onset of headache, visual disturbance, polymyalgia rheumatica and/or fever unknown cause. Glucocorticoides remain the cornerstone of therapy. Takayasu arteritis is a chronic panarteritis of the aorta ant its major branches presenting commonly in young ages. Although all large arteries can be affected, the aorta, subclavian and carotid arteries are most commonly involved. The most common symptoms included upper extremity claudication, hypertension, pain over the carotid arteries (carotidynia), dizziness and visual disturbances. Early diagnosis and treatment has improved the outcome in patients with TA.

  15. Reactor pressure vessel embrittlement

    International Nuclear Information System (INIS)

    1992-07-01

    Within the framework of the IAEA extrabudgetary programme on the Safety of WWER-440/230 NPPs, a list of safety issues requiring broad studies of generic interest have been agreed upon by an Advisory Group who met in Vienna in September 1990. The list was later revised in the light of the programme findings. The information on the status of the issues, and on the amount of work already completed and under way in the various countries, needs to be compiled. Moreover, an evaluation of what further work is required to resolve each one of the issues is also necessary. In view of this, the IAEA has started the preparation of a series of status reports on the various issues. This report on the generic safety issue ''Reactor Pressure Vessel Embrittlement'' presents a comprehensive survey of technical information available in the field and identifies those aspects which require further investigation. 39 refs, 21 figs, 4 tabs

  16. Reactor containment vessel

    International Nuclear Information System (INIS)

    Ochiai, Kanehiro; Hayagumo, Sunao; Morikawa, Matsuo.

    1981-01-01

    Purpose: To safety and simplify the structure in a reactor containment vessel. Constitution: Steam flow channels with steam jetting ports communicating to coolants are provided between a communication channel and coolants in a pressure suppression chamber. Upon loss of coolant accidents, pressure in a dry well will increase, then force downwards water in an annulus portion and further flow out the water through steam jetting ports into a suppression pool. Thus, the steam flow channel is filled with steams or airs present in the dry well, which are released through the steam jetting ports into the pressure suppression chamber. Even though water is violently vibrated owing to the upward movement of air bubbles and condensation of steam bubbles, the annular portion and the steam jetting ports are filled with steams or the like, direct dynamic loads onto the structures such as communication channels can be avoided. (J.P.N.)

  17. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1984-12-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training the head was safely removed and stored and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  18. TMI-2 reactor vessel head removal

    International Nuclear Information System (INIS)

    Bengel, P.R.; Smith, M.D.; Estabrook, G.A.

    1985-09-01

    This report describes the safe removal and storage of the Three Mile Island Unit 2 (TMI-2) reactor vessel head. The head was removed in July 1984 to permit the removal of the plenum and the reactor core, which were damaged during the 1979 accident. From July 1982, plans and preparations were made using a standard head removal procedure modified by the necessary precautions and changes to account for conditions caused by the accident. After data acquisition, equipment and structure modifications, and training, the head was safely removed and stored; and the internals indexing fixture and a work platform were installed on top of the vessel. Dose rates during and after the operation were lower than expected; lessons were learned from the operation which will be applied to the continuing fuel removal operations activities

  19. Autonomous radiation monitoring of small vessels

    International Nuclear Information System (INIS)

    Ziock, K.P.; Cheriyadat, A.; Fabris, L.; Goddard, J.; Hornback, D.; Karnowski, T.; Kerekes, R.; Newby, J.

    2011-01-01

    Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the land-based approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. In contrast to roadways, where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. A unique solution to this problem is based on a portal-less portal monitor designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. This paper presents the results of a recent test of the system in a maritime setting.

  20. Irradiation embrittlement of pressure vessel steels

    International Nuclear Information System (INIS)

    Brumovsky, M.; Vacek, M.

    1975-01-01

    A Standard Research Programme on Irradiation Embrittlement of Pressure Vessel Steels was approved by the Coordinating Meeting on the 12th May 1972 at the Working Group on Engineering Aspects of Irradiation Embrittlement of Pressure Vessel Steels. This Working Group was set up by the International Atomic Energy Agency in Vienna. Seven countries with their research institutes agreed on doing irradiation experiments according to the approved programme on steel A533 B from the U.S. HSST Programme. The Czechoslovak contribution covering tensile and impact testing of non-irradiated steel and steel irradiated at 280degC to 1.3 x 10 23 n/m 2 (E above 1 MeV) is presented in this report. As an additional part the same set of experiments was carried out on two additional steels - A 542 and A 543, made in SKODA Works for comparison of their irradiation embrittlement and hardening with A533 B steel. (author)

  1. Advanced Instrumentation, Information and Control (II&C) Research and Development Facility Buildout and Project Execution of LWRS II&C Pilot Projects 1 and 3

    Energy Technology Data Exchange (ETDEWEB)

    Ronald Farris; Johanna Oxstrand; Gregory Weatherby

    2011-09-01

    The U.S. Department of Energy (DOE) is sponsoring research, development, and deployment on light water reactor sustainability (LWRS), in which the Idaho National Laboratory (INL) is working closely with nuclear utilities to develop technologies and solutions to help ensure the safe operational life extension of current reactors. As technologies are introduced that change the operation of the plant, the LWRS pilot projects can help identify their best-advanced uses and help demonstrate the safety of these technologies. In early testing of operator performance given these emerging technologies will ensure the safety and usability of systems prior to large-scale deployment and costly verification and validation at the plant. The aim of these collaborations, demonstrations, and approaches are intended to lessen the inertia that sustains the current status quo of today's II&C systems technology, and to motivate transformational change and a shift in strategy to a long-term approach to II&C modernization that is more sustainable. Research being conducted under Pilot Project 1 regards understanding the conditions and behaviors that can be modified, either through process improvements and/or technology deployment, to improve the overall safety and efficiency of outage control at nuclear facilities. The key component of the research in this pilot project is accessing the delivery of information that will allow researchers to simulate the control room, outage control center (OCC) information, and plant status data. The simulation also allows researchers to identify areas of opportunity where plant operating status and outage activities can be analyzed to increase overall plant efficiency. For Pilot Project 3 the desire is to demonstrate the ability of technology deployment and the subsequent impact on maximizing the 'Collective Situational Awareness' of the various stakeholders in a commercial nuclear power plant. Specifically, the desire is to show positive

  2. Advanced Instrumentation, Information and Control (II and C) Research and Development Facility Buildout and Project Execution of LWRS II and C Pilot Projects 1 and 3

    International Nuclear Information System (INIS)

    Farris, Ronald; Oxstrand, Johanna; Weatherby, Gregory

    2011-01-01

    The U.S. Department of Energy (DOE) is sponsoring research, development, and deployment on light water reactor sustainability (LWRS), in which the Idaho National Laboratory (INL) is working closely with nuclear utilities to develop technologies and solutions to help ensure the safe operational life extension of current reactors. As technologies are introduced that change the operation of the plant, the LWRS pilot projects can help identify their best-advanced uses and help demonstrate the safety of these technologies. In early testing of operator performance given these emerging technologies will ensure the safety and usability of systems prior to large-scale deployment and costly verification and validation at the plant. The aim of these collaborations, demonstrations, and approaches are intended to lessen the inertia that sustains the current status quo of today's II and C systems technology, and to motivate transformational change and a shift in strategy to a long-term approach to II and C modernization that is more sustainable. Research being conducted under Pilot Project 1 regards understanding the conditions and behaviors that can be modified, either through process improvements and/or technology deployment, to improve the overall safety and efficiency of outage control at nuclear facilities. The key component of the research in this pilot project is accessing the delivery of information that will allow researchers to simulate the control room, outage control center (OCC) information, and plant status data. The simulation also allows researchers to identify areas of opportunity where plant operating status and outage activities can be analyzed to increase overall plant efficiency. For Pilot Project 3 the desire is to demonstrate the ability of technology deployment and the subsequent impact on maximizing the 'Collective Situational Awareness' of the various stakeholders in a commercial nuclear power plant. Specifically, the desire is to show positive results

  3. Recipient vessel selection in immediate breast reconstruction with free abdominal tissue transfer after nipple-sparing mastectomy.

    Science.gov (United States)

    Yang, Sung Jun; Eom, Jin Sup; Lee, Taik Jong; Ahn, Sei Hyun; Son, Byung Ho

    2012-05-01

    Nipple-sparing mastectomy (NSM) is gaining popularity due to its superior aesthetic results. When reconstructing the breast with free abdominal tissue transfer, we must readdress the recipient vessel, because NSM can cause difficulty in access to the chest vessel. Between June 2006 and March 2011, a total of 92 women underwent NSM with free abdominal tissue transfer. A lateral oblique incision was used for the nipple-sparing mastectomy. For recipient vessels, the internal mammary vessels were chosen if the mastectomy flap did not block access to the vessels. If it did, the thoracodorsal vessels were used. Age, degree of breast ptosis, weight of the mastectomy specimen, and related complications of the internal mammary vessel group and the thoracodorsal vessel group were compared. Thoracodorsal vessels were used as recipient vessels in 59 cases, and internal mammary vessels in 33 cases including 4 cases with perforators of the internal mammary vessels. Breast reconstruction was successful in all cases except one case involving a total flap failure, which was replaced by a silicone gel implant. The internal mammary group and the thoracodorsal group were similar in terms of age, height, breast weight, and degree of ptosis. The flap related complications such as flap loss and take-back operation rates were not significantly different between the two groups. The rate of nipple necrosis was higher in the internal mammary group. The thoracodorsal vessels could produce comparable outcomes in breast reconstruction after nipple-sparing mastectomies. If access to internal mammary vessels is difficult, the thoracodorsal vessel can be a better choice.

  4. Recipient Vessel Selection in Immediate Breast Reconstruction with Free Abdominal Tissue Transfer after Nipple-Sparing Mastectomy

    Directory of Open Access Journals (Sweden)

    Sung Jun Yang

    2012-05-01

    Full Text Available BackgroundNipple-sparing mastectomy (NSM is gaining popularity due to its superior aesthetic results. When reconstructing the breast with free abdominal tissue transfer, we must readdress the recipient vessel, because NSM can cause difficulty in access to the chest vessel.MethodsBetween June 2006 and March 2011, a total of 92 women underwent NSM with free abdominal tissue transfer. A lateral oblique incision was used for the nipple-sparing mastectomy. For recipient vessels, the internal mammary vessels were chosen if the mastectomy flap did not block access to the vessels. If it did, the thoracodorsal vessels were used. Age, degree of breast ptosis, weight of the mastectomy specimen, and related complications of the internal mammary vessel group and the thoracodorsal vessel group were compared.ResultsThoracodorsal vessels were used as recipient vessels in 59 cases, and internal mammary vessels in 33 cases including 4 cases with perforators of the internal mammary vessels. Breast reconstruction was successful in all cases except one case involving a total flap failure, which was replaced by a silicone gel implant. The internal mammary group and the thoracodorsal group were similar in terms of age, height, breast weight, and degree of ptosis. The flap related complications such as flap loss and take-back operation rates were not significantly different between the two groups. The rate of nipple necrosis was higher in the internal mammary group.ConclusionsThe thoracodorsal vessels could produce comparable outcomes in breast reconstruction after nipple-sparing mastectomies. If access to internal mammary vessels is difficult, the thoracodorsal vessel can be a better choice.

  5. Pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    1975-01-01

    The invention applies to a pressure vessel for nuclear reactors whose shell, made of cast metal segments, has a steel liner. This liner must be constructed to withstand all operational stresses and to be easily repairable. The invention solves this problem by installing the liner at a certain distance from the inner wall of the pressure vessel shell and by filling this clearance with supporting concrete. Both the concrete and the steel liner must have a lower prestress than the pressure vessel shell. In order to avoid damage to the liner when prestressing the pressure vessel shell, special connecting elements are provided which consist of welded-on fastening elements projecting into recesses in the cast metal segments of the pressure vessel. Their design is described in detail. (TK) [de

  6. Containment vessel stability analysis

    International Nuclear Information System (INIS)

    Harstead, G.A.; Morris, N.F.; Unsal, A.I.

    1983-01-01

    The stability analysis for a steel containment shell is presented herein. The containment is a freestanding shell consisting of a vertical cylinder with a hemispherical dome. It is stiffened by large ring stiffeners and relatively small longitudinal stiffeners. The containment vessel is subjected to both static and dynamic loads which can cause buckling. These loads must be combined prior to their use in a stability analysis. The buckling loads were computed with the aid of the ASME Code case N-284 used in conjunction with general purpose computer codes and in-house programs. The equations contained in the Code case were used to compute the knockdown factors due to shell imperfections. After these knockdown factors were applied to the critical stress states determined by freezing the maximum dynamic stresses and combining them with other static stresses, a linear bifurcation analysis was carried out with the aid of the BOSOR4 program. Since the containment shell contained large penetrations, the Code case had to be supplemented by a local buckling analysis of the shell area surrounding the largest penetration. This analysis was carried out with the aid of the NASTRAN program. Although the factor of safety against buckling obtained in this analysis was satisfactory, it is claimed that the use of the Code case knockdown factors are unduly conservative when applied to the analysis of buckling around penetrations. (orig.)

  7. Pressure vessel design

    International Nuclear Information System (INIS)

    Annaratone, D.

    2007-01-01

    This book guides through general and fundamental problems of pressure vessel design. It moreover considers problems which seem to be of lower importance but which turn out to be crucial in the design phase. The basic approach is rigorously scientific with a complete theoretical development of the topics treated, but the analysis is always pushed so far as to offer concrete and precise calculation criteria that can be immediately applied to actual designs. This is accomplished through appropriate algorithms that lead to final equations or to characteristic parameters defined through mathematical equations. The first chapter describes how to achieve verification criteria, the second analyzes a few general problems, such as stresses of the membrane in revolution solids and edge effects. The third chapter deals with cylinders under pressure from the inside, while the fourth focuses on cylinders under pressure from the outside. The fifth chapter covers spheres, and the sixth is about all types of heads. Chapter seven discusses different components of particular shape as well as pipes, with special attention to flanges. The eighth chapter discusses the influence of holes, while the ninth is devoted to the influence of supports. Finally, chapter ten illustrates the fundamental criteria regarding fatigue analysis. Besides the unique approach to the entire work, original contributions can be found in most chapters, thanks to the author's numerous publications on the topic and to studies performed ad hoc for this book. (orig.)

  8. Light-induced bird strikes on vessels in Southwest Greenland

    DEFF Research Database (Denmark)

    Merkel, Flemming Ravn; Johansen, Kasper Lambert

    2011-01-01

    Light-induced bird strikes are known to occur when vessels navigate during darkness in icy waters using powerful searchlight. In Southwest Greenland, which is important internationally for wintering seabirds, we collected reports of incidents of bird strikes over 2–3 winters (2006–2009) from navy...... vessels, cargo vessels and trawlers (total n = 19). Forty-one incidents were reported: mainly close to land (birds were reported killed in a single incident. All occurred between 5 p.m. and 6 a.m. and significantly more birds were involved when...... visibility was poor (snow) rather than moderate or good. Among five seabird species reported, the common eider (Somateria mollissima) accounted for 95% of the bird casualties. Based on spatial analyses of data on vessel traffic intensity and common eider density we are able to predict areas with high risk...

  9. Stress analysis and evaluation of a rectangular pressure vessel

    International Nuclear Information System (INIS)

    Rezvani, M.A.; Ziada, H.H.; Shurrab, M.S.

    1992-10-01

    This study addresses structural analysis and evaluation of an abnormal rectangular pressure vessel, designed to house equipment for drilling and collecting samples from Hanford radioactive waste storage tanks. It had to be qualified according to ASME boiler and pressure vessel code, Section VIII; however, it had the cover plate bolted along the long face, a configuration not addressed by the code. Finite element method was used to calculate stresses resulting from internal pressure; these stresses were then used to evaluate and qualify the vessel. Fatigue is not a concern; thus, it can be built according to Section VIII, Division I instead of Division 2. Stress analysis was checked against the code. A stayed plate was added to stiffen the long side of the vessel

  10. Innovations in prestressed concrete pressure vessel design

    International Nuclear Information System (INIS)

    Chow, P.Y.; Ngo, D.; Lin, T.Y.

    1979-01-01

    The study explored a new approach to the design of a high-pressure PCPV that accepts tension and tension cracks in the outer region of the PCPV. It examined the possibility of incorporating artificially-introduced preformed separations that pre-determined crack locations in the design as a method of controlling high tensile stresses generated by internal temperature and pressure. The results showed that the PCPV so designed was, in the extreme case of the DSV, approximately 70% cheaper than the 18 steel vessels of equivalent capacity it replaces. (orig.)

  11. PWR vessel flaw distribution development

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

    1990-01-01

    This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

  12. Nuclear reactor vessel decontamination systems

    International Nuclear Information System (INIS)

    McGuire, P. J.

    1985-01-01

    There is disclosed in the present application, a decontamination system for reactor vessels. The system is operatable without entry by personnel into the contaminated vessel before the decontamination operation is carried out and comprises an assembly which is introduced into the vertical cylindrical vessel of the typical boiling water reactor through the open top. The assembly includes a circular track which is centered by guideways permanently installed in the reactor vessel and the track guides opposed pairs of nozzles through which water under very high pressure is directed at the wall for progressively cutting and sweeping a tenacious radioactive coating as the nozzles are driven around the track in close proximity to the vessel wall. The whole assembly is hoisted to a level above the top of the vessel by a crane, outboard slides on the assembly brought into engagement with the permanent guideways and the assembly progressively lowered in the vessel as the decontamination operation progresses. The assembly also includes a low pressure nozzle which forms a spray umbrella above the high pressure nozzles to contain radioactive particles dislodged during the decontamination

  13. Gammatography of thick lead vessels

    International Nuclear Information System (INIS)

    Raghunath, V.M.; Bhatnagar, P.K.; Sundaram, V.M.

    1979-01-01

    Radiography, scintillation and GM counting and dose measurements using ionisation chamber equipment are commonly used for detecting flaws/voids in materials. The first method is mostly used for steel vessels and to a lesser extent thin lead vessels also and is essentially qualitative. Dose measuring techniques are used for very thick and large lead vessels for which high strength radioactive sources are required, with its inherent handling problems. For vessels of intermediate thicknesses, it is ideal to use a small strength source and a GM or scintillation counter assembly. At the Reactor Research Centre, Kalpakkam, such a system was used for checking three lead vessels of thicknesses varying from 38mm to 65mm. The tolerances specified were +- 4% variation in lead thickness. The measurements also revealed the non concentricity of one vessel which had a thickness varying from 38mm to 44mm. The second vessel was patently non-concentric and the dimensional variation was truly reproduced in the measurements. A third vessel was fabricated with careful control of dimensions and the measurements exhibited good concentricity. Small deviations were observed, attributable to imperfect bondings between steel and lead. This technique has the following advantages: (a) weaker sources used result in less handling problems reducing the personnel exposures considerably; (b) the sensitivity of the instrument is quite good because of better statistics; (c) the time required for scanning a small vessel is more, but a judicious use of a scintillometer for initial fast scan will help in reducing the total scanning time; (d) this method can take advantage of the dimensional variations themselves to get the calibration and to estimate the deviations from specified tolerances. (auth.)

  14. BY FRUSTUM CONFINING VESSEL

    Directory of Open Access Journals (Sweden)

    Javad Khazaei

    2016-09-01

    Full Text Available Helical piles are environmentally friendly and economical deep foundations that, due to environmental considerations, are excellent additions to a variety of deep foundation alternatives available to the practitioner. Helical piles performance depends on soil properties, the pile geometry and soil-pile interaction. Helical piles can be a proper alternative in sensitive environmental sites if their bearing capacity is sufficient to support applied loads. The failure capacity of helical piles in this study was measured via an experimental research program that was carried out by Frustum Confining Vessel (FCV. FCV is a frustum chamber by approximately linear increase in vertical and lateral stresses along depth from top to bottom. Due to special geometry and applied bottom pressure, this apparatus is a proper choice to test small model piles which can simulate field stress conditions. Small scale helical piles are made with either single helix or more helixes and installed in fine grained sand with three various densities. Axial loading tests including compression and tension tests were performed to achieve pile ultimate capacity. The results indicate the helical piles behavior depends essentially on pile geometric characteristics, i.e. helix configuration and soil properties. According to the achievements, axial uplift capacity of helical model piles is about equal to usual steel model piles that have the helixes diameter. Helical pile compression bearing capacity is too sufficient to act as a medium pile, thus it can be substituted other piles in special geoenvironmental conditions. The bearing capacity also depends on spacing ratio, S/D, and helixes diameter.

  15. Vessel eddy current characteristics in SST-1 tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Jana, Subrata; Pradhan, Subrata, E-mail: pradhan@ipr.res.in; Dhongde, Jasraj; Masand, Harish

    2016-11-15

    Highlights: • Eddy current distribution in the SST-1 vacuum vessel. • Circuit model analysis of eddy current. • A comparison of the field lines with and without the plasma column in identical conditions. • The influence of eddy current in magnetic NULL dynamics. - Abstract: Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285 m and minor axis of 0.81 m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

  16. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Hagiwara, Koji; Imura, Yasuya.

    1979-01-01

    Purpose: To provide constituted method for easily performing baking of vacuum vessel, using short-circuiting segments. Constitution: At the time of baking, one turn circuit is formed by the vacuum vessel and short-circuiting segments, and current transformer converting the one turn circuit into a secondary circuit by the primary coil and iron core is formed, and the vacuum vessel is Joule heated by an induction current from the primary coil. After completion of baking, the short-circuiting segments are removed. (Kamimura, M.)

  17. PWR vessel inspection performance improvements

    International Nuclear Information System (INIS)

    Blair Fairbrother, D.; Bodson, Francis

    1998-01-01

    A compact robot for ultrasonic inspection of reactor vessels has been developed that reduces setup logistics and schedule time for mandatory code inspections. Rather than installing a large structure to access the entire weld inspection area from its flange attachment, the compact robot examines welds in overlapping patches from a suction cup anchor to the shell wall. The compact robot size allows two robots to be operated in the vessel simultaneously. This significantly reduces the time required to complete the inspection. Experience to date indicates that time for vessel examinations can be reduced to fewer than four days. (author)

  18. Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

    2014-09-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

  19. Special enclosure for a pressure vessel

    International Nuclear Information System (INIS)

    Wedellsborg, B.W.; Wedellsborg, U.W.

    1993-01-01

    A pressure vessel enclosure is described comprising a primary pressure vessel, a first pressure vessel containment assembly adapted to enclose said primary pressure vessel and be spaced apart therefrom, a first upper pressure vessel jacket adapted to enclose the upper half of said first pressure vessel containment assembly and be spaced apart therefrom, said upper pressure vessel jacket having an upper rim and a lower rim, each of said rims connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, mean for connecting in a sealable relationship said upper rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, means for connecting in a sealable relationship said lower rim of said first upper pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a first lower pressure vessel jacket adapted to enclose the lower half of said first pressure vessel containment assembly and be spaced apart therefrom, said lower pressure vessel jacket having an upper rim connected in a slidable relationship to the outer surface of said first pressure vessel containment assembly, and means for connecting in a sealable relationship said upper rim of said first lower pressure vessel jacket to the outer surface of said first pressure vessel containment assembly, a second upper pressure vessel jacket adapted to enclose said first upper pressure vessel jacket and be spaced apart therefrom, said second upper pressure vessel jacket having an upper rim and a lower rim, each of said rims adapted to slidably engage the outer surface of said first upper pressure vessel jacket, means for sealing said rims, a second lower pressure vessel jacket adapted to enclose said first lower pressure vessel jacket and be spaced apart therefrom

  20. CT and MR angiographic findings in dissection of cervical vessels

    International Nuclear Information System (INIS)

    Link, J.; Brinkmann, G.; Heuser, K.; Heller, M.

    1996-01-01

    Purpose: To determine the usefulness of CT angiography (CTA) and MR angiography (MRA) for evaluation of dissection in cervical vessels. Material and methods: Dissection of cervical vessels was revealed by conventional angiography in 4 patients (two female, two male) of 30-62 years of age. Dissection was located in the carotid artery (n=3) and in the vertebral artery (n=1). In two patients CTA and in two patients MRA was performed. Results: Diagnosis of dissection was possible by CTA (internal carotid artery: n=2) and by MRA (internal carotid artery and vertebral artery). Imaging of the dissection membrane of the vessel wall was possible in one case with MRA. Conclusion: CT and MR angiography was successful for detection of typical morphology of dissection in all cases. If results in a greater number can be obtained it seems to be conceivable that both methods can be used in primary diagnosis. (orig.) [de

  1. 2013 West Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  2. 2011 West Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  3. 2013 Great Lakes Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  4. 2011 Great Lakes Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  5. 2011 East Coast Vessel Tracklines

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Automatic Identification Systems (AIS) are a navigation safety device that transmits and monitors the location and characteristics of many vessels in U.S. and...

  6. Coastal Discard Logbook Survey (Vessels)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains data on the type and amount of marine resources that are discarded or interacted with by vessels that are selected to report to the Southeast...

  7. SC/OQ Vessel Database

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Data tables holding information for the Surf Clam/Ocean Quahog vessel and dealer/processor logbooks (negative and positive), as well as individual tag information...

  8. Vessel Permit System Data Set

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — GARFO issues federal fishing permits annually to owners of fishing vessels who fish in the Greater Atlantic region, as required by federal regulation. These permits...

  9. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

    Strosnider, J.; Wichman, K.; Elliot, B.

    1994-12-01

    This report gives a brief description of the reactor pressure vessel (RPV), followed by a discussion of the radiation embrittlement of RPV beltline materials and the two indicators for measuring embrittlement, the end-of-license (EOL) reference temperature and the EOL upper-shelf energy. It also summarizes the GL 92-01 effort and presents, for all 37 boiling water reactor plants and 74 pressurized water reactor plants in the United States, the current status of compliance with regulatory requirements related to ensuring RPV integrity. The staff has evaluated the material data needed to predict neutron embrittlement of the reactor vessel beltline materials. These data will be stored in a computer database entitled the reactor vessel integrity database (RVID). This database will be updated annually to reflect the changes made by the licensees in future submittals and will be used by the NRC staff to assess the issues related to vessel structural integrity

  10. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-07-01

    A successful technical demonstration of simulated reactor vessel sectioning was completed using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel: an air arc gouger was chosen to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. After the simulated vessel was successfully cut from the SS side, another cut was made, starting from the carbon steel side. This cut was also successful. Cutting from the carbon steel side has the advantages of cost reduction since the air arc gouging step is eliminated and contamination controlled because the molten metal is blown inward

  11. Power reactor pressure vessel benchmarks

    International Nuclear Information System (INIS)

    Rahn, F.J.

    1978-01-01

    A review is given of the current status of experimental and calculational benchmarks for use in understanding the radiation embrittlement effects in the pressure vessels of operating light water power reactors. The requirements of such benchmarks for application to pressure vessel dosimetry are stated. Recent developments in active and passive neutron detectors sensitive in the ranges of importance to embrittlement studies are summarized and recommendations for improvements in the benchmark are made. (author)

  12. Maritime Training Serbian Autonomous Vessel Protection Detachment

    Directory of Open Access Journals (Sweden)

    Šoškić Svetislav D.

    2014-06-01

    Full Text Available The crisis in Somalia has caused appearance of piracy at sea in the Gulf of Aden and the Western Indian Ocean. Somali pirates have become a threat to economic security of the world because almost 30 percent of world oil and 20 percent of global trade passes through the Gulf of Aden. Solving the problem of piracy in this part of the world have included international organizations, institutions, military alliances and the states, acting in accordance with international law and UN Security Council resolutions. The European Union will demonstrate the application of a comprehensive approach to solving the problem of piracy at sea and the crisis in Somalia conducting naval operation — EU NAVFOR Atalanta and operation EUTM under the Common Security and Defense Policy. The paper discusses approaches to solving the problem of piracy in the Gulf of Aden and the crisis in Somalia. Also, the paper points to the complexity of the crisis in Somalia and dilemmas correctness principles that are applied to solve the problem piracy at sea. One of goals is protections of vessels of the World Food Programme (WFP delivering food aid to displaced persons in Somalia. Republic of Serbia joined in this mission and trained and sent one a autonomous team in this military operation for protection WFP. This paper consist the problem of modern piracy, particularly in the area of the Horn of Africa became a real threat for the safety of maritime ships and educational process of Serbian Autonomous vessel protection detachment. Serbian Military Academy adopted and developed educational a training program against piracy applying all the provisions and recommendations of the IMO conventions and IMO model courses for Serbian Autonomous vessel protection detachment.

  13. Prosopomorphic vessels from Moesia Superior

    Directory of Open Access Journals (Sweden)

    Nikolić Snežana

    2008-01-01

    Full Text Available The prosopomorphic vessels from Moesia Superior had the form of beakers varying in outline but similar in size. They were wheel-thrown, mould-made or manufactured by using a combination of wheel-throwing and mould-made appliqués. Given that face vessels are considerably scarcer than other kinds of pottery, more than fifty finds from Moesia Superior make an enviable collection. In this and other provinces face vessels have been recovered from military camps, civilian settlements and necropolises, which suggests that they served more than one purpose. It is generally accepted that the faces-masks gave a protective role to the vessels, be it to protect the deceased or the family, their house and possessions. More than forty of all known finds from Moesia Superior come from Viminacium, a half of that number from necropolises. Although tangible evidence is lacking, there must have been several local workshops producing face vessels. The number and technological characteristics of the discovered vessels suggest that one of the workshops is likely to have been at Viminacium, an important pottery-making centre in the second and third centuries.

  14. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Kurita, Gen-ichi; Onozuka, Masaki; Suzuki, Masaru.

    1997-01-01

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and γ rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  15. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Kurita, Gen-ichi [Japan Atomic Energy Research Inst., Tokyo (Japan); Onozuka, Masaki; Suzuki, Masaru

    1997-07-31

    Heat of inner walls of a vacuum vessel that receive radiation heat from plasmas by way of first walls is removed by a cooling medium flowing in channels for cooling the inner walls. Nuclear heat generation of constitutional materials of the vacuum vessel caused by fast neutrons and {gamma} rays is removed by a cooling medium flowing in cooling channels disposed in the vacuum vessel. Since the heat from plasmas and the nuclear heat generation are removed separately, the amount of the cooling medium flowing in the channels for cooling inner walls is increased for cooling a great amount of heat from plasmas while the amount of the cooling medium flowing in the channels for cooling the inside of the vacuum vessel is reduced for cooling the small amount of nuclear heat generation. Since the amount of the cooling medium can thus be optimized, the capacity of the facilities for circulating the cooling medium can be reduced. In addition, since the channels for cooling the inner walls and the channels of cooling medium formed in the vacuum vessel are disposed to the inner walls of the vacuum vessel on the side opposite to plasmas, integrity of the channels relative to leakage of the cooling medium can be ensured. (N.H.)

  16. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    Energy Technology Data Exchange (ETDEWEB)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10{sup -4} Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  17. ITER cryostat main chamber and vacuum vessel pressure suppression system design

    International Nuclear Information System (INIS)

    Ito, Akira; Nakahira, Masataka; Takahashi, Hiroyuki; Tada, Eisuke; Nakashima, Yoshitane; Ueno, Osamu

    1999-03-01

    Design of Cryostat Main Chamber and Vacuum Vessel Pressure Suppression System (VVPS) of International Thermonuclear Experimental Reactor (ITER) has been conducted. The cryostat is a cylindrical vessel that includes in-vessel component such as vacuum vessel, superconducting toroidal coils and poloidal coils. This cryostat provides the adiabatic vacuum about 10 -4 Pa for the superconducting coils operating at 4 K and forms the second confinement barrier to tritium. The adiabatic vacuum is to reduce thermal loads applied to the superconducting coils and their supports so as to keep their temperature 4 K. The VVPS consists of a suppression tank located under the lower bio-shield and 4 relief pipes to connect the vacuum vessel and the suppression tank. The VVPS is to keep the maximum pressure rise of the vacuum vessel below the design value of 0.5 MPa in case of the in-vessel LOCA (water spillage from in-vessel component). The spilled water and steam are lead to the suppression tank through the relief pipes when the internal pressure of vacuum vessel is over 0.2 MPa, and then the internal pressure is kept below 0.5 MPa. This report summarizes the structural design of the cryostat main chamber and pressure suppression system, together with their fabrication and installation. (author)

  18. Apparatus for locating inspection device in a nuclear reactor vessel

    International Nuclear Information System (INIS)

    1980-01-01

    A method for accurately locating an inspection device with a PWR or BWR pressure vessel uses a plurity of guide members and an internal location element, the exact position of which is known. Used for defining the size, orientation and position of a flow. (U.K.)

  19. Concrete containment vessels (CCV) for nuclear power plants, (1)

    International Nuclear Information System (INIS)

    Ibe, Yukimi; Kitajima, Masatake

    1977-01-01

    Containment vessels (CV) and the construction of concrete containment vessels (CCV) for nuclear power plants are described generally, and their use and techniques in foreign countries are illustrated, in connection with the introduction of CCV to Japanese nuclear power plants. The introduction deals with the construction plan of Japanese nuclear power plants, and with the difficulties in the steel CV for large scale construction. The investigations, tests and researches are not yet sufficient. The prompt establishment of safety supported by technical criteria, analytical methods and experiments is desired. The second part deals with the consideration for aseismatic design, construction, function and characteristics of CCV. The classification and currently employed CCV, which is mainly reinforced concrete containment vessels (RCCV), are described, and the typical CCV employed for BWR is illustrated. Further, the typical arrangement of reinforcing steels at the cylindrical portion and the dome portion of RCCV is illustrated. The third part deals with the present state of CCV abroad. A prestressed concrete containment vessel (PCCV) of Turkey Point power plant is illustrated as a typical example of CCV. The tests reported in the international meeting for the design, construction and operation of concrete pressure vessels and concrete containment vessels at York University in England in 1975 are reviewed. Typical examples of the design conditions, the size and form, and the construction procedure for PCCV and RCCV abroad are reviewed. (Iwakiri, K.)

  20. ITER vacuum vessel dynamic stress analysis of a disruption

    International Nuclear Information System (INIS)

    Riemer, B.W.; Conner, D.L.; Strickler, D.J.; Williamson, D.E.

    1994-01-01

    Dynamic stress analysis of the International Thermonuclear Experimental Reactor vacuum vessel loaded by disruption forces was performed. The deformation and stress results showed strong inertial effects when compared to static analyses. Maximum stress predicted dynamically was 300 MPa, but stress shown by static analysis from loads at the same point in time reached only 80 MPa. The analysis also provided a reaction load history in the vessel's supports which is essential in evaluating support design. The disruption forces were estimated by assuming a 25-MA plasma current decaying at 1 MA/ms while moving vertically. In addition to forces developed within the vessel, vertical loadings from the first wall/strong back assemblies and the divertor were applied to the vessel at their attachment points. The first 50 natural modes were also determined. The first mode's frequency was 6.0 Hz, and its shape is characterized by vertical displacement of the vessel inner leg. The predicted deformation of the vessel appeared similar to its first mode shape combined with radial contraction. Kinetic energy history from the analysis also correlated with the first mode frequency

  1. Automated method for identification and artery-venous classification of vessel trees in retinal vessel networks.

    Science.gov (United States)

    Joshi, Vinayak S; Reinhardt, Joseph M; Garvin, Mona K; Abramoff, Michael D

    2014-01-01

    The separation of the retinal vessel network into distinct arterial and venous vessel trees is of high interest. We propose an automated method for identification and separation of retinal vessel trees in a retinal color image by converting a vessel segmentation image into a vessel segment map and identifying the individual vessel trees by graph search. Orientation, width, and intensity of each vessel segment are utilized to find the optimal graph of vessel segments. The separated vessel trees are labeled as primary vessel or branches. We utilize the separated vessel trees for arterial-venous (AV) classification, based on the color properties of the vessels in each tree graph. We applied our approach to a dataset of 50 fundus images from 50 subjects. The proposed method resulted in an accuracy of 91.44% correctly classified vessel pixels as either artery or vein. The accuracy of correctly classified major vessel segments was 96.42%.

  2. 75 FR 56903 - Pacific Halibut Fisheries; Limited Access for Guided Sport Charter Vessels in Alaska

    Science.gov (United States)

    2010-09-17

    .... 100503209-0430-02] RIN 0648-AY85 Pacific Halibut Fisheries; Limited Access for Guided Sport Charter Vessels... program for charter vessels in the guided sport fishery for Pacific halibut in the waters of International... necessary to achieve the halibut fishery management goals of the North Pacific Fishery Management Council...

  3. 76 FR 26178 - Modifications to Treatment of Aircraft and Vessel Leasing Income

    Science.gov (United States)

    2011-05-06

    ... Modifications to Treatment of Aircraft and Vessel Leasing Income AGENCY: Internal Revenue Service (IRS... final regulations addressing the treatment of certain income and assets related to the leasing of... controlled foreign corporations that derive income from the leasing of aircraft or vessels in foreign...

  4. Complete versus culprit-only revascularisation in ST elevation myocardial infarction with multi-vessel disease

    DEFF Research Database (Denmark)

    Bravo, Claudio A.; Hirji, Sameer A.; Bhatt, Deepak L.

    2017-01-01

    Background: Multi-vessel coronary disease in people with ST elevation myocardial infarction (STEMI) is common and is associated with worse prognosis after STEMI. Based on limited evidence, international guidelines recommend intervention on only the culprit vessel during STEMI. This, in turn, leaves...

  5. Computational study of the mixed cooling effects on the in-vessel retention of a molten pool in a nuclear reactor

    International Nuclear Information System (INIS)

    Kim, Byung Seok; Sohn, Chang Hyun; Ahn, Kwang Il

    2004-01-01

    The retention of a molten pool vessel cooled by internal vessel reflooding and/or external vessel reactor cavity flooding has been considered as one of severe accident management strategies. The present numerical study investigates the effect of both internal and external vessel mixed cooling on an internally heated molten pool. The molten pool is confined in a hemispherical vessel with reference to the thermal behavior of the vessel wall. In this study, our numerical model used a scaled-down reactor vessel of a KSNP (Korea Standard Nuclear Power) reactor design of 1000 MWe (a pressurized water reactor with a large and dry containment). Well-known temperature-dependent boiling heat transfer curves are applied to the internal and external vessel cooling boundaries. Radiative heat transfer has been considered in the case of dry internal vessel boundary condition. Computational results show that the external cooling vessel boundary conditions have better effectiveness than internal vessel cooling in the retention of the melt pool vessel failure

  6. Clay Corner: Recreating Chinese Bronze Vessels.

    Science.gov (United States)

    Gamble, Harriet

    1998-01-01

    Presents a lesson where students make faux Chinese bronze vessels through slab or coil clay construction after they learn about the history, function, and design of these vessels. Utilizes a variety of glaze finishes in order to give the vessels an aged look. Gives detailed guidelines for creating the vessels. (CMK)

  7. Vacuum vessel for thermonuclear device

    International Nuclear Information System (INIS)

    Kikuchi, Mitsuru; Nagashima, Keisuke; Suzuki, Masaru; Onozuka, Masaki.

    1997-01-01

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  8. Vacuum vessel for thermonuclear device

    Energy Technology Data Exchange (ETDEWEB)

    Kikuchi, Mitsuru; Nagashima, Keisuke [Japan Atomic Energy Research Inst., Tokyo (Japan); Suzuki, Masaru; Onozuka, Masaki

    1997-07-11

    A vacuum vessel main body and structural members at the inside and the outside of the vacuum vessel main body are constituted by structural materials activated by irradiation of neutrons from plasmas such as stainless steels. Shielding members comprising tungsten or molybdenum are disposed on the surface of the vacuum vessel main body and the structural members of the inside and the outside of the main body. The shielding members have a function also as first walls or a seat member for the first walls. Armor tiles may be disposed to the shielding members. The shielding members and the armor tiles are secured to a securing seat member disposed, for example, to an inner plate of the vacuum vessel main body by bolts. Since the shielding members are disposed, it is not necessary to constitute the vacuum vessel main body and the structural members at the inside and the outside thereof by using a low activation material which is less activated, such as a titanium alloy. (I.N.)

  9. Molten material-containing vessel

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko

    1998-01-01

    The molten material-containing vessel of the present invention comprises a vessel main body having an entrance opened at the upper end, a lid for closing the entrance, an outer tube having an upper end disposed at the lower surface of the lid, extended downwardly and having an closed lower end and an inner tube disposed coaxially with the outer tube. When a molten material is charged from the entrance to the inside of the vessel main body of the molten material-containing vessel and the entrance is closed by the lid, the outer tube and the inner tube are buried in the molten material in the vessel main body, accordingly, a fluid having its temperature elevated by absorption of the heat of the molten material rises along the inner circumferential surface of the outer tube, abuts against the lower surface of the lid and cooled by exchanging heat with the lid and forms a circulating flow. Since the heat in the molten material is continuously absorbed by the fluid, transferred to the lid and released from the lid to the atmospheric air, heat releasing efficiency can be improved compared with conventional cases. (N.H.)

  10. Visual tritium imaging of In-Vessel surfaces

    International Nuclear Information System (INIS)

    Gentile, C. A.; Zweben, S. J.; Skinner, C. H.; Young, K. M.; Langish, S. W.; Nishi, M. F.; Shu, W. M.; Parker, J.; Isobe, K.

    2000-01-01

    A imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

  11. Visual tritium imaging of in-vessel surfaces

    International Nuclear Information System (INIS)

    Gentile, C.A.; Zweben, S.J.; Skinner, C.H.; Young, K.M.; Langish, S.W.; Nishi, M.F.; Shu, W.M.; Parker, J.; Isobe, K.

    2000-01-01

    An imaging detector has been developed for the purpose of providing a non-destructive, real time method of determining tritium concentrations on the surface of internal TFTR vacuum vessel components. The detector employs a green phosphor screen (P31, zinc sulfide: copper) with a wave length peak of 530 nm, a charge-coupled device (CCD) camera linked to a computer, and a detection chamber for inserting components recovered from the vacuum vessel. This detector is capable of determining tritium concentrations on the surfaces. The detector provides a method of imaging tritium deposition on the surfaces in a fairly rapid fashion

  12. MELCOR ex-vessel LOCA simulations for ITER+

    International Nuclear Information System (INIS)

    Gaeta, M.J.; Merrill, B.J.; Bartels, H.W.

    1995-01-01

    Ex-vessel Loss-of-Coolant-Accident (LOCA) simulations for the International Thermonuclear Experimental Reactor (ITER) were performed using the MELCOR code. The main goals of this work were to estimate the ultimate pressurization of the heat transport system (HTS) vault in order to gauge the potential for stack releases and to estimate the total amount of hydrogen generated during a design basis ex-vessel LOCA. Simulation results indicated that the amount of hydrogen produced in each transient was below the flammability limit for the plasma chamber. In addition, only moderate pressurization of the HTS vault indicated a very small potential for releases through the stack

  13. Latest feedback from a major reactor vessel dismantling project

    International Nuclear Information System (INIS)

    Boucau, J.; Segerud, P.; Sanchez, M.; Garcia, R.

    2015-01-01

    Westinghouse performed two large segmentation projects in 2010-2013 and then 2013-2015 at the Jose Cabrera nuclear power plant in Spain. The power plant is located in Almonacid de Zorita, 43 miles east of Madrid, Spain and was in operation between 1968 and 2006. This paper will describe the sequential steps required to prepare, segment, separate, and package the individual component segments using under water mechanical techniques. The paper will also include experiences and lessons learned that Westinghouse has collected from the activities performed during the reactor vessel and vessel internals segmentation projects. (authors)

  14. Emergency venting of pressure vessels

    International Nuclear Information System (INIS)

    Steinkamp, H.

    1995-01-01

    With the numerical codes developed for safety analysis the venting of steam vessel can be simulated. ATHLET especially is able to predict the void fraction depending on the vessel height. Although these codes contain a one-dimensional model they allow the description of complex geometries due to the detailed nodalization of the considered apparatus. In chemical reactors, however, the venting process is not only influenced by the flashing behaviour but additionally by the running chemical reaction in the vessel. Therefore the codes used for modelling have to consider the kinetics of the chemical reaction. Further multi-component systems and dissolving processes have to be regarded. In order to preduct the fluid- and thermodynamic process it could be helpful to use 3-dimensional codes in combination with the one-dimensional codes as used in nuclear industry to get a more detailed describtion of the running processes. (orig./HP)

  15. Development of Catamaran Fishing Vessel

    Directory of Open Access Journals (Sweden)

    A. Jamaluddin

    2010-11-01

    Full Text Available Multihull due to a couple of advantages has been the topic of extensive research work in naval architecture. In this study, a series of investigation of fishing vessel to save fuel energy was carried out at ITS. Two types of ship models, monohull (round bilge and hard chine and catamaran, a boat with two hulls (symmetrical and asymmetrical were developed. Four models were produced physically and numerically, tested (towing tank and simulated numerically (CFD code. The results of the two approaches indicated that the catamaran mode might have drag (resistance smaller than those of monohull at the same displacement. A layout of catamaran fishing vessel, proposed here, indicates the freedom of setting the deck equipments for fishing vessel.

  16. 33 CFR 90.3 - Pushing vessel and vessel being pushed: Composite unit.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 1 2010-07-01 2010-07-01 false Pushing vessel and vessel being... HOMELAND SECURITY INLAND NAVIGATION RULES INLAND RULES: INTERPRETATIVE RULES § 90.3 Pushing vessel and vessel being pushed: Composite unit. Rule 24(b) of the Inland Rules states that when a pushing vessel and...

  17. LWR pressure vessel irradiation surveillance dosimetry. Quarterly progress report, July--September 1978

    Energy Technology Data Exchange (ETDEWEB)

    Guthrie, G L; McElroy, W N; Lippincott, E P; Gold, R

    1978-12-01

    Program objectives and progress to date by the national laboratories in LWR pressure vessel irradiation surveillance dosimetry are summarized. Participants in the program include: Rockwell International, Hanford Engineering Development Laboratory, National Bureau of Standards, and Oak Ridge National Laboratory.

  18. Containment vessel design and practice

    International Nuclear Information System (INIS)

    Bangash, Y.

    1983-01-01

    The state of the art of analysis and design of the concrete containment vessels required for BWR and PWR is reviewed. A step-by-step critical appraisal of the existing work is given. Elastic, inelastic and cracking conditions under extreme loads are fully discussed. Problems associated with these structures are highlighted. A three-dimensional finite element analysis is included to cater for service, overload and dynamic cracking of such structures. Missile impact and seismic effects are included in this work. The second analysis is known as the limit state analysis, which is given to design such vessels for any kind of load. (U.K.)

  19. Stress analysis of pressure vessels

    International Nuclear Information System (INIS)

    Kim, B.K.; Song, D.H.; Son, K.H.; Kim, K.S.; Park, K.B.; Song, H.K.; So, J.Y.

    1979-01-01

    This interim report contains the results of the effort to establish the stress report preparation capability under the research project ''Stress analysis of pressure vessels.'' 1978 was the first year in this effort to lay the foundation through the acquisition of SAP V structural analysis code and a graphic terminal system for improved efficiency of using such code. Software programming work was developed in pre- and post processing, such as graphic presentation of input FEM mesh geometry and output deformation or mode shope patterns, which was proven to be useful when using the FEM computer code. Also, a scheme to apply fracture mechanics concept was developed in fatigue analysis of pressure vessels. (author)

  20. Vessel dilatation in coronary angiograms

    International Nuclear Information System (INIS)

    Hinterauer, L.; Goebel, N.

    1983-01-01

    Amongst 166 patients with aneurysms, ectasia or megaloarteries shown on coronary angiograms, 86.1% had dilated vessels as part of generalised coronary sclerosis (usually in patients with three-vessel disease). In 9%, dilatation was of iatrogenic origin and in 4.8% it was idiopathic. One patient had Marfan's syndrome. Amongst 9 000 patients, there were eight with megalo-arteries without stenosis; six of these had atypical angina and three suffered an infarct. Patients with definite dilatation of the coronary artery and stagnation of contrast flow required treatment. (orig.) [de

  1. Vessel dilatation in coronary angiograms

    Energy Technology Data Exchange (ETDEWEB)

    Hinterauer, L.; Goebel, N.

    1983-11-01

    Amongst 166 patients with aneurysms, ectasia or megaloarteries shown on coronary angiograms, 86.1% had dilated vessels as part of generalised coronary sclerosis (usually in patients with three-vessel disease). In 9%, dilatation was of iatrogenic origin and in 4.8% it was idiopathic. One patient had Marfan's syndrome. Amongst 9 000 patients, there were eight with megalo-arteries without stenosis; six of these had atypical angina and three suffered an infarct. Patients with definite dilatation of the coronary artery and stagnation of contrast flow required treatment.

  2. Limit analysis and design of containment vessels

    International Nuclear Information System (INIS)

    Save, M.

    1984-01-01

    In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness. In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon. The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed. A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods. The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the 'limit states' design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper. (orig.)

  3. Mortality of German travellers on passenger vessels.

    Science.gov (United States)

    Oldenburg, Marcus; Herzog, Jan; Püschel, Klaus; Harth, Volker

    2016-01-01

    In the past two decades, more and more Germans decided to spend their holidays on a passenger vessel. This study examined the frequencies and causes of deaths of German travellers aboard passenger vessels of all flags. The shipboard deaths of all German travellers within the time period from 1998 to 2008 were counted using the German civil central register in Berlin. The available documentation in this register provides information on frequencies, circumstances and causes of deaths on ships. In the above-mentioned period of time, the total cohort of German travellers on cruise ships is estimated to be 5.97 million persons. During the 11-year examination period, 135 shipboard deaths of German passengers [102 males (75.6%) and 33 females (24.4%)] were recorded. Out of these travellers, 110 died on cruise ships. When considering only the passengers on cruise ships (without those on ferries) an average crude mortality rate of 1.8 per 100,000 German passengers was calculated. The crude mortality rate of shipboard death for males and females was 2.5 and 0.8 per 100,000 German passengers with a mean age of 71.2 years [standard deviation (SD) 16.0 years] and 73.3 years (SD 16.0 years), respectively. Significantly, more deceased travellers older than 70 years were observed on traditional cruise ships and resort vessels than on passenger ferries (P = 0.001). The causes of death were documented in 85 cases (63.0%). Out of these documented deaths, 82 (96.5%) cases were regarded to be natural causes (particularly circulatory diseases) and 3 (3.5%) as unnatural causes (twice drowning and once an accidental fall). In spite of the large proportion of unknown causes of death, this study argues for a high significance of internal causes of deaths among German passengers. Thus, ship's doctors-particularly those on traditional cruise ships-should be well experienced in internal and geriatric medicines. © The Author 2016. Published by Oxford University Press on behalf of

  4. Experiments on performance of the multi-layered in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, K.H.; Kim, S.B.; Park, R.J.; Cheung, F.B.; Suh, K.Y.; Rempe, J.L.

    2004-01-01

    LAVA-GAP experiments are in progress to investigate the performance of the in-vessel core catcher using alumina melt as a corium simulant. The hemispherical in-vessel core catcher made of carbon steel was installed inside the lower head vessel with a uniform gap of 10 mm. Until now, two types of the in-vessel core catcher were used in this study. The first one is a single layered in-vessel core catcher without an internal coating of the LAVA-GAP-2 test, and the other one is a two layered in-vessel core catcher with a 0.5 mm-thick ZrO 2 internal coating of the LAVA-GAP-3 test. Current LAVA-GAP experimental results indicate that an internally coated in-vessel core catcher has better thermal performance compared with an uncoated in-vessel core catcher. Metallurgical inspections on the test specimens of the LAVA-GAP-3 test have been performed to examine the performance of the coating material and the base carbon steel. Although the base carbon steel had experienced a severe thermal attack to the extent that the microstructures were changed and re-crystallization occurred, the carbon steel showed stable and pure chemical compositions without any oxidation and interaction with the coating layer. In terms of the material aspects, these metallurgical inspection results suggest that the ZrO 2 coating performed well. (authors)

  5. 30 seismic analysis of FBR vessels: Coupling between components and vessels, fluid communications, imperfections

    International Nuclear Information System (INIS)

    Gantenbein, F.; Gibert, R.J.; Aita, S.; Durandet, E.

    1988-01-01

    The internal structures of a loop type breeder reactors such as SUPERPHENIX are composed of axisymmetrical shells separated by fluid volumes. Seismic analysis is usually performed by axisymmetric finite element model taking into account fluid structure interaction but the geometry is in fact 3D due to components, small communications between fluid volumes and imperfections in the vessels. The methods to take this 3D behaviour into account are based on Fourier decomposition of the motion and substructuration. They are briefly described in the following chapters. The influence of components and of small communications on a block reactor similar to SPX1 will also be presented. 15 refs, 20 figs

  6. Design and fabrication of the vacuum vessel for the Advanced Toroidal Facility

    International Nuclear Information System (INIS)

    Chipley, K.K.; Frey, G.N.

    1985-01-01

    The vacuum vessel for the Advanced Toroidal Facility (ATF) is a heavily contoured and very complex formed vessel that is specifically designed to allow for maximum plasma volume in a pure stellarator arrangement. The design of the facility incorporates an internal vessel that is closely fitted to the two helical field coils following the winding law theta = 1/6phi. Metallic seals have been incorporated throughout the system to minimize impurities. The vessel has been fabricated utilizing a comprehensive set of tooling fixtures specifically designed for the task of forming 6-mm stainless steel plate to the complex shape. Computer programs were used to develop a series of ribs that essentially form an internal mold of the vessel. Plates were press-formed with multiple compound curves, fitted to the fixture, and joined with full-penetration welds. 7 refs., 8 figs

  7. Vulnerability analysis of a pressurized aluminum composite vessel against hypervelocity impacts

    Directory of Open Access Journals (Sweden)

    Hereil Pierre-Louis

    2015-01-01

    Full Text Available Vulnerability of high pressure vessels subjected to high velocity impact of space debris is analyzed with the response of pressurized vessels to hypervelocity impact of aluminum sphere. Investigated tanks are CFRP (carbon fiber reinforced plastics overwrapped Al vessels. Explored internal pressure of nitrogen ranges from 1 bar to 300 bar and impact velocity are around 4400 m/s. Data obtained from Xrays radiographies and particle velocity measurements show the evolution of debris cloud and shock wave propagation in pressurized nitrogen. Observation of recovered vessels leads to the damage pattern and to its evolution as a function of the internal pressure. It is shown that the rupture mode is not a bursting mode but rather a catastrophic damage of the external carbon composite part of the vessel.

  8. A system for the thermal insulation of a pre-stressed concrete vessel

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1975-01-01

    This invention concerns the thermal insulation of a pre-stressed concrete vessel for a pressurised water nuclear reactor, this vessel being fitted internally with a leak-proof metal lining. Two rings are placed at the lower and upper parts of the vessel respectively. The upper ring is closed with a cover. These rings differ in diameter, are fitted with a metal insulating and mark the limits of a chamber between the vaporisable fluid and the internal wall of the vessel. This chamber is filled with a fluid in the liquid phase up to the liquid/vapor interface level of the fluid and with a gas above that level, the covering of the rings forming a cold fluid liquid seal. Each ring is supported by the vessel. Leak-proof components take up the radial expansion of the rings [fr

  9. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    International Nuclear Information System (INIS)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun; Moriyama, Kiyofumi; Park, Jin Ho

    2016-01-01

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  10. Development of an ex-vessel corium debris bed with two-phase natural convection in a flooded cavity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eunho; Lee, Mooneon; Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr; Moriyama, Kiyofumi; Park, Jin Ho

    2016-03-15

    Highlights: • For ex-vessel severe accidents in LWRs with wet-cavity strategy, development of debris bed with two-phase natural convection flow due to thermal characteristics of prototypic corium particles was investigated experimentally by using simulant particles and local air bubble control system. • Based on the experimental results of this study, an analytical model was established to describe the spreading of the debris bed in terms of two-phase flow and the debris injection parameters. • This model was then used to analyze the formation of debris beds at the reactor scale, and a sensitivity analysis was carried out based on key accident parameters. - Abstract: During severe accidents of light water reactors (LWRs), the coolability of relocated corium from the reactor vessel is a significant safety issue and a threat to the integrity of containment. With a flooded cavity, a porous debris bed is expected to develop on the bottom of the pool due to breakup and fragmentation of the melt jet. As part of the coolability assessment under accident conditions, the geometrical configuration of the debris bed is important. The Debris Bed Research Apparatus for Validation of the Bubble-Induced Natural Convection Effect Issue (DAVINCI) experimental apparatus facility was constructed to investigate the formation of debris beds under the influence of a two-phase flow induced by steam generation due to the decay heat of the debris bed. Using this system, five kilograms of stainless steel simulant debris were injected from the top of the water level, while air bubbles simulating the vapor flow were injected from the bottom of the particle catcher plate. The airflow rate was determined based on the quantity of settled debris, which will form a heat source due to the decay of corium. The radial distribution of the settled debris was examined using a ‘gap–tooth’ approach. Based on the experimental results of this study, an analytical model was established to

  11. 75 FR 67386 - Policy for Banning of Foreign Vessels From Entry into United States Ports

    Science.gov (United States)

    2010-11-02

    ... International Maritime Organization (IMO) Resolution A.741 (18), titled ``International Management Code for the... condition have failed to recognize the importance of complying with international conventions and standards and put their crews, vessels, and the marine environment at risk. Occasionally, the U.S. Coast Guard...

  12. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    1983-01-01

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

  13. Structural design and manufacturing of TPE-RX vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y. [Mitsubishi Heavy Ind. Ltd., Kobe (Japan); Urata, K.; Kudough, F. [Mitsubishi Heavy Industries, Ltd., Tokyo (Japan); Hasegawa, M.; Oyabu, I. [Mitsubishi Electric Co., Tokyo (Japan); Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H. [Electrotechnical Laboratory, Tsukuba (Japan)

    1999-10-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  14. Structural design and manufacturing of TPE-RX vacuum vessel

    International Nuclear Information System (INIS)

    Sago, H.; Orita, J.; Kaguchi, H.; Ishigami, Y.; Urata, K.; Kudough, F.; Hasegawa, M.; Oyabu, I.; Yagi, Y.; Sekine, S.; Shimada, T.; Hirano, Y.; Sakakita, H.; Koguchi, H.

    1999-01-01

    TPE-RX is a newly constructed, large-sized reversed field pinch (RFP) machine installed at the Electrotechnical Laboratory of the Ministry of International Trade and Industry in Japan. This is the third largest RFP in the world. Major and minor radii of the plasma are 1.72 and 0.45 m, respectively. TPE-RX aims to optimize plasma confinement up to 1 MA. RFP plasma configuration was successfully obtained in March 1998. This paper reports the structural design and manufacturing of the vacuum vessel of TPE-RX. The supporting system on the bellows sections of the vessel was designed based on a detailed finite element method. The integrity of the vacuum vessel against a plasma disruption has been confirmed using dynamic inelastic analyses. (orig.)

  15. Seismic proving test of PWR reactor containment vessel

    International Nuclear Information System (INIS)

    Akiyama, H.; Yoshikawa, T.; Tokumaru, Y.

    1987-01-01

    The seismic reliability proving tests of nuclear power plant facilities are carried out by Nuclear Power Engineering Test Center (NUPEC), using the large-scale, high-performance vibration of Tadotsu Engineering Laboratory, and sponsored by the Ministry of International Trade and Industry (MITI). In 1982, the seismic reliability proving test of PWR containment vessel started using the test component of reduced scale 1/3.7 and the test component proved to have structural soundness against earthquakes. Subsequently, the detailed analysis and evaluation of these test results were carried out, and the analysis methods for evaluating strength against earthquakes were established. Whereupon, the seismic analysis and evaluation on the actual containment vessel were performed by these analysis methods, and the safety and reliability of the PWR reactor containment vessel were confirmed

  16. Optimized Baking of the DIII-D Vessel

    International Nuclear Information System (INIS)

    Anderson, P.M.; Kellman, A.G.

    1999-01-01

    The DIII-D tokamak vacuum vessel baking system is used to heat the vessel walls and internal hardware to an average temperature of 350 C to allow rapid conditioning of the vacuum surfaces. The system combines inductive heating and a circulating hot air system to provide rapid heating with temperature uniformity required by stress considerations. In recent years, the time to reach 350 C had increased from 9 hrs to 14 hrs. To understand and remedy this sluggish heating rate, an evaluation of the baking system was recently performed. The evaluation indicated that the mass of additional in-vessel hardware (50% increase in mass) was primarily responsible. This paper reports on this analysis and the results of the addition of an electric air heater and procedural changes that have been implemented. Preliminary results indicate that the time to 350 C has been decreased to 4.5 hours and the temperature uniformity has improved

  17. Aging of reactor vessels in LWR type reactors

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.

    2004-01-01

    Most of the degradation mechanisms of nuclear components were not included on the design so they have to be treated a posteriori, and that imply a loss of capacity. In this paper the state of the art on the reactor pressure vessel neutron embrittlement and on the irradiation assisted stress corrosion cracking that affects internal components, are explained. Special attention is devoted on the influence of the neutron fluence on IASCC process, on the material alterations promoted by irradiation and their consequences on the susceptibility to this phenomenon. Regarding the reactor pressure vessel degradation, this paper discuss the application of the Master Curve on the structural integrity evaluation of the vessel. Other aspects related to further developments are also mentioned and the importance of a good materials ageing management on the operation of the plant is pointed out. (Author) 12 refs

  18. Armed guards on vessels : insurance and liability

    Directory of Open Access Journals (Sweden)

    Mišo Mudrić

    2011-12-01

    Full Text Available The Paper examines the insurance and liability issues resulting from the use of armed guards on board vessels. The study begins with an overview of the available data on key economic fi gures representing the projected overall annual costs of modern piracy. The focus is then shifted to the issue of public versus private security, where possible dangers of private-based security options are discussed in general. After explaining why the Somalia region deserves a closer attention when compared to other pirate-infested waters, a brief summary of the international effort to combat piracy threat is presented, followed by a structured overview of the use of private maritime security options in the maritime sector in general. One security option is the use of armed guards on board vessels. This option is explored both from the political (the acceptance by stakeholders and legal standpoint (legal issues arising from the use of armed guards. An important remedy for the shipping companies/ operators threatened by the piracy hazard is the existence of affordable and effective (specialized marine insurance. A study of available piracy insurance policies is presented, followed by an analysis of case law and other legal issues arising from piracy attacks, which could prove important when considering the legal implications of armed guards employment. Finally, a simplifi ed economic analysis of available security options is presented, followed by the final assessment of benefi ts derived from the use of armed guards.

  19. Autonomous Radiation Monitoring of Small Vessels

    International Nuclear Information System (INIS)

    Fabris, Lorenzo; Hornback, Donald Eric

    2010-01-01

    Small private vessels are one avenue by which nuclear materials may be smuggled across international borders. While one can contemplate using the terrestrial approach of radiation portal monitors on the navigable waterways that lead to many ports, these systems are ill-suited to the problem. They require vehicles to pass at slow speeds between two closely-spaced radiation sensors, relying on the uniformity of vehicle sizes to space the detectors, and on proximity to link an individual vehicle to its radiation signature. In contrast to roadways where lanes segregate vehicles, and motion is well controlled by inspection booths; channels, inlets, and rivers present chaotic traffic patterns populated by vessels of all sizes. We have developed a unique solution to this problem based on our portal-less portal monitor instrument that is designed to handle free-flowing traffic on roadways with up to five-traffic lanes. The instrument uses a combination of visible-light and gamma-ray imaging to acquire and link radiation images to individual vehicles. It was recently tested in a maritime setting. In this paper we present the instrument, how it functions, and the results of the recent tests.

  20. Grounding Damage to Conventional Vessels

    DEFF Research Database (Denmark)

    Lützen, Marie; Simonsen, Bo Cerup

    2003-01-01

    The present paper is concerned with rational design of conventional vessels with regard to bottom damage generated in grounding accidents. The aim of the work described here is to improve the design basis, primarily through analysis of new statistical data for grounding damage. The current regula...

  1. PLANNING VESSEL BODY SECTION PRODUCTION

    Directory of Open Access Journals (Sweden)

    A. G. Grivachevsky

    2015-01-01

    Full Text Available A problem of planning production of a vessel body section is considered. The problem is reduced to the classic Johnson’s tree-machine flow-shop scheduling problem. A genetic algorithm and computer experiment to compare efficiency of this algorithm and the algorithm of full enumeration are described.

  2. Pressure vessel and method therefor

    Science.gov (United States)

    Saunders, Timothy

    2017-09-05

    A pressure vessel includes a pump having a passage that extends between an inlet and an outlet. A duct at the pump outlet includes at least one dimension that is adjustable to facilitate forming a dynamic seal that limits backflow of gas through the passage.

  3. Commercial Passenger Fishing Vessel Fishery

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data set contains the logbook data from U.S.A. Commercial Passenger Fishing Vessels (CPFV) fishing in the U.S.A. EEZ and in waters off of Baja California, from...

  4. Multi-purpose deployer for ITER in-vessel maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang-Hwan, E-mail: Chang-Hwan.CHOI@iter.org [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Tesini, Alessandro; Subramanian, Rajendran [ITER Organization, Route de Vinon-sur-Verdon, 13115 St Paul lez Durance (France); Rolfe, Alan; Mills, Simon; Scott, Robin; Froud, Tim; Haist, Bernhard; McCarron, Eddie [Oxford Technologies Ltd., 7 Nuffield Way, Abingdon, OXON (United Kingdom)

    2015-10-15

    Highlights: • ITER RH system called as the multi-purpose deployer (MPD) is introduced. • The MPD performs dust and tritium inventory control, in-service inspection. • The MPD performs leak localization, in-vessel diagnostics maintenance. • The MPD has nine degrees of freedom with a payload capacity up to 2 tons. - Abstract: The multi-purpose deployer (MPD) is a general purpose in-vessel remote handling (RH) system in the ITER RH system. The MPD provides the means for deployment and handling of in-vessel tools or components inside the vacuum vessel (VV) for dust and tritium inventory control, in-service inspection, leak localization, and in-vessel diagnostics. It also supports the operation of blanket first wall maintenance and neutral beam duct liner module maintenance operations. This paper describes the concept design of the MPD. The MPD is a cask based system, i.e. it stays in the hot cell building during the machine operation, and is deployed to the VV using the cask system for the in-vessel operations. The main part of the MPD is the articulated transporter which provides transportation and positioning of the in-vessel tools or components. The articulated transporter has nine degrees of freedom with a payload capacity up to 2 tons. The articulated transporter can cover the whole internal surface of the VV by switching between the four equatorial RH ports. Additionally it can use two non-RH equatorial ports to transfer large tools or components. A concept for in-cask tool exchange is developed which minimizes the cask transportation by allowing the MPD to stay in the VV during the tool exchange.

  5. Development of PWR pressure vessel steels

    International Nuclear Information System (INIS)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed

  6. Development of PWR pressure vessel steels

    Energy Technology Data Exchange (ETDEWEB)

    Druce, S.; Edwards, B.

    1982-01-01

    Requirements to be met by vessel steels for pressurized water reactors are analyzed. Chemicat composition of low-alloyed steels, mechanical properties of sheets and forgings made of these steels and changes in the composition and properties over the wall thickness of the reactor vessel are presented. Problems of the vessel manufacturing including welding and heat treatment processes of sheets and forgings are considered. Special attention is paid to steel embrittlement during vessel fabrication and operation (radiation embrittlement, thermal embrittlement). The role of non-metal inclusions and their effect on anisotropy of fracture toughness is discussed. Possible developments of vessel steels and procedures for producing reactor vessels are reviewed.

  7. Structural analysis of the JT-60SA cryostat vessel body

    Energy Technology Data Exchange (ETDEWEB)

    Botija, José, E-mail: jose.botija@ciemat.es [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Alonso, Javier; Fernández, Pilar; Medrano, Mercedes; Ramos, Francisco; Rincon, Esther; Soleto, Alfonso [Association EURATOM – CIEMAT, Avda. Complutense 40, 28040 Madrid (Spain); Davis, Sam; Di Pietro, Enrico; Tomarchio, Valerio [Fusion for Energy, JT-60SA European Home Team, 85748 Garching bei Munchen (Germany); Masaki, Kei; Sakasai, Akira; Shibama, Yusuke [JAEA – Japan Atomic Energy Agency, Naka Fusion Institute, Ibaraki 311-0193 (Japan)

    2013-10-15

    Highlights: ► Structural analysis to validate the JT-60SA cryostat vessel body design. ► Design code ASME 2007 “Boiler and Pressure Vessel Code. Section VIII”. ► First buckling mode: load multiplier of 10.644, higher than the minimum factor 4.7. ► Elastic and elastic–plastic stress analysis meets ASME against plastic collapse. ► Bolted fasteners have been analyzed showing small gaps closed by strong welding. -- Abstract: The JT-60SA cryostat is a stainless steel vacuum vessel (14 m diameter, 16 m height) which encloses the Tokamak providing the vacuum environment (10{sup −3} Pa) necessary to limit the transmission of thermal loads to the components at cryogenic temperature. It must withstand both external atmospheric pressure during normal operation and internal overpressure in case of an accident. The paper summarizes the structural analyses performed in order to validate the JT-60SA cryostat vessel body design. It comprises several analyses: a buckling analysis to demonstrate stability under the external pressure; an elastic and an elastic–plastic stress analysis according to ASME VIII rules, to evaluate resistance to plastic collapse including localized stress concentrations; and, finally, a detailed analysis with bolted fasteners in order to evaluate the behavior of the flanges, assuring the integrity of the vacuum sealing welds of the cryostat vessel body.

  8. Interpretation of strain measurements on nuclear pressure vessels

    International Nuclear Information System (INIS)

    Andersen, S.I.; Engbaek, P.

    1979-11-01

    Selected results from strain measurements on 4 nuclear pressure vessels are presented and discussed. The measurements were made in several different regions of the vessels: transition zones in vessel heads, flanges and bottom parts, nozzels, internal vessel structure and flange bolts. The results presented are based on data obtained by approximately 700 strain-gauges, and a comprehensive knowledge of the quality obtained by such measurements is established. It is shown that a thorough control procedure before and after the test as well as detailed knowledge of the behaviour of the signal from the individual gauges during the test is necessary. If this is omitted, it can be extremely difficult to distinguish between the real structural behaviour and a malfunctioning of a specific gauge installation. In general, most of the measuring results exhibit a very linear behaviour with a negligible zeroshift. However, deviations from linear behaviour are observed in several cases. This nonlinearity can be explained by friction (flange connections) or by gaps (concentrical nozzles) in certain regions, whereas local plastic deformations during the first pressure loadings of the vessel seem to be the reason in other regions. (author)

  9. Nonlinear analysis of end slabs in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.

    1978-01-01

    A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

  10. Cold source vessel development for the advanced neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Williams, P.T.; Lucas, A.T. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory (ORNL), will be a user-oriented neutron research facility that will produce the most intense flux of neutrons in the world. Among its many scientific applications, the productions of cold neutrons is a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410 mm diameter sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel`s inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design are being performed with multi-dimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This paper presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that will be used to verify the final design.

  11. Stress analysis of blanket vessel for JAERI experimental fusion reactor

    International Nuclear Information System (INIS)

    Sako, K.; Minato, A.

    1979-01-01

    A blanket structure of JAERI Experimental Fusion Reactor (JXFR) consists of about 2,300 blanket cells with round cornered rectangular cross sections (twelve slightly different shapes) and is placed in a vacuum vessel. Each blanket vessel is a double-walled thin-shell structure made of Type 316 stainless steel with a spherical domed surface at the plasma side. Ribs for coolant channel are provided between inner and outer walls. The blanket cell contains Li 2 O pebbles and blocks for tritium breeding and stainless steel blocks for neutron reflection. A coolant is helium gas at 10 kgf/cm 2 (0.98 MPa) and its inlet and outlet temperatures are 300 0 C and 500 0 C. The maxima of heat flux and nuclear heating rate at the first wall are 12 W/cm 2 and 2 W/cc. A design philosophy of the blanket structure is based on high tritium breeding ratio and more effective shielding performance. The thin-shell vessel with a rectangular cross section satisfies the design philosophy. We have designed the blanket structure so that the adjacent vessels are mutually supporting in order to decrease the large deformation and stress due to internal pressure in case of the thin-shell vessel. (orig.)

  12. TMI-2 Vessel Investigation Project (VIP) Metallurgical Program

    International Nuclear Information System (INIS)

    Diercks, D.R.; Neimark, L.A.

    1990-06-01

    The TMI-2 Vessel Investigation Project (VIP) Metallurgical Program is a part of the international TMI-2 Vessel Investigation Project being conducting jointly by the US Nuclear Regulatory Commission and the Organization for Economic Co-operation and Development (OECD). The overall project consists of three phases, namely (1) recovery of material samples from the lower head of the TMI-2 reactor, (2) examination and analysis of the lower head samples and the preparation and testing of archive material subjected to a similar thermal history, and (3) procurement, examination, and analysis of companion core material located adjacent to or near the lower head material. The specific objectives of the ANL Metallurgical Program, which comprises a major portion of Phase 2, are to prepare metallographic and mechanical test specimen blanks from the TMI-2 lower head material, prepare similar test specimen blanks from suitable archive material subjected to the appropriate thermal processing, determine the mechanical properties of the lower vessel head and archive materials under the conditions of the core-melt accident, and assess the lower head integrity and margin-to-failure during the accident. The ANL work consists of three tasks: (1) archive materials program, (2) fabrication of metallurgical and mechanical test specimens from the TMI-2 pressure vessel samples, and (3) mechanical property characterization of TMI-2 lower pressure vessel head and archive material

  13. Method of measuring density of gas in a vessel

    International Nuclear Information System (INIS)

    Shono, Kosuke.

    1981-01-01

    Purpose: To accurately measure the density of a gas in a vessel even at a loss-of-coolant accident in a BWR type reactor. Method: When at least one of the pressure or the temperature of gas in a vessel exceeds the usable range of a gas density measuring instrument due to a loss-of-coolant accident, the gas in the vessel is sampled, and the pressure or the temperature of the sampled gas are measured by matching them to the usable conditions of the gas density measuring instrument. Hydrogen gas and oxygen gas densities exceeding the usable range of the gas density measuring instrument are calculated by the following formulae based on the measured values. C'sub(O) = P sub(T).C sub(O)/P sub(T), C'sub(H) = C''sub(H).C'sub(O)/C''sub(O), where C sub(O), P sub(T), C'sub(H) represent the oxygen density, the total pressure and the hydrogen density of the internal pressure gas of the vessel after the respective gas density measuring instruments exceed the usable ranges; C sub(O), P sub(T) represent the oxygen density and the total pressure of the gas in the vessel before the gas density measuring instruments exceeded the usable range, and C''sub(H), C''sub(O) represent the hydrogen density and oxygen density of the respective sampled gases. (Kamimura, M.)

  14. Design optimization of a thin walled pressure vessel

    International Nuclear Information System (INIS)

    Sadiq, S.

    2001-01-01

    Design evaluation of a pressure vessel is not only to build confidence on its integrity but also to reduce structural weight and enhance the performance of the structure. Pressure vessel, e.g., a rocket motor not only has to withstand the high operating temperatures but it must also be able to survive the internal pressures and external aerodynamic forces and bending stresses during its operation in flight. A research program was devised to study the stresses, which are generated in a thin walled pressure vessel during actual operation and its simulation with cold testing technique, i.e., by means of hydrostatic testing employing electrical resistance strain gauges on the external surface of the cylinder. The objective of the research was to uphold the performance of the vessel by reducing its thickness from 6.09 to 5.5 mm (which of course reduces the safety factor margin from 1.8 to 1.5); thereby curtailing the overall structural weight and maintaining the efficiency of the vessel itself during its live operation. The techniques employed were hydrostatic testing, data acquisition system for obtaining data on strains from the electrical resistance strain gauges and later employing V on Mises yield criterion empirical relation to computer the stresses in hoop and longitudinal directions. (author)

  15. Pressure vessel for nuclear reactor plant consisting of several pre-stressed cast pressure vessels

    International Nuclear Information System (INIS)

    Bodmann, E.

    1984-01-01

    Several cylindrical pressure vessel components made of pressure castings are arranged on a sector of a circle around the cylindrical cast pressure vessel for accommodating the helium cooled HTR. Each component pressure vessel is connected to the reactor vessel by a horizontal gas duct. The contact surfaces between reactor and component pressure vessel are in one plane. In the spaces between the individual component pressure vessels, there are supporting blocks made of cast iron, which are hollow and also have flat surfaces. With the reactor vessel and the component pressure vessels they form a disc-shaped connecting part below and above the gas ducts. (orig./PW)

  16. Vacuum vessel for the tandem Mirror Fusion Test Facility

    International Nuclear Information System (INIS)

    Gerich, J.W.

    1986-01-01

    In 1980, the US Department of Energy gave the Lawrence Livermore National Laboratory approval to design and build a tandem Mirror Fusion Test Facility (MFTF-B) to support the goals of the National Mirror Program. We designed the MFTF-B vacuum vessel both to maintain the required ultrahigh vacuum environment and to structurally support the 42 superconducting magnets plus auxiliary internal and external equipment. During our design work, we made extensive use of both simple and complex computer models to arrive at a cost-effective final configuration. As part of this work, we conducted a unique dynamic analysis to study the interaction of the 32,000-tonne concrete-shielding vault with the 2850-tonne vacuum vessel system. To maintain a vacuum of 2 x 10 -8 torr during the physics experiments inside the vessel, we designed a vacuum pumping system of enormous capacity. The vacuum vessel (4200-m 3 internal volume) has been fabricated and erected, and acceptance tests have been completed at the Livermore site. The rest of the machine has been assembled, and individual systems have been successfully checked. On October 1, 1985, we began a series of integrated engineering tests to verify the operation of all components as a complete system

  17. Identification Of Damages Of Tribological Associations In Crankshaft And Piston Systems Of Two-Stroke Internal Combustion Engines Used As Main Propulsion In Sea-Going Vessels And Proposal Of Probabilistic Description Of Loads As Causes Of These Damages

    Directory of Open Access Journals (Sweden)

    Girtler Jerzy

    2015-04-01

    Full Text Available The article discusses damages of essential tribological associations in crankshaft and piston systems of large power two-stroke engines used as main engines, which take place during transport tasks performed by those ships. Difficulties are named which make preventing those damages impossible, despite the fact that the technical state of engines of this type is identified with the aid of complex diagnostic systems making use of advanced computer technology. It is demonstrated that one of causes of the damages is the lack of research activities oriented on recognising random properties of the loads leading to those damages. A proposal is made for the loads acting at a given time t on tribological associations in crankshaft and piston systems of internal combustion engines used as main engines to be considered as random variables Qt. At the same time the loads examined within a given time interval tr ≤ t ≤ tz would be considered stochastic processes {Q(t: t ≥ 0}. Essential properties of the loads of the abovementioned tribological associations are named and explained by formulating hypotheses which need empirical verification. Interval estimation is proposed for estimating the expected value E(Qt of the load Qt acting at time t. A relation is indicated between the mechanical load and the thermal load acting on tribological associations in the ship main engine crankshaft and piston system. A suggestion is formulated that a stochastic form of the relation between these types of load is to be searched for, rather than statistic relation, and a proposal is made to measure the intensity (strength of the stochastic relation using the Czuprow’s convergence coefficient.

  18. AFSC/FMA/Vessel Assessment Logging

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Vessels fishing trawl gear, vessels fishing hook-and-line and pot gear that are also greater than 57.5 feet overall, and shoreside and floating processing facilities...

  19. 2013 EPA Vessels General Permit (VGP)

    Data.gov (United States)

    U.S. Environmental Protection Agency — Information for any vessel that submitted a Notice of Intent (NOI), Notice of Termination (NOT), or annual report under EPA's 2013 Vessel General Permit (VGP)....

  20. Interstitial Cells of Blood Vessels

    Directory of Open Access Journals (Sweden)

    Vladimír Pucovský

    2010-01-01

    Full Text Available Blood vessels are made up of several distinct cell types. Although it was originally thought that the tunica media of blood vessels was composed of a homogeneous population of fully differentiated smooth muscle cells, more recent data suggest the existence of multiple smooth muscle cell subpopulations in the vascular wall. One of the cell types contributing to this heterogeneity is the novel, irregularly shaped, noncontractile cell with thin processes, termed interstitial cell, found in the tunica media of both veins and arteries. While the principal role of interstitial cells in veins seems to be pacemaking, the role of arterial interstitial cells is less clear. This review summarises the knowledge of the functional and structural properties of vascular interstitial cells accumulated so far, offers hypotheses on their physiological role, and proposes directions for future research.

  1. The Vessel Schedule Recovery Problem

    DEFF Research Database (Denmark)

    Brouer, Berit Dangaard; Plum, Christian Edinger Munk; Vaaben, Bo

    Maritime transportation is the backbone of world trade and is accountable for around 3% of the worlds CO2 emissions. We present the Vessel Schedule Recovery Problem (VSRP) to evaluate a given disruption scenario and to select a recovery action balancing the trade off between increased bunker cons...... consumption and the impact on the remaining network and the customer service level. The model is applied to 4 real cases from Maersk Line. Solutions are comparable or superior to those chosen by operations managers. Cost savings of up to 58% may be achieved.......Maritime transportation is the backbone of world trade and is accountable for around 3% of the worlds CO2 emissions. We present the Vessel Schedule Recovery Problem (VSRP) to evaluate a given disruption scenario and to select a recovery action balancing the trade off between increased bunker...

  2. Expanded Fermilab pressure vessel directory program

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

  3. Expanded Fermilab pressure vessel directory program

    International Nuclear Information System (INIS)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

  4. Vacuum vessel for plasma devices

    International Nuclear Information System (INIS)

    Yamada, Masao; Taguchi, Masami.

    1975-01-01

    Object: To permit effective utility of the space in the inner and outer sides of the container wall and also permit repeated assembly for use. Structure: Vacuum vessel wall sections are sealed together by means of welding bellows, and also flange portions formed at the end of the wall sections are coupled together by bolts and are sealed together with a seal ring and a seal cap secured by welding. (Nakamura, S.)

  5. Numerical investigation of the reactor pressure vessel behaviour under severe accident conditions taking into account the combined processes of the vessel creep and the molten pool natural convection

    International Nuclear Information System (INIS)

    Loktionov, V.D.; Mukhtarov, E.S.; Yaroshenko, N.I.; Orlov, V.E.

    1999-01-01

    Analysis of the WWER lower head behaviour and its failure has been performed for several molten pool structures and internal overpressure levels in a reactor pressure vessel (RPV). The different types of the molten pools (homogeneous, conventionally homogeneous, conventionally stratified, stratified) cover the bounding scenarios during a hypothetical severe accident. The parametric investigations of the failure mode and RPV behaviour for various molten pool types, its heights and internal overpressure levels are presented herein. A coupled treatment in this investigation includes: (i) a 2-D thermohydraulic analysis of a molten pool natural convection. Domestic NARAUFEM code has been used in this detailed analysis for prediction of the heat flux from the molten pool to the RPV inner surface; and (ii) a detailed 3-D transient thermal analysis of the RPV lower head. Domestic 3-D ASHTER-VVR finite element code has been used for the numerical simulations of the high temperature creep and failure of the lower head. The effect of an external RPV cooling, temperature-dependent physical properties of the molten pool and vessel steel, the hydrostatic forces and vessel dead-weight were taken into account in this study. The obtained results show that lower head failure occurs as a result of the vessel creep process which is significantly dependent on both an internal overpressure level and the type of molten pool structure. In particular, it was found that there were combinations of 'overpressure-molten pool structure' when the vessel failure started at the 'hot' layers of the vessel. (orig.)

  6. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  7. Reactor-vessel-sectioning demonstration

    International Nuclear Information System (INIS)

    Lundgren, R.A.

    1981-09-01

    A technical demonstration was successfully completed of simulated reactor vessel sectioning using the combined techniques of air arc gouging and flame cutting. A 4-ft x 3-ft x 9-in. thick sample was fabricated of A36 carbon steel to simulate a reactor vessel wall. A 1/4-in. layer of stainless steel (SS) was tungsten inert gas (TIG)-welded to the carbon steel. Several techniques were considered to section the simulated reactor vessel; air arc gouging was selected to penetrate the stainless steel, and flame cutting was selected to sever the carbon steel. Three sectioning operations were demonstrated. For all three, the operating parameters were the same; but the position of the sample was varied. For the first cut, the sample was placed in a horizontal position, and it was successfully severed from the SS side. For the second cut, the sample was turned over and cut from the carbon steel side. Cutting from the carbon steel side has the advantages of cost reduction

  8. 33 CFR 151.1512 - Vessel safety.

    Science.gov (United States)

    2010-07-01

    ... 33 Navigation and Navigable Waters 2 2010-07-01 2010-07-01 false Vessel safety. 151.1512 Section... River § 151.1512 Vessel safety. Nothing in this subpart relieves the master of the responsibility for ensuring the safety and stability of the vessel or the safety of the crew and passengers, or any other...

  9. Reactor vessel supported by flexure member

    International Nuclear Information System (INIS)

    Crawford, J.D.; Pankow, B.

    1975-01-01

    A description is given of a reactor pressure vessel which is provided with vertical support means in the form of circumferentially spaced columns upon which the vessel is mounted. The columns are adapted to undergo flexure in order to accommodate the thermally induced displacements experienced by the vessel during operational transients

  10. 19 CFR 4.97 - Salvage vessels.

    Science.gov (United States)

    2010-04-01

    ... United States and Great Britain ‘concerning reciprocal rights for United States and Canada in the... meaning of this statute. (e) A Mexican vessel may engage in a salvage operation on a Mexican vessel in any territorial waters of the United States in which Mexican vessels are permitted to conduct such operations by...

  11. Model tests for prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Stoever, R.

    1975-01-01

    Investigations with models of reactor pressure vessels are used to check results of three dimensional calculation methods and to predict the behaviour of the prototype. Model tests with 1:50 elastic pressure vessel models and with a 1:5 prestressed concrete pressure vessel are described and experimental results are presented. (orig.) [de

  12. 50 CFR 648.4 - Vessel permits.

    Science.gov (United States)

    2010-10-01

    ... carrying passengers for hire. (8) Atlantic bluefish vessels. (i) Commercial. Any vessel of the United... lands Atlantic bluefish in or from the EEZ in excess of the recreational possession limit specified at § 648.164 must have been issued and carry on board a valid commercial bluefish vessel permit. (ii) Party...

  13. Large vessel imaging using cosmic-ray muons

    International Nuclear Information System (INIS)

    Jenneson, P.M.

    2004-01-01

    Cosmic-ray muons are assessed for their practical use in the tomographic imaging of the internal composition of large vessels over 2 m in diameter. The technique is based on the attenuation and scattering of cosmic-ray muons passing through a vessel and has advantages over photon-based methods of tomography that it is extendable to object containing high-density materials over many tens of metres. The main disadvantage is the length of time required to produce images of sufficient resolution and hence cosmic ray muon tomography will be most suited to the imaging of large structures whose internal composition is effectively static for the duration of the imaging period. Simulation and theoretical results are presented here which demonstrate the feasibility of cosmic ray muon tomography

  14. In service inspection of superphenix 1 vessels: MIR

    International Nuclear Information System (INIS)

    Asty, M.; Viard, J.; Lerat, B.; Saglio, R.

    1985-02-01

    Presentation of the in-service inspection device, MIR, which has been specially developed for the visual and ultrasonic examination of Super Phenix 1 vessels (surface and internal defects). The inspections take place during fuel handling operations. The inspection device is a robot with a four-wheel drive vehicle which guidance along the welds is achieved by eddy-current devices; visual examination is performed by a television camera and ultrasonic probes are specially resistent to high temperatures

  15. Integrating Multiple Autonomous Underwater Vessels, Surface Vessels and Aircraft into Oceanographic Research Vessel Operations

    Science.gov (United States)

    McGillivary, P. A.; Borges de Sousa, J.; Martins, R.; Rajan, K.

    2012-12-01

    Autonomous platforms are increasingly used as components of Integrated Ocean Observing Systems and oceanographic research cruises. Systems deployed can include gliders or propeller-driven autonomous underwater vessels (AUVs), autonomous surface vessels (ASVs), and unmanned aircraft systems (UAS). Prior field campaigns have demonstrated successful communication, sensor data fusion and visualization for studies using gliders and AUVs. However, additional requirements exist for incorporating ASVs and UASs into ship operations. For these systems to be optimally integrated into research vessel data management and operational planning systems involves addressing three key issues: real-time field data availability, platform coordination, and data archiving for later analysis. A fleet of AUVs, ASVs and UAS deployed from a research vessel is best operated as a system integrated with the ship, provided communications among them can be sustained. For this purpose, Disruptive Tolerant Networking (DTN) software protocols for operation in communication-challenged environments help ensure reliable high-bandwidth communications. Additionally, system components need to have considerable onboard autonomy, namely adaptive sampling capabilities using their own onboard sensor data stream analysis. We discuss Oceanographic Decision Support System (ODSS) software currently used for situational awareness and planning onshore, and in the near future event detection and response will be coordinated among multiple vehicles. Results from recent field studies from oceanographic research vessels using AUVs, ASVs and UAS, including the Rapid Environmental Picture (REP-12) cruise, are presented describing methods and results for use of multi-vehicle communication and deliberative control networks, adaptive sampling with single and multiple platforms, issues relating to data management and archiving, and finally challenges that remain in addressing these technological issues. Significantly, the

  16. Manufacture of EAST VS In-Vessel Coil

    International Nuclear Information System (INIS)

    Long, Feng; Wu, Yu; Du, Shijun; Jin, Huan; Yu, Min; Han, Qiyang; Wan, Jiansheng; Liu, Bin; Qiao, Jingchun; Liu, Xiaochuan; Li, Chang; Cai, Denggang; Tong, Yunhua

    2013-01-01

    Highlights: • ITER like Stainless Steel Mineral Insulation Conductor (SSMIC) used for EAST Tokamak VS In-Vessel Coil manufacture first time. • Research on SSMIC fabrication was introduced in detail. • Two sets totally four single-turn VS coils were manufactured and installed in place symmetrically above and below the mid-plane in the vacuum vessel of EAST. • The manufacture and inspection of the EAST VS coil especially the joint for the SSMIC connection was described in detail. • The insulation resistances of all the VS coils have no significant reduction after endurance test. -- Abstract: In the ongoing latest update round of EAST (Experimental Advanced Superconducting Tokamak), two sets of two single-turn Vertical Stabilization (VS) coils were manufactured and installed symmetrically above and below the mid-plane in the vacuum vessel of EAST. The Stainless Steel Mineral Insulated Conductor (SSMIC) developed for ITER In-Vessel Coils (IVCs) in Institute of Plasma Physics, Chinese Academy of Science (ASIPP) was used for the EAST VS coils manufacture. Each turn poloidal field VS coil includes three internal joints in the vacuum vessel. The middle joint connects two pieces of conductor which together form an R2.3 m arc segment inside the vacuum vessel. The other two joints connect the arc segment with the two feeders near the port along the toroidal direction to bear lower electromagnetic loads during operation. Main processes and tests include material performances checking, conductor fabrication, joint connection and testing, coil forming, insulation performances measurement were described herein

  17. Acrolein generation stimulates hypercontraction in isolated human blood vessels

    International Nuclear Information System (INIS)

    Conklin, D.J.; Bhatnagar, A.; Cowley, H.R.; Johnson, G.H.; Wiechmann, R.J.; Sayre, L.M.; Trent, M.B.; Boor, P.J.

    2006-01-01

    Increased risk of vasospasm, a spontaneous hyperconstriction, is associated with atherosclerosis, cigarette smoking, and hypertension-all conditions involving oxidative stress, lipid peroxidation, and inflammation. To test the role of the lipid peroxidation- and inflammation-derived aldehyde, acrolein, in human vasospasm, we developed an ex vivo model using human coronary artery bypass graft (CABG) blood vessels and a demonstrated acrolein precursor, allylamine. Allylamine induces hypercontraction in isolated rat coronary artery in a semicarbazide-sensitive amine oxidase activity (SSAO) dependent manner. Isolated human CABG blood vessels (internal mammary artery, radial artery, saphenous vein) were used to determine: (1) vessel responses and sensitivity to acrolein, allylamine, and H 2 O 2 exposure (1 μM-1 mM), (2) SSAO dependence of allylamine-induced effects using SSAO inhibitors (semicarbazide, 1 mM; MDL 72274-E, active isomer; MDL 72274-Z, inactive isomer; 100 μM), (3) the vasoactive effects of two other SSAO amine substrates, benzylamine and methylamine, and (4) the contribution of extracellular Ca 2+ to hypercontraction. Acrolein or allylamine but not H 2 O 2 , benzylamine, or methylamine stimulated spontaneous and pharmacologically intractable hypercontraction in CABG blood vessels that was similar to clinical vasospasm. Allylamine-induced hypercontraction and blood vessel SSAO activity were abolished by pretreatment with semicarbazide or MDL 72274-E but not by MDL 72274-Z. Allylamine-induced hypercontraction also was significantly attenuated in Ca 2+ -free buffer. In isolated aorta of spontaneously hypertensive rat, allylamine-induced an SSAO-dependent contraction and enhanced norepinephrine sensitivity but not in Sprague-Dawley rat aorta. We conclude that acrolein generation in the blood vessel wall increases human susceptibility to vasospasm, an event that is enhanced in hypertension

  18. Magnetic resonance angiography of the neck vessels: technique and anatomy

    International Nuclear Information System (INIS)

    Carriero, A.; Salute, L.

    1990-01-01

    The authors identified the standard projections for studying neck vessels with magnetic resonance angiography. Sixty volunteers underwent angio-MR of the arterial neck vessels with FISP 3D FT sequences obtained on the coronal and sagittal planes. The gradient-echo sequence (FISP 3D FT) was acquired with TR=0.04-0.08 s and TE=15 ms, with 25 grade flip angle. Single excitated slices of thickness ranging from 1-2 mm were included in the acquisition volume. Theses sequences were subsequently processed by the maximum intensity projection method. Two radiologist examined our results to choose the optimal projections. We used a semi-quantitative scale which allowed us to distinguish 3 different diagnostic levels for each projection: well-visualized vessels, poorly-visualized, and non-visualized ones. For each section axial rotations were performed ranging from 0 grade to 180 grade, with 15 grade i ntervals. On the coronal plane, rotations from 45 grade to 45 grade were the optimal ones to visualize the studied vessels. The 0 grade- 15 grade- 30 grade- 45 grade- 135 grade- 165 grade- 180 grade projections allowed the common carotids to be clearly demonstrated together with the verterbal arteries. The other projections appeared to be useless for diagnostic purposes. On the saggittal plane, rotations from 60 grade to 120 grade were the optimal ones. The 90 grade projection allowed the demonstration of all the big arterial vessel of the neck, including carotid bifurcation and internal and external carotids. The assessment of the optimal diagnostic projections for angio-MR of the neck vessels is helpful to reduce post-processing time. As a matter of fact, the immediate visualization, during the examination, of the standard projections allows further acquisitions to be obtained- if needed- to try to solve specific diagnostic doubts

  19. Flaw distribution development from vessel ISI data

    International Nuclear Information System (INIS)

    Foulds, J.R.; Kennedy, E.L.; Basin, S.L.; Rosinski, S.T.

    1991-01-01

    Previous attempts to develop flaw distributions for use in the structural integrity evaluation of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all vessels. In contrast, this paper describes the analysis of vessel-specific in-service inspection (ISI) data for the development of a flaw distribution reliably representative of the condition of the particular vessel inspected. The application of the methodology may be extended to other vessels, but has been primarily developed for PWR reactor vessels. For this study, the flaw data analyzed included data obtained from three recently performed PWR vessel ISIs and from laboratory inspection of selected weldment sections of the Midland reactor vessel. The variability in both the character of the reviewed data (size range of flaws, number of flaws) and the UT (ultrasonic test) inspection system performance identified a need for analyzing the inspection results on a vessel-, or data set-specific basis. For this purpose, traditional histogram-based methods were inadequate, and a new methodology that can accept a very small number of flaws (typical of vessel-specific ISI results) and that includes consideration of inspection system flaw detection reliability, flaw sizing accuracy and flaw detection threshold, was developed. Results of the application of the methodology to each of the four PWR reactor vessel cases studied are presented and discussed

  20. Review of the current understanding of the potential for containment failure from in-vessel steam explosions

    International Nuclear Information System (INIS)

    1985-06-01

    A group of experts was convened to review the current understanding of the potential for containment failure from in-vessel steam explosions during core meltdown accidents in LWRs. The Steam Explosion Review Group (SERG) was requested to provide assessments of: (1) the conditional probability of containment failure due to a steam explosion, (2) a Sandia National Laboratory (SNL) report entitled ''An Uncertainty Study of PWR Steam Explosions,'' NUREG/CR-3369, (3) a SNL proposed steam explosion research program. This report summarizes the results of the deliberations of the review group. It also presents the detailed response of each individual member to each of the issues. The consensus of the SERG is that the occurrence of a steam explosion of sufficient energetics which could lead to alpha-mode containment failure has a low probability. The SERG members disagreed with the methodology used in NUREG/CR-3369 for the purpose of establishing the uncertainty in the probability of containment failure by a steam explosion. A consensus was reached among SERG members on the need for a continuing steam explosion research program which would improve our understanding of certain aspects of steam explosion phenomenology

  1. Primo vessel inside a lymph vessel emerging from a cancer tissue.

    Science.gov (United States)

    Lee, Sungwoo; Ryu, Yeonhee; Cha, Jinmyung; Lee, Jin-Kyu; Soh, Kwang-Sup; Kim, Sungchul; Lim, Jaekwan

    2012-10-01

    Primo vessels were observed inside the lymph vessels near the caudal vena cava of a rabbit and a rat and in the thoracic lymph duct of a mouse. In the current work we found a primo vessel inside the lymph vessel that came out from the tumor tissue of a mouse. A cancer model of a nude mouse was made with human lung cancer cell line NCI-H460. We injected fluorescent nanoparticles into the xenografted tumor tissue and studied their flow in blood, lymph, and primo vessels. Fluorescent nanoparticles flowed through the blood vessels quickly in few minutes, and but slowly in the lymph vessels. The bright fluorescent signals of nanoparticles disappeared within one hour in the blood vessels but remained much longer up to several hours in the case of lymph vessels. We found an exceptional case of lymph vessels that remained bright with fluorescence up to 24 hours. After detailed examination we found that the bright fluorescence was due to a putative primo vessel inside the lymph vessel. This rare observation is consistent with Bong-Han Kim's claim on the presence of a primo vascular system in lymph vessels. It provides a significant suggestion on the cancer metastasis through primo vessels and lymph vessels. Copyright © 2012. Published by Elsevier B.V.

  2. What is cerebral small vessel disease?

    International Nuclear Information System (INIS)

    Onodera, Osamu

    2011-01-01

    An accumulating amount of evidence suggests that the white matter hyperintensities on T 2 weighted brain magnetic resonance imaging predict an increased risk of dementia and gait disturbance. This state has been proposed as cerebral small vessel disease, including leukoaraiosis, Binswanger's disease, lacunar stroke and cerebral microbleeds. However, the concept of cerebral small vessel disease is still obscure. To understand the cerebral small vessel disease, the precise structure and function of cerebral small vessels must be clarified. Cerebral small vessels include several different arteries which have different anatomical structures and functions. Important functions of the cerebral small vessels are blood-brain barrier and perivasucular drainage of interstitial fluid from the brain parenchyma. Cerebral capillaries and glial endfeet, take an important role for these functions. However, the previous pathological investigations on cerebral small vessels have focused on larger arteries than capillaries. Therefore little is known about the pathology of capillaries in small vessel disease. The recent discoveries of genes which cause the cerebral small vessel disease indicate that the cerebral small vessel diseases are caused by a distinct molecular mechanism. One of the pathological findings in hereditary cerebral small vessel disease is the loss of smooth muscle cells, which is an also well-recognized finding in sporadic cerebral small vessel disease. Since pericytes have similar character with the smooth muscle cells, the pericytes should be investigated in these disorders. In addition, the loss of smooth muscle cells may result in dysfunction of drainage of interstitial fluid from capillaries. The precise correlation between the loss of smooth muscle cells and white matter disease is still unknown. However, the function that is specific to cerebral small vessel may be associated with the pathogenesis of cerebral small vessel disease. (author)

  3. 46 CFR 4.03-40 - Public vessels.

    Science.gov (United States)

    2010-10-01

    ... INVESTIGATIONS Definitions § 4.03-40 Public vessels. Public vessel means a vessel that— (a) Is owned, or demise... Department (except a vessel operated by the Coast Guard or Saint Lawrence Seaway Development Corporation...

  4. Proceedings of the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 28-29 April 2014, OECD-NEA HQ

    International Nuclear Information System (INIS)

    Kurata, Masaki; Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Sowder, Andrew; Koo, Yang-Hyun; Yang, Jae-Ho; Kim, Hyun-Gil; Zhou, Y.; Forgeron, T.; Guedeney, Ph.; Brachet, J.C.; Michaux, A.; Chauvin, Nathalie; Waeckel, N.; Ambard, A.; Blanpain, P.; Bischoff, J.; Zvonarev, Yu.; Verwerft, M.; Weber, M.; Lambrinou, K.; Koonen, E.; Van Dyck, S.; PETIT, Marc; Cornet, Stephanie; ); YAMAJI, Akifumi; ); Inozemtsev, V.; )

    2014-04-01

    Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Final Agenda; 2 - Draft mandate of EGATFL: Discussion of Scope and Objectives (K. Pasamehmetoglu, INL); 3 - Technical updates since the 2. meeting on ATF (28-29 October 2013): - Overview on ATF R and D in Japan (M. Kurata, JAEA); - Update on Development of Enhanced Accident Tolerant Fuel for Light Water Reactors in the United States (S. Bragg-Sitton, INL); - EPRI Update Since the 2. OECD/NEA Meeting on ATF - 28-29 October 2013 (A. Sowder, EPRI); - Accident Tolerant Fuel (ATF) Development: KAERI's R and D Status (Y.H. Koo, KAERI); - Accident Tolerant Fuel Research Activities in China General Nuclear Power Corporation - CGN (Y. Zhou, CGN); - ATF R and D Status and Perspectives (Th. Forgeron, CEA); - Proposals of NRC 'Kurchatov Institute' on Contributions to Collaborative Framework on ATF Activity (Y. Zvonarev, NRC KI); - Input to the

  5. Progress of ITER vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K., E-mail: Kimihiro.Ioki@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Bayon, A. [F4E, c/ Josep Pla, No. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Kim, B.C. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Kuzmin, E. [NTC “Sintez”, Efremov Inst., 189631 Metallostroy, St. Petersburg (Russian Federation); Le Barbier, R.; Martinez, J.-M. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Pathak, H. [ITER-India, A-29, GIDC Electronic Estate, Sector 25, Gandhinagar 382025 (India); Preble, J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); Sa, J.W. [NFRI, 52 Yeoeundong Yuseonggu, Daejeon 305-333 (Korea, Republic of); Terasawa, A.; Utin, Yu. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-lez-Durance (France); and others

    2013-10-15

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

  6. Progress of ITER vacuum vessel

    International Nuclear Information System (INIS)

    Ioki, K.; Bayon, A.; Choi, C.H.; Daly, E.; Dani, S.; Davis, J.; Giraud, B.; Gribov, Y.; Hamlyn-Harris, C.; Jun, C.; Levesy, B.; Kim, B.C.; Kuzmin, E.; Le Barbier, R.; Martinez, J.-M.; Pathak, H.; Preble, J.; Sa, J.W.; Terasawa, A.; Utin, Yu.

    2013-01-01

    Highlights: ► This covers the overall status and progress of the ITER vacuum vessel activities. ► It includes design, R and D, manufacturing and approval process of the regulators. ► The baseline design was completed and now manufacturing designs are on-going. ► R and D includes ISI, dynamic test of keys and lip-seal welding/cutting technology. ► The VV suppliers produced full-scale mock-ups and started VV manufacturing. -- Abstract: Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R and D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure

  7. Evaluation on In-vessel Source Term in PGSFR (2015 Results)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Seung Won; Chang, Won-Pyo; Ha, Kwi-Seok; Ahn, Sang June; Kang, Seok Hun; Choi, Chi-Woong; Lee, Kwi Lim; Jeong, Jae-Ho; Kim, Jin Su; Jeong, Taekyeong [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    This strategy requires nuclear plants to have features that prevent radionuclide release and multiple barriers to the escape from the plants of any radionuclides that are released despite preventive measures. Considerations of the ability to prevent and mitigate release of radionuclides arise at numerous places in the safety regulations of nuclear plants. The effectiveness of mitigative capabilities in nuclear plants is subject to quantitative analysis. The radionuclide input to these quantitative analyses of effectiveness is the Source Term (ST). All features of the composition, magnitude, timing, chemical form and physical form of accidental radionuclide release constitute the ST. Also, ST is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment. Since the TMI accident in 1979, extensive experimental and analytical information has been accumulated on the accident ST for LWRs. Such mechanistic models and computer codes as the MELCOR and MAAP have been developed. The results of extensive calculations and experiments have been used to formulate an alternative to the simple TID-14844 ST for regulatory purpose. The in-vessel STs are calculated through several phases: The inventory of each radionuclide is calculated by ORIGEN-2 code using the peak burnup conditions. The nominal value of the radiological inventory is multiplied by a factor of 1.1 as an uncertainty margin to give the radiological inventory. ST in the release from the core to primary sodium is calculated by using the assumption of 4S methodology. Lastly, ST in the release from the primary sodium to cover gas space is calculated by using the assumption of 4S methodology.

  8. The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

    International Nuclear Information System (INIS)

    Paladino, D.; Guentay, S.; Andreani, M.; Tkatschenko, I.; Brinster, J.; Dabbene, F.; Kelm, S.; Allelein, H. J.; Visser, D. C.; Benz, S.; Jordan, T.; Liang, Z.; Porcheron, E.; Malet, J.; Bentaib, A.; Kiselev, A.; Yudina, T.; Filippov, A.; Khizbullin, A.; Kamnev, M.; Zaytsev, A.; Loukianov, A.

    2012-01-01

    During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). The tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)

  9. Innovative decontamination technology by abrasion in vibratory vessels

    International Nuclear Information System (INIS)

    Fabbri, Silvio; Ilarri, Sergio

    2007-01-01

    Available in abstract form only. Full text of publication follows: The possibility of using conventional vibratory vessel technology as a decontamination technique is the motivation for the development of this project. The objective is to explore the feasibility of applying the vibratory vessel technology for decontamination of radioactively-contaminated materials such as pipes and metal structures. The research and development of this technology was granted by the U.S. Department of Energy (DOE). Abrasion processes in vibratory vessels are widely used in the manufacture of metals, ceramics, and plastics. Samples to be treated, solid abrasive media and liquid media are set up into a vessel. Erosion results from the repeated impact of the abrasive particles on the surface of the body being treated. A liquid media, generally detergents or surfactants aid the abrasive action. The amount of material removed increases with the time of treatment. The design and construction of the machine were provided by Vibro, Argentina private company. Tests with radioactively-contaminated aluminum tubes and a stainless steel bar, were performed at laboratory level. Tests showed that it is possible to clean both the external and the internal surface of contaminated tubes. Results show a decontamination factor around 10 after the first 30 minutes of the cleaning time. (authors)

  10. Design features of the KSTAR in-vessel control coils

    Energy Technology Data Exchange (ETDEWEB)

    Kim, H.K. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)], E-mail: hkkim@nfri.re.kr; Yang, H.L.; Kim, G.H.; Kim, Jin-Yong; Jhang, Hogun; Bak, J.S.; Lee, G.S. [National Fusion Research Institute (NFRI), 52 Yeoeun-dong, Yusung-ku, Daejeon, 305-333 (Korea, Republic of)

    2009-06-15

    In-vessel control coils (IVCCs) are to be used for the fast plasma position control, field error correction (FEC), and resistive wall mode (RWM) stabilization for the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The IVCC system comprises 16 segments to be unified into a single set to achieve following remarkable engineering advantages; (1) enhancement of the coil system reliability with no welding or brazing works inside the vacuum vessel, (2) simplification in fabrication and installation owing to coils being fabricated outside the vacuum vessel and installed after device assembly, and (3) easy repair and maintenance of the coil system. Each segment is designed in 8 turns coil of 32 mm x 15 mm rectangular oxygen free high conductive copper with a 7 mm diameter internal coolant hole. The conductors are enclosed in 2 mm thick Inconel 625 rectangular welded vacuum jacket with epoxy/glass insulation. Structural analyses were implemented to evaluate structural safety against electromagnetic loads acting on the IVCC for the various operation scenarios using finite element analysis. This paper describes the design features and structural analysis results of the KSTAR in-vessel control coils.

  11. An experimental study on in-vessel debris retention through gap cooling

    International Nuclear Information System (INIS)

    Kang, Kyung Ho; Kim, Jong Whan; Cho, Young Ro; Chang, Young Cho; Park, Rae Jun; Gu, Kil Mo; Kim, Sang Baik; Kim, Hee Dong

    1999-04-01

    LAVA experiments have been performed using high temperature molten material to be relocated into the 1/8 linear scaled vessel of a reactor lower plenum filled with water. An Al 2 O 3 /Fe tehrmite melt (Al 2 O 3 only) was used as a corium simulant. In this study, the influence of various initial conditions, such as internal pressure load across the vessel wall, the material composition of the melt simulant, water subcooling and depth on gap formation were investigated. As well, the thermal and mechanical behaviors of the vessel were examined. In case the internal pressure load was imposed, the gap formation between the continuous solidified debris and the vessel wall was clearly shown with post-test examination. On the other hand, in case the internal pressure load was not imposed, the iron welded to the inner surface of the vessel and the vessel experienced ablation to about 5 mm. The cooling rate of the vessel was very slow in the tests using Al 2 O 3 /Fe thermite melt but it was rather fast using Al 2 O 3 melt. It is postulated that in the Al 2 O 3 /Fe thermite tests, the iron melt layer is so dense that water ingression into the gap is difficult due to the high pressure of escaping steam. On the other hand, in the porosity of an Al 2 O 3 melt layer could enhance water ingression into the gap by giving the flow path of the evaporated steam through the porous media. The water height and subcooling could affect the melt pool formation and the initial thermal attack to the vessel. However, at present stage, the effect of water subcooling on the thermal behavior of the vessel couldn't be generalized. For clear confirmation of the effect of water subcooling, the tests will be performed at the saturated water condition. (author). 19 refs., 3 tabs., 36 figs

  12. BWR internals life assurance

    International Nuclear Information System (INIS)

    Herrera, M.L.; Stancavage, P.P.

    1988-01-01

    Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

  13. Nuclear reactor pressure vessel flaw distribution development

    International Nuclear Information System (INIS)

    Kennedy, E.L.; Foulds, J.R.; Basin, S.L.

    1991-12-01

    Previous attempts to develop flaw distributions for probabilistic fracture mechanics analyses of pressurized water reactor (PWR) vessels have aimed at the estimation of a ''generic'' distribution applicable to all PWR vessels. In contrast, this report describes (1) a new flaw distribution development analytic methodology that can be applied to the analysis of vessel-specific inservice inspection (ISI) data, and (2) results of the application of the methodology to the analysis of flaw data for each vessel case (ISI data on three PWR vessels and laboratory inspection data on sections of the Midland reactor vessel). Results of this study show significant variation among the flaw distributions derived from the various data sets analyzed, strongly suggesting than a vessel-specific flaw distribution (for vessel integrity prediction under pressurized thermal shock) is preferred over a ''generic'' distribution. In addition, quantitative inspection system flaw sizing accuracy requirements have been identified for developing a flaw distribution from vessel ISI data. The new flaw data analysis methodology also permits quantifying the reliability of the flaw distribution estimate. Included in the report are identified needs for further development of several aspects of ISI data acquisition and vessel integrity prediction practice

  14. Performance experiments on the in-vessel core catcher during severe accidents

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Rae Joon; Cho, Young Rho; Kim, Sang Baik

    2004-01-01

    A US-Korean International Nuclear Energy Research Initiative (INERI) project has been initiated by the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korean Atomic Energy Research Institute (KAERI) to determine if IVR is feasible for high power reactors up to 1500 MWe by investigating the performance of enhanced ERVC and in-vessel core catcher. This program is initially focusing on the Korean Advanced Power Reactor 1400 MWe (APR1400) design. As for the enhancement of the coolability through the ERVC, boiling tests are conducted by using appropriate coating material on the vessel outer surface to promote downward facing boiling and selecting an improved vessel/insulation design to facilitate water flow and steam venting through the insulation in this program. Another approach for successful IVR are investigated by applying the in-vessel core catcher to provide an 'engineered gap' between the relocated core materials and the water-filled reactor vessel and a preliminary design for an in-vessel core catcher was developed during the first year of this program. Feasibility experiments using the LAVA facility, named LAVA-GAP experiments, are in progress to investigate the core catcher performance based on the conceptual design of the in-vessel core catcher proposed in this INERI project. The experiments were performed using 60kg of Al 2 O 3 thermite melt as a core material simulant with a 1/8 linear scale mock-up of the reactor vessel lower plenum. The hemispherical in-vessel core catcher was installed inside the lower head vessel maintaining a uniform gap of 10mm from the inner surface of the lower head vessel. Two types of the core catchers were used in these experiments. The first one was a single layered in-vessel core catcher without internal coating and the second one was a two layered in-vessel core catcher with an internal coating of 0.5mm-thick ZrO 2 via the plasma

  15. A reactor vessel and its internals disassembled and packaged

    International Nuclear Information System (INIS)

    Eisenmann, B.; Prechtl, E.; Suessdorf, W.

    2008-01-01

    2007 was a successful year for the Disassembly Unit of the Karlsruhe Research Center: dismantling the highly activated Karlsruhe Multipurpose Research Reactor (MZFR) was completed successfully and without any incident. A vote of thanks is expressed at this point to all staff members, participating industries and institutions for their extraordinary commitment and their outstanding innovation and work. Preserving and advancing existing knowledge is one of the important pillars securing a sustainable future for generations to come in Germany. Securing and advancing know-how in nuclear technology was defined as a major duty last in late January 2008 by Dr. Peter Fritz, Chairman of the Kerntechnische Gesellschaft e.V. (KTG) and member of the Executive Board of the Karlsruhe Research Center, for instance, in cooperation between the KERNTECHNIK (i.e. nuclear technology) Southwestern Research and Teaching Association and the Karlsruhe Institute of Technology (KIT). The borders of Germany do not constitute a natural radiological barrier, despite the efforts by political groups in this country to convey this impression. This report therefore is to document that the demolition of a nuclear reactor with a high radioactivity inventory can be managed safely. In the light of the experience accumulated with the MZFR, this is feasible only if demolition is carried out immediately instead of the 'problem' of disassembly and conditioning being shifted to future generations. The article is also meant to be a piece of advice by showing the unplannable difficulties which came up, and the technical solutions implemented by the competence team successfully and efficiently. (orig.)

  16. Sampling of Reactor Pressure Vessel and Core Internals

    International Nuclear Information System (INIS)

    Oberhaeuser, R.

    2011-01-01

    Decommissioning and dismantling of nuclear power plants is a growing business, as a huge number of plants built in the 1970s have now reached their lifetime. It is well known that dismantling a nuclear power plant means an extraordinary expense for the owner respectively operator. Besides the dismantling works, the disposal of activated components and other nuclear waste is very expensive. Moreover, the fact that, in most countries, final disposal facilities are not available yet implies the need for interim storage on-site in specially built facilities. It can be concluded that a special attention is paid on producing a minimal radioactive waste volume. For this, optimized dismantling and packaging concepts have to be developed. The challenge is a fair balance between the obtainment of optimized packing and on the other side the fulfillment of stringent regulations set by the authorities and the storage requirements. The basis of a well-founded, optimized dismantling and packaging concept must always be the detailed knowledge of the radiological condition of the component to be dismantled. In the best case a 3- dimensional activation model contributes to this basis.

  17. Sodium steam generator within which are inlet and outlet ducts with pipe bundles in vessel

    International Nuclear Information System (INIS)

    1980-01-01

    The sodium steam generator with internal flow ducts for inlet and outlet to a vessel are provided as pipe bundles in the form of helically wound concentric layers terminating in inlet and outlet connections with chambers, characterised in that within the vessel, the pipe pieces which are connected to the pipe windings with the said vessel are arranged in substantially radially aligned rows so that each row measured in the circumferential direction at least on one side is at a spacing from the following row sufficiently large that between the rows or groups of rows an open sector is provided. (G.C.)

  18. Nuclear reactor installation with outer shell enclosing a primary pressure vessel

    International Nuclear Information System (INIS)

    1975-01-01

    The high temperature nuclear reactor installation described includes a fluid cooled nuclear heat source, a primary pressure vessel and outer shell around the primary pressure vessel and acting as a protection for it against outside projectiles. A floor is provided internally dividing the outside shell into two upper and lower sections and an inside wall dividing the lower section into one part containing the primary pressure vessel and a second part, both made pressure tight with respect to each other and with the outside shell and forming with the latter a secondary means of containment [fr

  19. Sense, enrich and classify: The scanmaris workshop for assessment of vessel's abnormal behavior in the EEZ

    OpenAIRE

    Jangal , Florent; Giraud , Marie-Annick; Morel , Michel; Mano , Jean-Pierre; Napoli , Aldo; Littaye , Anne

    2008-01-01

    International audience; Constant monitoring of the Exclusive Economic Zone cannot be performed only using high performance sensors. On the one hand, all available information on the observed area as juridical history of vessels or delineation of fishing zone is not necessarily measurable. On the other hand, even if the large amount of available information could be caught out they would be useless if none thorough sorting and analysis are carried on. So, we propose to sense vessel trail in Ex...

  20. LOFT reactor vessel 290/sup 0/ downcomer stalk instrument penetration flange stress analysis

    Energy Technology Data Exchange (ETDEWEB)

    Finicle, D.P.

    1978-06-06

    The LOFT Reactor Vessel 290/sup 0/ Downcomer Stalk Instrument Penetration Flange Stress Analysis has been completed using normal operational and blowdown loading. A linear elastic analysis was completed using simplified hand analysis techniques. The analysis was in accordance with the 1977 ASME Boiler and Pressure Vessel Code, Section III, for a Class 1 component. Loading included internal pressure, bolt preload, and thermal gradients due to normal operating and blowdown.

  1. Probabilistic approach to the analysis of reactor pressure vessel integrity during a pressurized thermal shock

    International Nuclear Information System (INIS)

    Adamec, P.

    2000-12-01

    Following a general summary of the issue, an overview of international experience (USA; Belgium, France, Germany, Russia, Spain, Sweden, The Netherlands, and the UK; and probabilistic PTS assessment for the reactor pressure vessel at Loviisa-1, Finland) is presented, and the applicable computer codes (VISA-II, OCA-P, FAVOR, ZERBERUS) are highlighted and their applicability to VVER type reactor pressure vessels is outlined. (P.A.)

  2. Investigation of flow stabilization in a compact reactor vessel of a FBR. Flow visualization in a reactor vessel

    International Nuclear Information System (INIS)

    Sato, Hiroyuki; Igarashi, Minoru; Kimura, Nobuyuki; Kamide, Hideki

    2002-01-01

    In the feasibility studies of Commercialized Fast Breeder Reactor Cycle System, a compact reactor vessel is considered from economical improvement point of a sodium cooled loop type fast reactor. The flow field was visualized by water experiment for a reactor vessel with 'a column type UIS (Upper Internal Structure)', which has a slit for fuel handling mechanism and is useful for a compact fast reactor. In this research, the 1/20 scale test equipment using water was made to understand coolant flow through a slit of a column type UIS' and fundamental behavior of reactor upper plenum flow. In the flow visualization tests, tracer particles were added in the water, and illuminated by the slit-shaped pulse laser. The flow visualization image was taken with a CCD camera. We obtained fluid velocity vectors from the visualization image using the Particle Imaging Velocimetry (PIV). The results are as follows. 1. Most of coolant flow through a slit of 'column type UIS' arrived the dip plate directly. In the opposite side of a slit, most of coolant flowed toward reactor vessel wall before it arrived the dip plate. 2. The PIV was useful to measure the flow field in the reactor vessel. The obtained velocity field was consistent with the flow visualization result. 3. The jet through the UIS slit was dependent on the UIS geometry. There is a possibility to control the jet by the UIS geometry. (author)

  3. Hydrogen storage in insulated pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, S.M.; Garcia-Villazana, O. [Lawrence Livermore National Lab., CA (United States)

    1998-08-01

    Insulated pressure vessels are cryogenic-capable pressure vessels that can be fueled with liquid hydrogen (LH{sub 2}) or ambient-temperature compressed hydrogen (CH{sub 2}). Insulated pressure vessels offer the advantages of liquid hydrogen tanks (low weight and volume), with reduced disadvantages (lower energy requirement for hydrogen liquefaction and reduced evaporative losses). This paper shows an evaluation of the applicability of the insulated pressure vessels for light-duty vehicles. The paper shows an evaluation of evaporative losses and insulation requirements and a description of the current analysis and experimental plans for testing insulated pressure vessels. The results show significant advantages to the use of insulated pressure vessels for light-duty vehicles.

  4. Neutron Assay System for Confinement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

  5. Automatic Vessel Segmentation on Retinal Images

    Institute of Scientific and Technical Information of China (English)

    Chun-Yuan Yu; Chia-Jen Chang; Yen-Ju Yao; Shyr-Shen Yu

    2014-01-01

    Several features of retinal vessels can be used to monitor the progression of diseases. Changes in vascular structures, for example, vessel caliber, branching angle, and tortuosity, are portents of many diseases such as diabetic retinopathy and arterial hyper-tension. This paper proposes an automatic retinal vessel segmentation method based on morphological closing and multi-scale line detection. First, an illumination correction is performed on the green band retinal image. Next, the morphological closing and subtraction processing are applied to obtain the crude retinal vessel image. Then, the multi-scale line detection is used to fine the vessel image. Finally, the binary vasculature is extracted by the Otsu algorithm. In this paper, for improving the drawbacks of multi-scale line detection, only the line detectors at 4 scales are used. The experimental results show that the accuracy is 0.939 for DRIVE (digital retinal images for vessel extraction) retinal database, which is much better than other methods.

  6. Electrical discharge machining for vessel sample removal

    International Nuclear Information System (INIS)

    Litka, T.J.

    1993-01-01

    Due to aging-related problems or essential metallurgy information (plant-life extension or decommissioning) of nuclear plants, sample removal from vessels may be required as part of an examination. Vessel or cladding samples with cracks may be removed to determine the cause of cracking. Vessel weld samples may be removed to determine the weld metallurgy. In all cases, an engineering analysis must be done prior to sample removal to determine the vessel's integrity upon sample removal. Electrical discharge machining (EDM) is being used for in-vessel nuclear power plant vessel sampling. Machining operations in reactor coolant system (RCS) components must be accomplished while collecting machining chips that could cause damage if they become part of the flow stream. The debris from EDM is a fine talclike particulate (no chips), which can be collected by flushing and filtration

  7. Exponential Stabilization of an Underactuated Surface Vessel

    Directory of Open Access Journals (Sweden)

    Kristin Y. Pettersen

    1997-07-01

    Full Text Available The paper shows that a large class of underactuated vehicles cannot be asymptotically stabilized by either continuous or discontinuous state feedback. Furthermore, stabilization of an underactuated surface vessel is considered. Controllability properties of the surface vessels is presented, and a continuous periodic time-varying feedback law is proposed. It is shown that this feedback law exponentially stabilizes the surface vessel to the origin, and this is illustrated by simulations.

  8. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  9. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    Frisch, E.; Andrews, H.N.

    1976-01-01

    A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

  10. How to replace a reactor pressure vessel

    International Nuclear Information System (INIS)

    Huber, R.

    1996-01-01

    A potential life extending procedure for a nuclear reactor after, say, 40 years of service life, might in some circumstances be the replacement of the reactor pressure vessel. Neutron induced degradation of the vessel might make replacement by one of a different material composition desirable, for example. Although the replacement of heavy components, such as steam generators, has been possible for many years, the pressure vessel presents a much more demanding task if only because it is highly irradiated. Some preliminary feasibility studies by Siemens are reported for the two removal strategies that might be considered. These are removal of the entire pressure vessel in one piece and dismantling it into sections. (UK)

  11. Prestressed reactor vessel for nuclear power plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1982-01-01

    With usual pressure vessels for nuclear reactor plants, especially for gas-cooled nuclear reactors, the load occurring due to the inner overpressure, especially the tensile load affecting the vessel top and/or bottom, their axis of inertia being horizontal, shall be compensated without a supplementary modification in design of the top and/or the bottom. This is attained by choosing an appropriate prestressing system of the vessel wall in the field the top and/or the bottom, so that the top and/or the bottom form a tension vault directed towards the interior of the vessel. (orig.) [de

  12. Reactor vessel decommissioning project. Final report

    International Nuclear Information System (INIS)

    Schoonen, D.H.

    1984-09-01

    This report describes a reactor vessel decommissioning project; it documents and explains the project objectives, scope, performance results, and sodium removal process. The project was successfully completed in FY-1983, within budget and without significant problems or adverse impact on the environment. Waste generated by the operation included the reactor vessel, drained sodium, and liquid, solid, and gaseous wastes which were significantly less than project estimates. Personnel radiation exposures were minimized, such that the project total was one-half the predicted exposure level. Except for the sodium removed, the material remaining in the reactor vessel is essentially the same as when the vessel arrived for processing

  13. Safety vessels for explosive fusion reactor

    International Nuclear Information System (INIS)

    Mineev, V.

    1994-01-01

    The failure of several types of geometrically similar cylindrical and spherical steel and glass fibers vessels filled with water or air was investigated when an explosive charge of TNT was detonated in the center. Vessels had radius 50-1000 mm, thickness of walls 2-20%. The detonation on TNT imitated energy release. The parameter: K = M/mf is a measure of the strength of the vessel where M is the mass of the vessel, and mf is the mass of TNT for which the vessel fails. This demanded 2-4 destroyed and nondestroyed shots. It may be showed that: K=A/σ f where σ f is the fracture stress of the material vessel, and A = const = F(energy TNT, characteristic of elasticity of vessel material). The chief results are the following: (1) A similar increase in the geometrical dimensions of steel vessels by a factor of 10 leads to the increase of parameter K in about 5 times and to decrease of failure deformation in 7 times (scale effect). (2) For glass fibers, scale effect is absent. (3) This problem is solved in terms of theory energetic scale effect. (4) The concept of TNT equivalent explosive makes it possible to use these investigations to evaluate the response of safety vessels for explosive fusion reactor

  14. Determinants of injuries in passenger vessel accidents.

    Science.gov (United States)

    Yip, Tsz Leung; Jin, Di; Talley, Wayne K

    2015-09-01

    This paper investigates determinants of crew and passenger injuries in passenger vessel accidents. Crew and passenger injury equations are estimated for ferry, ocean cruise, and river cruise vessel accidents, utilizing detailed data of individual vessel accidents that were investigated by the U.S. Coast Guard during the time period 2001-2008. The estimation results provide empirical evidence (for the first time in the literature) that crew injuries are determinants of passenger injuries in passenger vessel accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

  15. Increase of cyclic durability of pressure vessels

    International Nuclear Information System (INIS)

    Vorona, V.A.; Zvezdin, Yu.I.

    1980-01-01

    The durability of multilayer pressure vessels under cyclic loading is compared with single-layer vessels. The relative conditional durability is calculated taking into account the assumption on the consequent destruction of layers and viewing a vessel wall as an indefinite plate. It is established that the durability is mainly determined by the number of layers and to a lesser degree depends on the relative size of the defect for the given layer thickness. The advantage of the multilayer vessels is the possibility of selecting layer materials so that to exclude the effect of agressive corrosion media on the strength [ru

  16. Vessel size measurements in angiograms: Manual measurements

    International Nuclear Information System (INIS)

    Hoffmann, Kenneth R.; Dmochowski, Jacek; Nazareth, Daryl P.; Miskolczi, Laszlo; Nemes, Balazs; Gopal, Anant; Wang Zhou; Rudin, Stephen; Bednarek, Daniel R.

    2003-01-01

    Vessel size measurement is perhaps the most often performed quantitative analysis in diagnostic and interventional angiography. Although automated vessel sizing techniques are generally considered to have good accuracy and precision, we have observed that clinicians rarely use these techniques in standard clinical practice, choosing to indicate the edges of vessels and catheters to determine sizes and calibrate magnifications, i.e., manual measurements. Thus, we undertook an investigation of the accuracy and precision of vessel sizes calculated from manually indicated edges of vessels. Manual measurements were performed by three neuroradiologists and three physicists. Vessel sizes ranged from 0.1-3.0 mm in simulation studies and 0.3-6.4 mm in phantom studies. Simulation resolution functions had full-widths-at-half-maximum (FWHM) ranging from 0.0 to 0.5 mm. Phantom studies were performed with 4.5 in., 6 in., 9 in., and 12 in. image intensifier modes, magnification factor = 1, with and without zooming. The accuracy and reproducibility of the measurements ranged from 0.1 to 0.2 mm, depending on vessel size, resolution, and pixel size, and zoom. These results indicate that manual measurements may have accuracies comparable to automated techniques for vessels with sizes greater than 1 mm, but that automated techniques which take into account the resolution function should be used for vessels with sizes smaller than 1 mm

  17. Vessels for elevated temperature service

    International Nuclear Information System (INIS)

    O'Donnell, W.J.; Porowski, J.S.

    1983-01-01

    The subject is covered in chapters, entitled: introduction (background; elevated temperature concerns; design tools); design of pressure vessels for elevated temperature per ASME code; basic elevated temperature failure modes; allowable stresses and strains per ASME code (basic allowable stress limits; ASME code limits for bending; time-fraction summations; strain limits; buckling and instability; negligible creep and stress-rupture effects); combined membrane and bending stresses in creep regime; thermal stress cycles; bounding methods based on elastic core concept (bounds on accumulated strains; more accurate bounds; strain ranges; maximum stresses; strains at discontinuities); elastic follow-up; creep strain concentrations; time-dependent fatigue (combined creep rupture and fatigue damage; limits for inelastic design analyses; limits for elastic design analyses); flaw evaluation techniques; type 316 stainless steel; type 304 stainless steel; steel 2 1/4Cr1Mo; Inconel 718; Incolloy 800; Hastelloy X; detailed inelastic design analyses. (U.K.)

  18. Milestones in pressure vessel technology

    International Nuclear Information System (INIS)

    Spence, J.; Nash, D.H.

    2004-01-01

    The progress of pressure vessel technology over the years has been influenced by many important events. This paper identifies a number of 'milestones' which have provided a stimulus to analysis methods, manufacturing, operational processes and new pressure equipment. The formation of a milestone itself along with its subsequent development is often critically dependent on the work of many individuals. It is postulated that such developments takes place in cycles, namely, an initial idea, followed sometimes by unexpected failures, which in turn stimulate analysis or investigation, and when confidence is established, followed finally by the emergence of codes ad standards. Starting from the industrial revolution, key milestones are traced through to the present day and beyond

  19. Reactor pressure vessel thermal annealing

    International Nuclear Information System (INIS)

    Lee, A.D.

    1997-01-01

    The steel plates and/or forgings and welds in the beltline region of a reactor pressure vessel (RPV) are subject to embrittlement from neutron irradiation. This embrittlement causes the fracture toughness of the beltline materials to be less than the fracture toughness of the unirradiated material. Material properties of RPVs that have been irradiated and embrittled are recoverable through thermal annealing of the vessel. The amount of recovery primarily depends on the level of the irradiation embrittlement, the chemical composition of the steel, and the annealing temperature and time. Since annealing is an option for extending the service lives of RPVs or establishing less restrictive pressure-temperature (P-T) limits; the industry, the Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have assisted in efforts to determine the viability of thermal annealing for embrittlement recovery. General guidance for in-service annealing is provided in American Society for Testing and Materials (ASTM) Standard E 509-86. In addition, the American Society of Mechanical Engineers (ASME) Code Case N-557 addresses annealing conditions (temperature and duration), temperature monitoring, evaluation of loadings, and non-destructive examination techniques. The NRC thermal annealing rule (10 CFR 50.66) was approved by the Commission and published in the Federal Register on December 19, 1995. The Regulatory Guide on thermal annealing (RG 1.162) was processed in parallel with the rule package and was published on February 15, 1996. RG 1.162 contains a listing of issues that need to be addressed for thermal annealing of an RPV. The RG also provides alternatives for predicting re-embrittlement trends after the thermal anneal has been completed. This paper gives an overview of methodology and recent technical references that are associated with thermal annealing. Results from the DOE annealing prototype demonstration project, as well as NRC activities related to the

  20. Pressurization of Containment Vessels from Plutonium Oxide Contents

    International Nuclear Information System (INIS)

    Hensel, S.

    2012-01-01

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

  1. Single pressure vessel (SPV) nickel-hydrogen battery design

    Energy Technology Data Exchange (ETDEWEB)

    Coates, D.; Grindstaff, B.; Fox, C. [Eagle-Picher Industries, Inc., Joplin, MO (United States)

    1995-07-01

    Single pressure vessel (SPV) technology combines an entire multi-cell nickel-hydrogen (NiH{sub 2}) space battery within a single pressure vessel. SPV technology has been developed to improve the performance (volume/mass) of the NiH{sub 2} system at the battery level and ultimately to reduce overall battery cost and increase system reliability. Three distinct SPV technologies are currently under development and in production. Eagle-Picher has license to the COMSAT Laboratories technology, as well as internally developed independent SPV technology. A third technology resulted from the acquisition of Johnson Controls NiH{sub 2} battery assets in June, 1994. SPV batteries are currently being produced in 25 ampere-hour (Ah), 35 Ah and 50 Ah configurations. The battery designs have an overall outside diameter of 10 inches (25.4 centimeters).

  2. Lessons Learned From Developing Reactor Pressure Vessel Steel Embrittlement Database

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL

    2010-08-01

    Materials behaviors caused by neutron irradiation under fission and/or fusion environments can be little understood without practical examination. Easily accessible material information system with large material database using effective computers is necessary for design of nuclear materials and analyses or simulations of the phenomena. The developed Embrittlement Data Base (EDB) at ORNL is this comprehensive collection of data. EDB database contains power reactor pressure vessel surveillance data, the material test reactor data, foreign reactor data (through bilateral agreements authorized by NRC), and the fracture toughness data. The lessons learned from building EDB program and the associated database management activity regarding Material Database Design Methodology, Architecture and the Embedded QA Protocol are described in this report. The development of IAEA International Database on Reactor Pressure Vessel Materials (IDRPVM) and the comparison of EDB database and IAEA IDRPVM database are provided in the report. The recommended database QA protocol and database infrastructure are also stated in the report.

  3. Analysis of cracked pressure vessel nozzles by finite elements

    International Nuclear Information System (INIS)

    Reynen, J.

    1975-01-01

    In order to assess the safety of pressure vessel nozzles, the analysis should take into account cracks. The paper describes various algorithms, their computer implementations and relative merits to define in an effective way strain energy release rates along the tip front of arbitrary 3 D cracks under arbitary load including thermal strains. These techniques are basically equivalent to substructuring techniques and consequently they can be implemented to only FEM program able to deal with the data handling problems of the substructuring technique. Examples are given carried out with a substructure version of the BERSAFE system. These examples include a corner crack in a pressure vessel nozzle loaded by internal pressure and by thermal stresses. (Auth.)

  4. Requirements for thermal insulation on prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Neylan, A.J.; Wistrom, J.D.

    1979-01-01

    During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

  5. The coolability limits of a reactor pressure vessel lower head

    Energy Technology Data Exchange (ETDEWEB)

    Theofanous, T.G.; Syri, S. [Univ. of California, Santa Barbara, CA (United States)

    1995-09-01

    Configuration II of the ULPU experimental facility is described, and from a comprehensive set of experiments are provided. The facility affords full-scale simulations of the boiling crisis phenomenon on the hemispherical lower head of a reactor pressure vessel submerged in water, and heated internally. Whereas Configuration I experiments (published previously) established the lower limits of coolability under low submergence, pool-boiling conditions, with Configuration II we investigate coolability under conditions more appropriate to practical interest in severe accident management; that is, heat flux shapes (as functions of angular position) representative of a core melt contained by the lower head, full submergence of the reactor pressure vessel, and natural circulation. Critical heat fluxes as a function of the angular position on the lower head are reported and related the observed two-phase flow regimes.

  6. Proceedings of the Workshop on in-vessel core debris retention and coolability

    International Nuclear Information System (INIS)

    1999-01-01

    This conference on in-vessel core debris retention and coolability is composed of 37 papers grouped in three sessions: session 1 (Keynote papers: Key phenomena of late phase core melt progression, accident management strategies and status quo of severe fuel damage codes, In-vessel retention as a severe accident management scheme, GAREC analyses in support of in-vessel retention concept, Latest findings of RASPLAV project); session 2 - Experiments and model development with five sub-sessions: sub-session 1 (Debris bed heat transfer: Debris and Pool Formation/Heat Transfer in FARO-LWR: Experiments and Analyses, Evaporation and Flow of Coolant at the Bottom of a Particle-Bed modelling Relocated Debris, Investigations on the Coolability of Debris in the Lower Head with WABE-2D and MESOCO-2D, Uncertainty and Sensitivity Analysis of the Heat Transfer Mechanisms in the Lower Head, Simulation of the Arrival and Evolution of Debris in a PWR Lower Head with the SFD ICARE2 code), sub-session 2 (Corium properties, molten pool natural convection, and crust formation: Physico-chemistry and corium properties for in-vessel retention, Experimental data on heat flux distribution from volumetrically heated pool with frozen boundaries, Thermal hydraulic phenomena in corium pools - numerical simulation with TOLBIAC and experimental validation with BALI, TOLBIAC code simulations of some molten salt RASPLAV experiments, SIMECO experiments on in-vessel melt pool formation and heat transfer with and without a metallic layer, Numerical investigation of turbulent natural convection heat transfer in an internally-heated melt pool and metallic layer, Current status and validation of CON2D and 3D code, Free convection of heat-generating fluid in a constrained during experimental simulation of heat transfer in slice geometry), sub-session 3 (Gap formation and gap cooling: Quench of molten aluminum oxide associated with in-vessel debris retention by RPV internal water, Experimental investigations

  7. Emissions factors for gaseous and particulate pollutants from offshore diesel engine vessels in China

    Science.gov (United States)

    Zhang, F.; Chen, Y.; Tian, C.; Li, J.; Zhang, G.; Matthias, V.

    2015-09-01

    Shipping emissions have significant influence on atmospheric environment as well as human health, especially in coastal areas and the harbor districts. However, the contribution of shipping emissions on the environment in China still need to be clarified especially based on measurement data, with the large number ownership of vessels and the rapid developments of ports, international trade and shipbuilding industry. Pollutants in the gaseous phase (carbon monoxide, sulfur dioxide, nitrogen oxides, total volatile organic compounds) and particle phase (particulate matter, organic carbon, elemental carbon, sulfates, nitrate, ammonia, metals) in the exhaust from three different diesel engine power offshore vessels in China were measured in this study. Concentrations, fuel-based and power-based emissions factors for various operating modes as well as the impact of engine speed on emissions were determined. Observed concentrations and emissions factors for carbon monoxide, nitrogen oxides, total volatile organic compounds, and particulate matter were higher for the low engine power vessel than for the two higher engine power vessels. Fuel-based average emissions factors for all pollutants except sulfur dioxide in the low engine power engineering vessel were significantly higher than that of the previous studies, while for the two higher engine power vessels, the fuel-based average emissions factors for all pollutants were comparable to the results of the previous studies. The fuel-based average emissions factor for nitrogen oxides for the small engine power vessel was more than twice the International Maritime Organization standard, while those for the other two vessels were below the standard. Emissions factors for all three vessels were significantly different during different operating modes. Organic carbon and elemental carbon were the main components of particulate matter, while water-soluble ions and elements were present in trace amounts. Best-fit engine speeds

  8. Effect of fuel assembly mechanical design changes on dynamic response of reactor pressure vessel system under extreme loadings

    International Nuclear Information System (INIS)

    Bhandari, D.R.; Hankinson, M.F.

    1993-01-01

    This paper presents the results of a study to assess the effect of fuel assembly mechanical design changes on the dynamic response of a pressurized water reactor vessel and reactor internals under Loss-Of-Coolant Accident (LOCA) conditions. The results of this study show that the dynamic response of the reactor vessel internals and the core under extreme loadings, such as LOCA, is very sensitive to fuel assembly mechanical design changes. (author)

  9. MDCTA diagnosis of cerebral vessel disease among patients with arterial hypertension

    International Nuclear Information System (INIS)

    Romanko-Hrushchak, Nataliya

    2013-01-01

    to study changes involving cerebral vessels in patients with hypertension and various levels of total cardiovascular risk. One hundred and thirty-four patients underwent CT-angiography of intracranial vessels. Ninety-eight of them were diagnosed with hypertension. Taking into consideration high blood pressure, presence of risk factors and target organ damage subjects were divided into 4 groups: with low, medium, high and very high total cardiovascular risk. Control group included 36 patients. They were not diagnosed with hypertension at the time of examination. One hundred and five patients were examined using a 4-slice CT scanner (Toshiba Asteion 4, Toshiba Medical System, Japan), and 29 patients were examined using a 128-slice scanner (Siemens Definition AS+, Siemens Healthcare, Germany) with an injection system. We used iodine-containing contrast agents such as iodixanol and iopromide for angiography. Anatomical and topographic changes of cerebral vessels were most frequently found in hypertensive patients with high and very high total cardiovascular risk. Narrowing of vertebral vessels was the most common change (27 patients (27.55%), 21 patients (21.43%) had narrowing of the right artery, and 6 (6.12%) subjects – of the left one). Tortuous course of internal carotid arteries at the neck level was visualized in 11 patients (11.22%). Narrowing of A1 segment of anterior cerebral artery was noted in 9 patients (9.18%), of the right one – in 8 patients (8.16%), of the left one – in 1 patient (1.02%). Aneurysmal dilation of intracranial vessels was visualized in 6 patients (6.12%). Saccular aneurysm of left internal carotid artery was diagnosed in 2 patients (2.04%), one patient (1.02%) had right internal carotid artery aneurysm and one patient (1.02%) had an aneurysm of the basilar artery. the most common changes of cerebral vessels diagnosed in MDCTA among patients with hypertension included various degrees of narrowing of vertebral vessels, anterior

  10. Structural analysis of the KSTAR vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    In, Sang Ryul; Yoon, Byeong Joo [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    Structure analysis of the vacuum vessel for the KSTAR tokamak which, is in the end phase of the conceptual design have been performed. Mechanical stresses and deformations of the vessel produced by constant forces due to atmospheric pressure, dead weight, fluid pressure, etc and various transient electromagnetic forces induced during tokamak operations were calculated as well as modal characteristics and buckling properties were investigated. Influences of the temperature gradient and the constraint condition of the support on the thermal stress and deformation of the vessel were analyzed. The thermal stress due to the temperature distribution on the vessel as supplying the N{sub 2} gas of 400 deg C through poloidal channels according to the recent baking concept were calculated. No severe problem in the robustness of the vessel was found when applying the constant pressures on the vessel. However the mechanical stress due to the EM force induced by halo currents flowing on the vessel and the plasma facing components (PFCs) far exceeded the allowable limit. Some reinforcing components should be added on the boundary of the PFC support and the vessel, and that of the vessel support and the vessel. A steep temperature gradient in the vicinity of the inlet and oulet of the heating gas produced a thermal stress much higher than allowable. It is necessary to make the temperature of the vessel as uniform as possible and to develop a new support concept which is flexible enough to accommodate a thermal expansion of a few cm while sufficiently strong to resist mechanical impacts. (author). 5 refs., 41 figs., 9 tabs.

  11. Reducing catheter-related thrombosis using a risk reduction tool centered on catheter to vessel ratio.

    Science.gov (United States)

    Spencer, Timothy R; Mahoney, Keegan J

    2017-11-01

    In vascular access practices, the internal vessel size is considered important, and a catheter to vessel ratio (CVR) is recommended to assist clinicians in selecting the most appropriate-sized device for the vessel. In 2016, new practice recommendations stated that the CVR can increase from 33 to 45% of the vessels diameter. There has been evidence on larger diameter catheters and increased thrombosis risk in recent literature, while insufficient information established on what relationship to vessel size is appropriate for any intra-vascular device. Earlier references to clinical standards and guidelines did not clearly address vessel size in relation to the area consumed or external catheter diameter. The aim of this manuscript is to present catheter-related thrombosis evidence and develop a standardized process of ultrasound-guided vessel assessment, integrating CVR, Virchow's triad phenomenon and vessel health and preservation strategies, empowering an evidence-based approach to device placement. Through review, calculation and assessment on the areas of the 33 and 45% rule, a preliminary clinical tool was developed to assist clinicians make cognizant decisions when placing intravascular devices relating to target vessel size, focusing on potential reduction in catheter-related thrombosis. Increasing the understanding and utilization of CVRs will lead to a safer, more consistent approach to device placement, with potential thrombosis reduction strategies. The future of evidence-based data relies on the clinician to capture accurate vessel measurements and device-related outcomes. This will lead to a more dependable data pool, driving the relationship of catheter-related thrombosis and vascular assessment.

  12. Investigation of the failure of a reactor pressure vessel by plastic instability

    International Nuclear Information System (INIS)

    Laemmer, H.; Ritter, B.

    1994-01-01

    A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

  13. A computational algorithm addressing how vessel length might depend on vessel diameter

    Science.gov (United States)

    Jing Cai; Shuoxin Zhang; Melvin T. Tyree

    2010-01-01

    The objective of this method paper was to examine a computational algorithm that may reveal how vessel length might depend on vessel diameter within any given stem or species. The computational method requires the assumption that vessels remain approximately constant in diameter over their entire length. When this method is applied to three species or hybrids in the...

  14. Vessel classification method based on vessel behavior in the port of Rotterdam

    NARCIS (Netherlands)

    Zhou, Y.; Daamen, W.; Vellinga, T.; Hoogendoorn, S.P.

    2015-01-01

    AIS (Automatic Identification System) data have proven to be a valuable source to investigate vessel behavior. The analysis of AIS data provides a possibility to recognize vessel behavior patterns in a waterway area. Furthermore, AIS data can be used to classify vessel behavior into several

  15. Nuclear reactor with a suspended vessel

    International Nuclear Information System (INIS)

    Lemercier, Guy.

    1977-01-01

    This invention relates to a nuclear reactor with a suspended vessel and applies in particular when this is a fast reactor, the core or active part of the reactor being inside the vessel and immersed under a suitable volume of flowing liquid metal to cool it by extracting the calories released by the nuclear fission in the fuel assemblies forming this core [fr

  16. Fabrication of Separator Demonstration Facility process vessel

    International Nuclear Information System (INIS)

    Oberst, E.F.

    1985-01-01

    The process vessel system is the central element in the Separator Development Facility (SDF). It houses the two major process components, i.e., the laser-beam folding optics and the separators pods. This major subsystem is the critical-path procurement for the SDF project. Details of the vaious parts of the process vessel are given

  17. Irradiation embrittlement of reactor vessel steels

    International Nuclear Information System (INIS)

    Bros, J.

    2000-01-01

    From the historical decision of closing the Yankee Rowe NPP because of the uncertainties on the level of reactor pressure vessel neutron embrittlement, this paper reviews the technical-scientist bases of the degradation phenomena, and refers to the evolution of reactor pressure vessel radiation surveillance programs. (Author)

  18. Ultrasound periodic inspections of reactor pressure vessels

    International Nuclear Information System (INIS)

    Haniger, L.

    1980-01-01

    Two versions are described of ultrasonic equipment for periodic inspections of reactor pressure vessels. One uses the principle of exchangeable programmators with solid-state logic while the other uses programmable logic with semiconductor memories. The equipment is to be used for inspections of welded joints on the upper part of the V-1 reactor pressure vessel. (L.O.)

  19. Elastic plastic buckling of elliptical vessel heads

    International Nuclear Information System (INIS)

    Alix, M.; Roche, R.L.

    1981-08-01

    The risks of buckling of dished vessel head increase when the vessel is thin walled. This paper gives the last results on experimental tests of 3 elliptical heads and compares all the results with some empirical formula dealing with elastic and plastic buckling

  20. Analyzing Vessel Behavior Using Process Mining

    NARCIS (Netherlands)

    Maggi, F.M.; Mooij, A.J.; Aalst, W.M.P. van der

    2013-01-01

    In the maritime domain, electronic sensors such as AIS receivers and radars collect large amounts of data about the vessels in a certain geographical area. We investigate the use of process mining techniques for analyzing the behavior of the vessels based on these data. In the context of maritime