WorldWideScience

Sample records for vessel internal components

  1. 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

  2. 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)

  3. 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)

  4. 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

  5. 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

  6. 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.

  7. Integration of ITER in-vessel diagnostic components in the vacuum vessel

    International Nuclear Information System (INIS)

    Encheva, A.; Bertalot, L.; Macklin, B.; Vayakis, G.; Walker, C.

    2009-01-01

    The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement's deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence.

  8. 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.

  9. Development of design Criteria for ITER In-vessel Components

    International Nuclear Information System (INIS)

    Sannazzaro, G.; Barabash, V.; Kang, S.C.; Fernandez, E.; Kalinin, G.; Obushev, A.; Martínez, V.J.; Vázquez, I.; Fernández, F.; Guirao, J.

    2013-01-01

    Absrtract: The components located inside the ITER vacuum chamber (in-vessel components – IC), due to their specific nature and the environments they are exposed to (neutron radiation, high heat fluxes, electromagnetic forces, etc.), have specific design criteria which are, in this paper, referred as Structural Design Criteria for In-vessel Components (SDC-IC). The development of these criteria started in the very early phase of the ITER design and followed closely the criteria of the RCC-MR code. Specific rules to include the effect of neutron irradiation were implemented. In 2008 the need of an update of the SDC-IC was identified to add missing specifications, to implement improvements, to modernise rules including recent evolutions in international codes and regulations (i.e. PED). Collaboration was set up between ITER Organization (IO), European (EUDA) and Russian Federation (RFDA) Domestic Agencies to generate a new version of SDC-IC. A Peer Review Group (PRG) composed by members of the ITER Organization and all ITER Domestic Agencies and code experts was set-up to review the proposed modifications, to provide comments, contributions and recommendations

  10. 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

  11. 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

  12. 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

  13. 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

  14. ITER vacuum vessel design and electromagnetic analysis on in-vessel components

    International Nuclear Information System (INIS)

    Ioki, K.; Johnson, G.; Shimizu, K.; Williamson, D.; Iizuka, T.

    1995-01-01

    Major functional requirements for the vacuum vessel are to provide the first safety barrier and to support electromagnetic loads due to plasma disruptions and vertical displacement events, and to withstand plausible accidents without losing confinement. A double wall structure concept has been developed for the vacuum vessel due to its beneficial characteristics from the viewpoints of structural integrity and electrical continuity. An electromagnetic analysis of the blanket modules and the vacuum vessel has been performed to investigate force distributions on in-vessel components. According to the vertical displacement events (VDE) scenario, which assumes a critical q-value of 1.5, the total downward vertical force, induced by coupling between the eddy current and external fields, is about 110 MN. We have performed a stress analysis for the vacuum vessel using the VDE disruption forces acting on the blankets, and a maximum stress intensity of 112 MPa was obtained in the vicinity of the lower support of the vessel. (orig.)

  15. Assessment and selection of materials for ITER in-vessel components

    Science.gov (United States)

    Kalinin, G.; Barabash, V.; Cardella, A.; Dietz, J.; Ioki, K.; Matera, R.; Santoro, R. T.; Tivey, R.; ITER Home Teams

    2000-12-01

    During the international thermonuclear experimental reactor (ITER) engineering design activities (EDA) significant progress has been made in the selection of materials for the in-vessel components of the reactor. This progress is a result of the worldwide collaboration of material scientists and industries which focused their effort on the optimisation of material and component manufacturing and on the investigation of the most critical material properties. Austenitic stainless steels 316L(N)-IG and 316L, nickel-based alloys Inconel 718 and Inconel 625, Ti-6Al-4V alloy and two copper alloys, CuCrZr-IG and CuAl25-IG, have been proposed as reference structural materials, and ferritic steel 430, and austenitic steel 304B7 with the addition of boron have been selected for some specific parts of the ITER in-vessel components. Beryllium, tungsten and carbon fibre composites are considered as plasma facing armour materials. The data base on the properties of all these materials is critically assessed and briefly reviewed in this paper together with the justification of the material selection (e.g., effect of neutron irradiation on the mechanical properties of materials, effect of manufacturing cycle, etc.).

  16. 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)

  17. 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)

  18. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  19. 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)

  20. 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

  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. Manufacturing technology development for vacuum vessel and plasma facing components

    International Nuclear Information System (INIS)

    Laitinen, Arttu; Liimatainen, Jari; Hallila, Pentti

    2005-01-01

    Vacuum vessel and plasma facing components of the ITER construction including shield modules and primary first wall panels have great impact on the production costs and reliability of the installation. From the manufacturing technology point of view, accuracy of shape, properties of the various austenitic stainless steel/austenitic stainless steel interfaces or CuCrZr/austenitic stainless steel interfaces as well as those of the base materials are crucial for technical reliability of the construction. The current approach in plasma facing components has been utilisation of solid-HIP technology and solid-powder-HIP technology. Due to the large size of especially shield modules shape, control of the internal cavities and cooling channels is extremely demanding. This requires strict control of the raw materials and manufacturing parameters

  3. 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

  4. Materials requirements for the ITER vacuum vessel and in-vessel components - approaching the construction phase

    International Nuclear Information System (INIS)

    Barabash, V.; Ioki, K.; Pick, M.; Girard, J.P.; Merola, M.

    2007-01-01

    Full text of publication follows: The ITER activities are fully devoted toward its construction. In accordance with the ITER integrated project schedule, the procurement specifications for the manufacturing of the Vacuum Vessel should be prepared by March 2008 and the procurement specifications for the in-vessel components (first wall/blanket, divertor) by 2009. To update the design, considering design and technology evolution, the ITER Design Review has been launched. Among the various topics being discussed are the important issues related to selection of materials, material procurement, and assessment of performance during operation. The main requirements related to materials for the vacuum vessel and the in-vessel components are summarized in the paper. The specific licensing requirements are to be followed for structural materials of pressure and nuclear pressure equipment components for construction of ITER. In addition, the procurements in ITER will be done mostly 'in-kind' and it is assumed that materials for these components will be produced by different Parties. However, in accordance with the regulatory requirements and quality requirements for operation, common specifications and the general rules to fulfill these requirements are to be adopted. For some ITER components (e.g. first wall, divertor high heat flux components), the ultimate qualification of the joining technologies (Be/Cu, SS/Cu, CFC/Cu, W/Cu) is under final evaluation. Successful accomplishment of the qualification program will allow to proceed with procurements of the components for ITER. The criteria for acceptance of these components and materials after manufacturing are described and the main results will be reported. Additional materials issues, which come from the on-going manufacturing R and D program, will be also described. Finally, further materials activity during the construction phase, needs for final qualification and acceptance of materials are discussed. (authors)

  5. Design evolution and integration of the ITER in-vessel components

    International Nuclear Information System (INIS)

    Martin, A.; Calcagno, B.; Chappuis, Ph.; Daly, E.; Dellopoulos, G.; Furmanek, A.; Gicquel, S.; Heitzenroeder, P.; Jiming, Chen; Kalish, M.; Kim, D.-H.; Khomiakov, S.; Labusov, A.; Loarte, A.; Loughlin, M.; Merola, M.; Mitteau, R.; Polunovski, E.; Raffray, R.; Sadakov, S.

    2013-01-01

    Highlights: ► The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. ► A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. ► The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. ► The blanket manifold system has been redesigned to improve leak detection and localisation. ► The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. -- Abstract: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues

  6. 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.

  7. The impact of microwave stray radiation to in-vessel diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Hirsch, M.; Laqua, H. P.; Hathiramani, D.; Baldzuhn, J.; Biedermann, C.; Cardella, A.; Erckmann, V.; König, R.; Köppen, M.; Zhang, D. [Max-Planck-Institut für Plasmaphysik, Teilinstitut Greifswald, EURATOM Association, D-17489 Greifswald (Germany); Oosterbeek, J.; Brand, H. von der; Parquay, S. [Technische Universiteit Eindhoven, department Technische Natuurkunde, working group for Plasma Physics and Radiation Technology, Den Doelch 2, 5612 AZ Eindhoven (Netherlands); Jimenez, R. [Centro de Investigationes Energeticas, Medioambientales y Technológicas, Association EURATOM/CIEMAT, Avenida Complutense 22, Madrid 28040 (Spain); Collaboration: W7-X Teasm

    2014-08-21

    Microwave stray radiation resulting from unabsorbed multiple reflected ECRH / ECCD beams may cause severe heating of microwave absorbing in-vessel components such as gaskets, bellows, windows, ceramics and cable insulations. In view of long-pulse operation of WENDELSTEIN-7X the MIcrowave STray RAdiation Launch facility, MISTRAL, allows to test in-vessel components in the environment of isotropic 140 GHz microwave radiation at power load of up to 50 kW/m{sup 2} over 30 min. The results show that both, sufficient microwave shielding measures and cooling of all components are mandatory. If shielding/cooling measures of in-vessel diagnostic components are not efficient enough, the level of stray radiation may be (locally) reduced by dedicated absorbing ceramic coatings on cooled structures.

  8. 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)

  9. 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

  10. Improvement of retinal blood vessel detection using morphological component analysis.

    Science.gov (United States)

    Imani, Elaheh; Javidi, Malihe; Pourreza, Hamid-Reza

    2015-03-01

    Detection and quantitative measurement of variations in the retinal blood vessels can help diagnose several diseases including diabetic retinopathy. Intrinsic characteristics of abnormal retinal images make blood vessel detection difficult. The major problem with traditional vessel segmentation algorithms is producing false positive vessels in the presence of diabetic retinopathy lesions. To overcome this problem, a novel scheme for extracting retinal blood vessels based on morphological component analysis (MCA) algorithm is presented in this paper. MCA was developed based on sparse representation of signals. This algorithm assumes that each signal is a linear combination of several morphologically distinct components. In the proposed method, the MCA algorithm with appropriate transforms is adopted to separate vessels and lesions from each other. Afterwards, the Morlet Wavelet Transform is applied to enhance the retinal vessels. The final vessel map is obtained by adaptive thresholding. The performance of the proposed method is measured on the publicly available DRIVE and STARE datasets and compared with several state-of-the-art methods. An accuracy of 0.9523 and 0.9590 has been respectively achieved on the DRIVE and STARE datasets, which are not only greater than most methods, but are also superior to the second human observer's performance. The results show that the proposed method can achieve improved detection in abnormal retinal images and decrease false positive vessels in pathological regions compared to other methods. Also, the robustness of the method in the presence of noise is shown via experimental result. Copyright © 2015 Elsevier Ireland Ltd. All rights reserved.

  11. Development of computational methods of design by analysis for pressure vessel components

    International Nuclear Information System (INIS)

    Bao Shiyi; Zhou Yu; He Shuyan; Wu Honglin

    2005-01-01

    Stress classification is not only one of key steps when pressure vessel component is designed by analysis, but also a difficulty which puzzles engineers and designers at all times. At present, for calculating and categorizing the stress field of pressure vessel components, there are several computation methods of design by analysis such as Stress Equivalent Linearization, Two-Step Approach, Primary Structure method, Elastic Compensation method, GLOSS R-Node method and so on, that are developed and applied. Moreover, ASME code also gives an inelastic method of design by analysis for limiting gross plastic deformation only. When pressure vessel components design by analysis, sometimes there are huge differences between the calculating results for using different calculating and analysis methods mentioned above. As consequence, this is the main reason that affects wide application of design by analysis approach. Recently, a new approach, presented in the new proposal of a European Standard, CEN's unfired pressure vessel standard EN 13445-3, tries to avoid problems of stress classification by analyzing pressure vessel structure's various failure mechanisms directly based on elastic-plastic theory. In this paper, some stress classification methods mentioned above, are described briefly. And the computational methods cited in the European pressure vessel standard, such as Deviatoric Map, and nonlinear analysis methods (plastic analysis and limit analysis), are depicted compendiously. Furthermore, the characteristics of computational methods of design by analysis are summarized for selecting the proper computational method when design pressure vessel component by analysis. (authors)

  12. Alignment of in-vessel components by metrology defined adaptive machining

    International Nuclear Information System (INIS)

    Wilson, David; Bernard, Nathanaël; Mariani, Antony

    2015-01-01

    Highlights: • Advanced metrology techniques developed for large volume high density in-vessel surveys. • Virtual alignment process employed to optimize the alignment of 440 blanket modules. • Auto-geometry construct, from survey data, using CAD proximity detection and orientation logic. • HMI developed to relocate blanket modules if customization limits on interfaces are exceeded. • Data export format derived for Catia parametric models, defining customization requirements. - Abstract: The assembly of ITER will involve the precise and accurate alignment of a large number of components and assemblies in areas where access will often be severely constrained and where process efficiency will be critical. One such area is the inside of the vacuum vessel where several thousand components shall be custom machined to provide the alignment references for in-vessel systems. The paper gives an overview of the process that will be employed; to survey the interfaces for approximately 3500 components then define and execute the customization process.

  13. Alignment of in-vessel components by metrology defined adaptive machining

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, David [ITER Organization, Route de Vinon sur Verdon, CS90 046, St Paul-lez-Durance (France); Bernard, Nathanaël [G2Métric, Launaguet 31140 (France); Mariani, Antony [Spatial Alignment Ltd., Witney (United Kingdom)

    2015-10-15

    Highlights: • Advanced metrology techniques developed for large volume high density in-vessel surveys. • Virtual alignment process employed to optimize the alignment of 440 blanket modules. • Auto-geometry construct, from survey data, using CAD proximity detection and orientation logic. • HMI developed to relocate blanket modules if customization limits on interfaces are exceeded. • Data export format derived for Catia parametric models, defining customization requirements. - Abstract: The assembly of ITER will involve the precise and accurate alignment of a large number of components and assemblies in areas where access will often be severely constrained and where process efficiency will be critical. One such area is the inside of the vacuum vessel where several thousand components shall be custom machined to provide the alignment references for in-vessel systems. The paper gives an overview of the process that will be employed; to survey the interfaces for approximately 3500 components then define and execute the customization process.

  14. 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

  15. Welding of structural components and vessels

    International Nuclear Information System (INIS)

    1989-01-01

    'Welding of structural components and vessels' was chosen as the guiding topic for the 17th special conference in Munich so that current problems of this important area of application for welding engineering could be discussed in detail. The following topics were in the focus of the discussions: developments in steel, steel production and steel processing, reports on the practical application of welding in the manufacture of containers and pipes, quality assurance, product liability, safety considerations regarding creep-stressed components, problems of welding in large structures. 7 of the total number of 12 contributions were recorded separately for the data base ENERGY. (orig./MM) [de

  16. 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)

  17. 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

  18. 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)

  19. 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

  20. 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

  1. 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

  2. Feasibility of underwater welding of highly irradiated in-vessel components of boiling-water reactors: A literature review

    International Nuclear Information System (INIS)

    Lund, A.L.

    1997-11-01

    In February 1997, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research (RES), initiated a literature review to assess the state of underwater welding technology. In particular, the objective of this literature review was to evaluate the viability of underwater welding in-vessel components of boiling water reactor (BWR) in-vessel components, especially those components fabricated from stainless steels that are subjected to high neutron fluences. This assessment was requested because of the recent increased level of activity in the commercial nuclear industry to address generic issues concerning the reactor vessel and internals, especially those issues related to repair options. This literature review revealed a preponderance of general information about underwater welding technology, as a result of the active research in this field sponsored by the U.S. Navy and offshore oil and gas industry concerns. However, the literature search yielded only a limited amount of information about underwater welding of components in low-fluence areas of BWR in-vessel environments, and no information at all concerning underwater welding experiences in high-fluence environments. Research reported by the staff of the U.S. Department of Energy (DOE) Savannah River Site and researchers from the DOE fusion reactor program proved more fruitful. This research documented relevant experience concerning welding of stainless steel materials in air environments exposed to high neutron fluences. It also addressed problems with welding highly irradiated materials, and primarily attributed those problems to helium-induced cracking in the material. (Helium is produced from the neutron irradiation of boron, an impurity, and nickel.) The researchers found that the amount of helium-induced cracking could be controlled, or even eliminated, by reducing the heat input into the weld and applying a compressive stress perpendicular to the weld path

  3. 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.)

  4. 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)

  5. 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

  6. 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.

  7. 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

  8. Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Jacobs, Laurence [Georgia Inst. of Technology, Atlanta, GA (United States); Kim, Jin-Yeon [Georgia Inst. of Technology, Atlanta, GA (United States); Qu, Jisnmin [Northwestern Univ., Evanston, IL (United States); Ramuhalli, Pradeep [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wall, Joe [Electric Power Research Inst. (EPRI), Knoxville, TN (United States)

    2015-11-02

    The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor – a sensor that can continuously monitor a material’s damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks: (1

  9. Nonlinear Ultrasonic Techniques to Monitor Radiation Damage in RPV and Internal Components

    International Nuclear Information System (INIS)

    Jacobs, Laurence; Kim, Jin-Yeon; Qu, Jisnmin; Ramuhalli, Pradeep; Wall, Joe

    2015-01-01

    The objective of this research is to demonstrate that nonlinear ultrasonics (NLU) can be used to directly and quantitatively measure the remaining life in radiation damaged reactor pressure vessel (RPV) and internal components. Specific damage types to be monitored are irradiation embrittlement and irradiation assisted stress corrosion cracking (IASCC). Our vision is to develop a technique that allows operators to assess damage by making a limited number of NLU measurements in strategically selected critical reactor components during regularly scheduled outages. This measured data can then be used to determine the current condition of these key components, from which remaining useful life can be predicted. Methods to unambiguously characterize radiation related damage in reactor internals and RPVs remain elusive. NLU technology has demonstrated great potential to be used as a material sensor - a sensor that can continuously monitor a material's damage state. The physical effect being monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave. The degree of nonlinearity is quantified with the acoustic nonlinearity parameter, β, which is an absolute, measurable material constant. Recent research has demonstrated that nonlinear ultrasound can be used to characterize material state and changes in microscale characteristics such as internal stress states, precipitate formation and dislocation densities. Radiation damage reduces the fracture toughness of RPV steels and internals, and can leave them susceptible to IASCC, which may in turn limit the lifetimes of some operating reactors. The ability to characterize radiation damage in the RPV and internals will enable nuclear operators to set operation time thresholds for vessels and prescribe and schedule replacement activities for core internals. Such a capability will allow a more clear definition of reactor safety margins. The research consists of three tasks

  10. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    International Nuclear Information System (INIS)

    Lee, K. H.; Woo, H. K.; Im, K. H.; Cho, S. Y.; Kim, J. B.

    2000-01-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10 -6 ∼10 -7 Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses

  11. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K. H.; Woo, H. K. [Chungnam National Univ., Taejon (Korea, Republic of); Im, K. H.; Cho, S. Y. [korea Basic Science Institute, Taejon (Korea, Republic of); Kim, J. B. [Hyundai Heavy Industries Co., Ltd., Ulsan (Korea, Republic of)

    2000-07-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6}{approx}10{sup -7}Pa, to produce clean plasma with low impurity containments. For this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 .deg. C, 350 .deg. C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses.

  12. 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)

  13. Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

    Energy Technology Data Exchange (ETDEWEB)

    Li, C., E-mail: chuanfei.li@ipp.mpg.de; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

    2014-10-15

    Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements.

  14. Production management and quality assurance for the fabrication of the In-Vessel Components of the stellarator Wendelstein 7-X

    International Nuclear Information System (INIS)

    Li, C.; Boscary, J.; Dekorsy, N.; Junghanns, P.; Mendelevitch, B.; Peacock, A.; Pirsch, H.; Sellmeier, O.; Springer, J.; Stadler, R.; Streibl, B.

    2014-01-01

    Highlights: • Thousand parts for the divertor, first wall, cooling supply and diagnostics as W7-X In-Vessel Components. • Database building including part and assembly data, work and capacity organization, quality assurance documents. • Production management system to organize the fabrication and the associated quality assurance. • Successful use of an efficient and flexible product planning and scheduling tool for W7-X In-Vessel Components. - Abstract: The In-Vessel Components (IVC) of the stellarator Wendelstein 7-X consist of the divertor components and the first wall (FW) with their internal water cooling supply and a set of diagnostics. Due to the significant amount of different components, including many variants, a tool called Production Managing System (PMS) has been developed to organize the fabrication and the associated quality assurance. The PMS works by building a database containing the basic parts and assembly data, manufacturing and quality control plans, and available machine capacity. The creation of this database is based mainly on the parts lists, the manufacturing drawings, and details of the working flow organization. As a consequence of the learning process and technical adjustments during the design and manufacturing phase, the database needed to be permanently updated. Therefore an interface tool to optimize the data preparation has been developed. PMS has been demonstrated to be an efficient tool to support the IVC production activities providing reliable planning estimates, easily adaptable to problems encountered during the fabrication and provided a basis for the integration of quality assurance requirements

  15. The baking analysis for vacuum vessel and plasma facing components of the KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

    Lee, K.H. [Chungnam National University Graduate School, Taejeon (Korea); Im, K.H.; Cho, S.Y. [Korea Basic Science Institute, Taejeon (Korea); Kim, J.B. [Hyundai Heavy Industries Co., Ltd. (Korea); Woo, H.K. [Chungnam National University, Taejeon (Korea)

    2000-11-01

    The base pressure of vacuum vessel of the KSTAR (Korea Superconducting Tokamak Advanced Research) Tokamak is to be a ultra high vacuum, 10{sup -6} {approx} 10{sup -7} Pa, to produce clean plasma with low impurity containments. for this purpose, the KSTAR vacuum vessel and plasma facing components need to be baked up to at least 250 deg.C, 350 deg.C respectively, within 24 hours by hot nitrogen gas from a separate baking/cooling line system to remove impurities from the plasma-material interaction surfaces before plasma operation. Here by applying the implicit numerical method to the heat balance equations of the system, overall temperature distributions of the KSTAR vacuum vessel and plasma facing components are obtained during the whole baking process. The model for 2-dimensional baking analysis are segmented into 9 imaginary sectors corresponding to each plasma facing component and has up-down symmetry. Under the resulting combined loads including dead weight, baking gas pressure, vacuum pressure and thermal loads, thermal stresses in the vacuum vessel during bakeout are calculated by using the ANSYS code. It is found that the vacuum vessel and its supports are structurally rigid based on the thermal stress analyses. (author). 9 refs., 11 figs., 1 tab.

  16. A comparison between plaque-based and vessel-based measurement for plaque component using volumetric intravascular ultrasound radiofrequency data analysis.

    Science.gov (United States)

    Shin, Eun-Seok; Garcia-Garcia, Hector M; Garg, Scot; Serruys, Patrick W

    2011-04-01

    Although percent plaque components on plaque-based measurement have been used traditionally in previous studies, the impact of vessel-based measurement for percent plaque components have yet to be studied. The purpose of this study was therefore to correlate percent plaque components derived by plaque- and vessel-based measurement using intravascular ultrasound virtual histology (IVUS-VH). The patient cohort comprised of 206 patients with de novo coronary artery lesions who were imaged with IVUS-VH. Age ranged from 35 to 88 years old, and 124 patients were male. Whole pullback analysis was used to calculate plaque volume, vessel volume, and absolute and percent volumes of fibrous, fibrofatty, necrotic core, and dense calcium. The plaque and vessel volumes were well correlated (r = 0.893, P measurement was also highly correlated with vessel-based measurement. Therefore, the percent plaque component volume calculated by vessel volume could be used instead of the conventional percent plaque component volume calculated by plaque volume.

  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. Design, Analysis and R&D of the EAST In-Vessel Components

    Science.gov (United States)

    Yao, Damao; Bao, Liman; Li, Jiangang; Song, Yuntao; Chen, Wenge; Du, Shijun; Hu, Qingsheng; Wei, Jing; Xie, Han; Liu, Xufeng; Cao, Lei; Zhou, Zibo; Chen, Junling; Mao, Xinqiao; Wang, Shengming; Zhu, Ning; Weng, Peide; Wan, Yuanxi

    2008-06-01

    In-vessel components are important parts of the EAST superconducting tokamak. They include the plasma facing components, passive plates, cryo-pumps, in-vessel coils, etc. The structural design, analysis and related R&D have been completed. The divertor is designed in an up-down symmetric configuration to accommodate both double null and single null plasma operation. Passive plates are used for plasma movement control. In-vessel coils are used for the active control of plasma vertical movements. Each cryo-pump can provide an approximately 45 m3/s pumping rate at a pressure of 10-1 Pa for particle exhaust. Analysis shows that, when a plasma current of 1 MA disrupts in 3 ms, the EM loads caused by the eddy current and the halo current in a vertical displacement event (VDE) will not generate an unacceptable stress on the divertor structure. The bolted divertor thermal structure with an active cooling system can sustain a load of 2 MW/m2 up to a 60 s operation if the plasma facing surface temperature is limited to 1500 °C. Thermal testing and structural optimization testing were conducted to demonstrate the analysis results.

  19. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    Energy Technology Data Exchange (ETDEWEB)

    Urbonavičius, E., E-mail: Egidijus.Urbonavicius@lei.lt; Povilaitis, M., E-mail: Mantas.Povilaitis@lei.lt; Kontautas, A., E-mail: Aurimas.Kontautas@lei.lt

    2015-11-15

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  20. Assessment of W7-X plasma vessel pressurisation in case of LOCA taking into account in-vessel components

    International Nuclear Information System (INIS)

    Urbonavičius, E.; Povilaitis, M.; Kontautas, A.

    2015-01-01

    Highlights: • Analysis of the vacuum vessel response to the LOCA in W7-X was performed using lumped-parameter codes COCOSYS and ASTEC. • Benchmarking of the results received with two codes provides more confidence in results and helps in identification of possible important differences in the modelling. • The performed analysis answered the questions set in the installed plasma vessel venting system during overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. • Differences in time until opening the burst disk observed in ASTEC and COCOSYS results are caused by differences in heat transfer modelling. - Abstract: This paper presents the analysis of W7-X vacuum vessel response taking into account in-vessel components. A detailed analysis of the vacuum vessel response to the loss of coolant accident was performed using lumped-parameter codes COCOSYS and ASTEC. The performed analysis showed that the installed plasma vessel venting system prevents overpressure of PV in case of 40 mm diameter LOCA in “baking” mode. The performed analysis revealed differences in heat transfer modelling implemented in ASTEC and COCOSYS computer codes, which require further investigation to justify the correct approach for application to fusion facilities.

  1. 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)

  2. 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)

  3. 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)

  4. 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)

  5. FFTF thermal-hydraulic testing results affecting piping and vessel component design in LMFBR's

    International Nuclear Information System (INIS)

    Stover, R.L.; Beaver, T.R.; Chang, S.C.

    1983-01-01

    The Fast Flux Test Facility completed four years of pre-operational testing in April 1982. This paper describes thermal-hydraulic testing results from this period which impact piping and vessel component design in LMFBRs. Data discussed are piping flow oscillations, piping thermal stratification and vessel upper plenum stratification. Results from testing verified that plant design limits were met

  6. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 2. Comprehending the divertor structure

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

    2006-01-01

    Divertor is given the largest heat load in the in-vessel components of fusion machine. The functions and conditions of divertor are stated from the point of view of thermal and structural dynamics. The way of thinking of structure design of divertor of JT-60 and the ITER (International Thermonuclear Experimental Reactor) is explained. As the conditions of divertor, the materials for large heat load, heat removal, pressure boundary, control of damage, and thermal stress/strain are considered. The divertor has to be changed periodically. The materials are required the heat removal function for high heat load. CuCrZr will be used to cooling tube and heat sink, and CFC materials for the surface. The cross section of ITER, a part of divertor, heat load of divertor and other components, the thermal conductivity of CFC and metal materials, conditions of cooling water for divertor of BWR, PWR and ITER, the thermal stress produced on rod, vertical target of ITER, structure of cooling tube, distribution of temperature and critical heart flux of inner wall of cooling tube, and fatigue clack of cooling tube are shown. (S.Y.)

  7. 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

  8. 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)

  9. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    2004-01-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  10. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  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. Technical meeting on materials for in-vessel components of ITER

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.

    2000-01-01

    The Technical meeting on materials for in-vessel components of ITER was held at the ITER Joint Work Site in Garching from 31 January to 4 February. The main objectives of the meetings were: 1. to summarize the requirements, 2. to review new data, 3. to discuss in detail the R and D program and to discuss the material assessment report

  14. 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

  15. 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

  16. Design criteria for high-temperature-affected, metallic and ceramic components, and for the prestressed concrete reactor pressure vessel of future HTR systems. Final report. Vol. 1-4

    International Nuclear Information System (INIS)

    1988-08-01

    This work in five separate volumes reports on the elaboration of basic data for the formulation of design criteria for HTR components and is arranged into the four following subject areas : (1) safety-specific limiting conditions; (2) metallic components; (3) prestressed concrete reactor pressure vessels; (4) graphitic reactor internals. Under item 2, the mechanical and physical characteristics of the materials X20CrMoV 12 1, X10NiCrAlTi 32 20, and NiCr23Co12Mo are examined up to temperatures of 950deg C. Stress-strain rate laws are elaborated for description of the inelastic deformation behavior. The representation of the subject area reactor pressure vessels deals with four main topics: Prestressed concrete support structure, liner, vessel closures, thermal protection system. Quality-assurance classes are defined under item 4 for graphitic components and load levels for load categories. The material evaluation is discussed in detail (e.g. manufacturing monitoring from the raw material to the graphitization and manufacturing testing up to the acceptance test). In addition, the corrosion behavior and irradiation behavior of graphite is examined and rules for computation of stresses in irradiated and unirradiated graphitic components are elaborated. (MM) [de

  17. 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)

  18. 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)

  19. 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.

  20. 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

  1. 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

  2. VDE/disruption EM analysis for ITER in-vessel components

    International Nuclear Information System (INIS)

    Miki, N.; Ioki, K.; Ilio, F.; Kodama, T.; Chiocchio, S.; Williamson, D.; Roccella, M.; Barabaschi, P.; Sayer, R.S.

    1998-01-01

    This paper summarises the results of EM analyses for ITER in-vessel components, such as blanket modules, backplate and divertor modules. In the ITER design the following two disruption scenarios are taken into account: centered or radial disruption, and vertical displacement event (VDE). Eddy currents and forces due to plasma disruption were calculated using the 3D shell element code EDDYCUFF and the 3D solid element code EMAS. The plasma motion and current decay used in the EM analysis was supplied by 2-D axisymmetric plasma equilibrium codes, TSC and MAXFEA. (authors)

  3. 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

  4. A simple in-vessel/FW component viewing system for SST-1

    International Nuclear Information System (INIS)

    Santra, Prosenjit; Biswas, Prabal; Vasava, Kirit R.; Jaiswal, Snehal; Parekh, Tejas; Chauhan, Pradeep; Patel, Hiteshkumar; Pradhan, Subrata

    2015-01-01

    A simple compact system is being proposed for in-situ visual inspection of around 3800 First Wall (FW) graphite (armour) tiles in the vacuum vessel of SST-1 tokamak. The 2 DOF, manual driven system (permanently stationed inside vacuum vessel behind outer passive stabilizer) at top and bottom mid-plane locations consist of a rack and pinion mechanism operating a arm with a CCD camera/LED mounted on it, moving over a cam profile to cover approximately 1/8 th of the toroidal span of the vacuum vessel both at interior top/bottom locations with in the FW modules. The camera and LED light should withstand the ultrahigh vacuum conditions, prolonged baking temperatures of around 200°C along with high electromagnetic forces inside the vessel. This system can be operated remotely in-between shots from outside the VV through a linear motion feed through providing linear moment to a rack and pinion mechanism connected to the arm. This mechanism provides a better viewing of the inside FW components and vessel wall surface of tokamak with simple engineering and operational effort. Any information can be acquired from system regarding damages to FWC due to interaction with plasma as well as damage of other support structures inside VV. In comparison to more complicated and complex inspection system used in other tokamaks, this mechanism can be used for frequent in vessel visual inspection, which limits the system to be small, simple, occupying less space and custom made. This system is cheap with a minimum time for realization of the concept. The paper will present the conceptual and engineering design aspect of the in-viewing system, CAD images, its advantages and limitations, camera and LED details, data acquisition and the present status of realization of the project. (author)

  5. 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

  6. 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

  7. 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...

  8. Design and issues of the ITER in-vessel components: ITER Joint central team and home teams

    International Nuclear Information System (INIS)

    Parker, R.R.

    1998-01-01

    This paper surveys the status of the design of the in-vessel components for ITER, in particular the major components, namely the vacuum vessel, blanket and first wall, and divertor, and the interface of selected ancillary systems such as those used for RF heating and current drive, and for diagnostics. The vacuum vessel is a double-walled structure constructed from two toroidal shells joined by ribs. The space between the skins is filled with shield plates directly cooled by water. The structural material is 316 LN IG (ITER grade). Toroidal supports joining the vessel midplane ports with the TF structure limit possible differential toroidal displacements, as might occur due to seismic or vertical displacement events (VDEs). A variety of load conditions corresponding to normal and off-normal loads have been considered and in all cases peak vessel stresses are within allowables. The blanket system consists of approximately 700 modules, each weighing ∝4 t. The integrated first wall consists of a beryllium-tiled copper mat bonded to the water-cooled SS shield block. The copper mat functions as a heat sink and has imbedded in it an array of SS tubes providing water cooling. The modules are mechanically attached to a toroidal backplate. Loads due to centered disruptions are reacted via hoop stress in the backplate, whereas net vertical and horizontal loads such as those arising from VDEs are transferred through the backplate and divertor supports to the vessel. (orig.)

  9. 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

  10. 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

  11. 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.

  12. 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.

  13. 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

  14. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  15. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 1. Comprehending the vacuum vessel structure

    International Nuclear Information System (INIS)

    Onozuka, Masanori; Nakahira, Masataka

    2006-01-01

    The functions, conditions and structure of vacuum vessel using tokamak fusion machines are explained. The structural standard and code of vacuum vessel, process of vacuum vessel design, and design of ITER vacuum vessel are described. Production and maintenance of ultra high vacuum, confinement of radioactive materials, support of machines in vessel and electromagnetic force, radiation shield, plasma vertical stability, one-turn electric resistance, high temperature baking heat and remove of nuclear heat, reduce of troidal ripple, structural standard, features of safety of nuclear fusion machines, subjects of structural standard of fusion vacuum vessel, design flow of vacuum vessel, establishment of radial build, selections of materials, baking and cooling method, basic structure, structure of special parts, shield structure, and of support structure, and example of design of structure, ITER, are stated. (S.Y.)

  16. 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

  17. Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Utoh, Hiroyasu, E-mail: uto.hiroyasu@jaea.go.jp [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Takase, Haruhiko [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Sakamoto, Yoshiteru; Tobita, Kenji [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); Mori, Kazuo; Kudo, Tatsuya [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan); International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Someya, Youji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke [Japan Atomic Energy Agency, Obuchi, Rokkasho-mura, Aomori-ken 039-3212 (Japan)

    2016-02-15

    Highlights: • Conceptual design of in-vessel component including conducting shell has been investigated. • The conducting shell design for plasma vertical stability was clarified from the plasma vertical stability analysis. • The calculation results showed that the double-loop shell has the most effect on plasma vertical stability. - Abstract: In order to realize a feasible DEMO, we designed an in-vessel component including the conducting shell. The project is affiliated with the broader approach DEMO design activities and is conceptualized from a plasma vertical stability and engineering viewpoint. The dependence of the plasma vertical stability on the conducing shell parameters and the electromagnetic force at plasma disruption were investigated in numerical simulations (programmed in the 3D eddy current analysis code and a plasma position control code). The simulations assumed the actual shape and position of the vacuum vessel and in-vessel components. The plasma vertical stability was most effectively maintained by the double-loop shell.

  18. 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.

  19. 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.

  20. 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)

  1. Influence of prolonged service of steam turbines on the properties of materials of rotor and vessel components

    International Nuclear Information System (INIS)

    Anfimov, V.M.; Artamonov, V.V.; Chizhik, T.A.

    1984-01-01

    The structure and mechanical properties of steam turbine elements of 25Kh1MF, 25Kh1M1FA (rotors), 15Kh1M1FL (vessel components) steels have been investigated both in initial state and after 200 000 h operation. The structure stability and phase composition of rotor steels providing conservation of heat resistance at a required level was established. Examination of vessel components showed a decrease in the yield strength by 15-20% and durability - by 10% as compared to initial ones. The conclusion on a possible prolongation of the steam turbine service life to 200 000 h is drawn. The nominal service life equals 100 000 h

  2. 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

  3. 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...

  4. To the application of TV and optical equipment for in-service inspection of reactor vessel and primary circuit component materials

    International Nuclear Information System (INIS)

    Afonin, Eh.M.; Bachelis, I.M.; Tokarev, E.A.; Yastrebov, V.E.

    1985-01-01

    Some problems of application of TV and optical equipment for inspection of reactor vessel and primary circuit component materials are considered taking the most widespread WWER-440 type reactor as an example. The most advanrageous objects of the inspection and typical zones of equipment arrangement are shown. Methods and peculiarities of the inspection with the use of TV and optical equipment are presented. Recommendations on rational application of the equipment for the inspection of WWER-440 reactor vessel components are given

  5. Aging of metal components in US nuclear reactors

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Strosnider, J.R.

    1998-01-01

    This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

  6. Development of radiation hard components for remote maintenance

    International Nuclear Information System (INIS)

    Oka, Kiyoshi; Obara, Kenjiro; Kakudate, Satoshi; Tominaga, Ryuichiro; Akada, Tamio; Morita, Hirosuke.

    1997-01-01

    In International Thermonuclear Experimental Reactor (ITER), in-vessel remote-handling is inevitably required to assemble and maintain activated in-vessel components due to D-T operation. The components of the in-vessel remote-handling system must have sufficient radiation hardness to allow for operation under an intense gamma-ray radiation of over 30 kGy/h for periods up to more than 1,000 hours. To this end, extensive irradiation tests and quality improvements including the optimization of material composition have been conducted through the ITER R and D program in order to develop radiation hard components which satisfy radiation doses from 10 MGy to 100 MGy at the dose rate of 10 kGy/h. This paper outlines the latest status of the radiation hard component development that has been conducted as the Japan Home Team's contribution to ITER. The remote-handling components tested are categorized into either robotics, viewing systems or common components. The irradiation tests include commercial base products for screening both modified and newly developed products to improve their radiation hardness. (author)

  7. 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)

  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.

    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

  9. T-stresses for internally cracked components

    International Nuclear Information System (INIS)

    Fett, T.

    1997-12-01

    The failure of cracked components is governed by the stresses in the vicinity of the crack tip. The singular stress contribution is characterised by the stress intensity factor K, the first regular stress term is represented by the so-called T-stress. T-stress solutions for components containing an internal crack were computed by application of the Bundary Collocation Method (BCM). The results are compiled in form of tables or approximative relations. In addition a Green's function of T-stresses is proposed for internal cracks which enables to compute T-stress terms for any given stress distribution in the uncracked body. (orig.) [de

  10. Development of in-vessel neutron flux monitor equipped with microfission chambers to withstand the extreme ITER environment

    Energy Technology Data Exchange (ETDEWEB)

    Ishikawa, Masao, E-mail: ishikawa.masao@jaea.go.jp; Takeda, Keigo; Itami, Kiyoshi

    2016-11-01

    Highlights: • The in-vessel components of MFC system must withstand the extreme ITER environment. • To verify this, the thermal cycle test and the vibration tests were conducted. • Both tests were conducted under much severer conditions than ITER environment. • Soundness verification tests after the tests indicated that no problemswere found. • It is shown that the in-vessel component is sufficiently robust ITER environment. - Abstract: Via thermal cycling and vibration tests, this study aims to demonstrate that the in-vessel components of the microfission chamber (MFC) system can withstand the extreme International Thermonuclear Experimental Reactor (ITER) environment. In thermal cycle tests, the signal cable of the device was bent into a smaller radius and it was given more bends than those in its actual configuration within ITER. A faster rate of temperature change than that under the typical ITER baking scenario was then imposed on in-vessel components. For the vibration tests, strong 10 G vibrational accelerations with frequencies ranging from 30 Hz to 2000 Hz were imposed to the detector and the connector of the in-vessel components to simulate various types of electromagnetic events. Soundness verification tests of the in-vessel components conducted after thermal cycling and vibration testing indicated that problems related to the signal transmission cable functioning were not found. Thus, it was demonstrated that the in-vessel components of the MFC can withstand the extreme environment within ITER.

  11. 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)

  12. 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

  13. ITER vacuum vessel, in vessel components and plasma facing materials

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Enoeda, M.; Federici, G.

    2007-01-01

    Design of the NB ports including duct liners under heat loads of the neutral beams has been developed. Design of the in-wall shielding has been developed in more details considering the supporting structure and the assembly method. The ferromagnetic inserts have previously not been installed in the outboard midplane region due to irregularity caused by the tangential ports for NB injection. Due to this configuration, the maximum ripple is relatively large (∝1 %) in a limited region of the plasma and the toroidal field flux lines fluctuate ∝10 mm in the FW region. To avoid these problems, additional ferromagnetic inserts are to be installed in the equatorial port region. Detailed studies were carried out on the ITER vacuum vessel to define appropriate codes and standards in the context of the ITER licensing in France. A set of draft documents regarding the ITER vacuum vessel structural code were prepared including an RCC-MR Addendum for the ITER VV with justified exceptions or modifications. The main deviation from the base Code is the extensive use of UT in lieu of radiography for the volumetric examination of all one-side access welds of the outer shell and field joint. The procurement allocation of blanket modules among 6 parties was fixed and the blanket module design has progressed in cooperation with parties. Fabrication of mock-ups for prequalification testing is under way and the tests will be performed in 2007-2008. Development of new beryllium materials is progressing in China and Russia. The ITER limiters will be installed in equatorial ports at two toroidal locations. The limiter plasma-facing surface protrudes ∝8 cm from the FW during the start-up and shutdown phase. In the new limiter concept, the limiters are retracted by ∝8 cm during the plasma flat top phase. This concept gives important advantages; (i) mitigation of the particle and heat loads due to disruptions, ELMs and blobs, (ii) improvement of the power coupling with the ICRH antenna

  14. 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)

  15. Locking mechanism for in-vessel components of tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, S.; Shimizu, K.; Koizumi, K.; Tada, E.

    1992-01-01

    The locking and unlocking mechanism for in-vessel replaceable components such as blanket modules, is one of the most critical issues of the tokamak fusion reactor, since the sufficient stiffness against the enormous electromagnetic loads and the easy replaceability are required. In this paper, the authors decide that a caulking cotter joint is worth initiating the R and D from veiwpoints of an effective use of space, a replaceability, a removability of nuclear heating, and a reliability. In this approach, the cotter driving (thrusting and plucking) mechanism is a critical technology. A flexible tube concept has been developed as the driving mechanism, where the stroke and driving force are obtained by a fat shape by the hydraulic pressure. The original normal tube is subjected to the working percentage of more than several hundreds percentage (from thickness of 1.2 mm to 0.2 mm) for plastically forming the flexible tube

  16. Advanced inspection and repair techniques for primary side components

    International Nuclear Information System (INIS)

    Elm, Ralph

    1998-01-01

    The availability of nuclear power plant mainly depends on the components of the Nuclear Steam Supply System (NSSS) such as reactor pressure vessel, core internals and steam generators. The last decade has been characterized by intensive inspection and repair work on PWR steam generators. In the future, it can be expected, that the inspection of the reactor pressure vessel and the inspection and repair of its internals, in both PWR and BWR will be one of the challenges for the nuclear community. Due to this challenge, new, advanced inspection and repair techniques for the vital primary side components have been developed and applied, taking into account such issues as: use of reliable and fast inspection methods, repair of affected components instead of costly replacement, reduction of outage time compared to conventional methods, minimized radiation exposure, acceptable costs. This paper reflects on advanced inspection and repair techniques such as: Baffle Former Bolt inspection and replacement, Barrel Former Bolt inspection and replacement, Mechanized UT and visual inspection of reactor pressure vessels, Steam Generator repair by advanced sleeving technology. The techniques described have been successfully applied in nuclear power plants and improved the operation performance of the components and the NPP. (author). 6 figs

  17. 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

  18. 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.

  19. 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.

  20. 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)

  1. Issues and strategies for DEMO in-vessel component integration

    International Nuclear Information System (INIS)

    Bachmann, C.; Arbeiter, F.; Boccaccini, L.V.; Coleman, M.; Federici, G.; Fischer, U.; Kemp, R.; Maviglia, F.; Mazzone, G.; Pereslavtsev, P.; Roccella, R.; Taylor, N.; Villari, R.; Villone, F.; Wenninger, R.; You, J.-H.

    2016-01-01

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  2. Issues and strategies for DEMO in-vessel component integration

    Energy Technology Data Exchange (ETDEWEB)

    Bachmann, C., E-mail: christian.bachmann@euro-fusion.org [EUROfusion PMU, Garching (Germany); Arbeiter, F.; Boccaccini, L.V. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Coleman, M.; Federici, G. [EUROfusion PMU, Garching (Germany); Fischer, U. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Kemp, R. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Maviglia, F. [EUROfusion PMU, Garching (Germany); Mazzone, G. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Pereslavtsev, P. [Karlsruhe Institute of Technology (KIT), Karlsruhe (Germany); Roccella, R. [ITER Organization, St. Paul Lez Durance (France); Taylor, N. [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, Oxfordshire OX14 3DB (United Kingdom); Villari, R. [ENEA Dipartimento Fusione e Sicurezza Nucleare C. R. Frascati – via E. Fermi 45, 00044 Frascati, Roma (Italy); Villone, F. [ENEA-CREATE Association, DIEI, Università di Cassino e del Lazio Meridiona (Italy); Wenninger, R. [EUROfusion PMU, Garching (Germany); You, J.-H. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, Garching (Germany)

    2016-11-15

    In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

  3. Aging Management Strategy and Requirements of Pressurized Water Reactor Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jun Seog; Oh, Sung Jin; Won, Se Yol; Jeong, Sun Mi [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    The demonstration that the effects of degradation in the components of PWR internals are adequately managed is essential for maintaining a healthy fleet and ensuring the continued functionality of the reactor internals. It is also very important to determine when and where irradiation susceptibility may occur for the continued operation. This paper introduces the aging management strategies and requirements for PWR internals components and discusses effects of irradiation aging results from the functionality assessments based on the categorization of internal components. This paper introduces aging management strategies and requirements for PWR internals components. The aging management requirements for PWR internals are specified in four final component groups, which are Primary, Expansion, Existing Program and No Additional Measures. Among these groups, Primary groups include any restriction on general applicability, degradation mechanism, forward link to any Expansion components, examination method, initial examination and frequency, and examination coverage and accessibility. Expansion groups are backward link to the Primary component.

  4. 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.

  5. 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

  6. Frequency and Magnitude Analysis of the Macro-instability Related Component of the Tangential Force Affecting Radial Baffles in a Stirred Vessel

    Directory of Open Access Journals (Sweden)

    P. Hasal

    2002-01-01

    Full Text Available Experimental data obtained by measuring the tangential component of force affecting radial baffles in a flat-bottomed cylindrical mixing vessel stirred with pitched blade impellers is analysed. The maximum mean tangential force is detected at the vessel bottom. The mean force value increases somewhat with decreasing impeller off-bottom clearance and is noticeably affected by the number of impeller blades. Spectral analysis of the experimental data clearly demonstrated the presence of its macro-instability (MI related low-frequency component embedded in the total force at all values of impeller Reynolds number. The dimensionless frequency of the occurrence of the MI force component is independent of stirring speed, position along the baffle, number of impeller blades and liquid viscosity. Its mean value is about 0.074. The relative magnitude (QMI of the MI-related component of the total force is evaluated by a combination of proper orthogonal decomposition (POD and spectral analysis. Relative magnitude QMI was analysed in dependence on the frequency of the impeller revolution, the axial position of the measuring point in the vessel, the number of impeller blades, the impeller off-bottom clearance, and liquid viscosity. Higher values of QMI are observed at higher impeller off-bottom clearance height and (generally QMI decreases slightly with increasing impeller speed. The QMI value decreases in the direction from vessel bottom to liquid level. No evident difference was observed between 4 blade and 6 blade impellers. Liquid viscosity has only a marginal impact on the QMI value.

  7. 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

  8. 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

  9. 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)

  10. 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)

  11. Case study for one-piece removal method of reactor vessel of nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagane, Satoru; Kitahara, Katsumi; Yoshikawa, Seiji; Miyasaka, Yasuhiko; Fukumura, Nobuo; Nisizawa, Ichiou

    2010-01-01

    A reactor installed at the center part of the nuclear ship 'Mutsu' has been stored safely and exhibited in a reactor room building since 1996. The reactor vessel and its internals are key components because of main radioactive wastes for the reasonable decommissioning plan in the future. This report describes the one-piece removal method as the one package of the reactor vessel with its internals intact with a shipping container or additional shields. The reactor vessel package (Max.100ton) will be classified acceptable for burial at the low level radioactive waste (LLW), which will be buried at a LLW pit facility under waste disposal regulations. And also, the package will be classified as an IP-2-equivalent package according to the requirement for Shipments and Packagings. (author)

  12. 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

  13. Principal component approach in variance component estimation for international sire evaluation

    Directory of Open Access Journals (Sweden)

    Jakobsen Jette

    2011-05-01

    Full Text Available Abstract Background The dairy cattle breeding industry is a highly globalized business, which needs internationally comparable and reliable breeding values of sires. The international Bull Evaluation Service, Interbull, was established in 1983 to respond to this need. Currently, Interbull performs multiple-trait across country evaluations (MACE for several traits and breeds in dairy cattle and provides international breeding values to its member countries. Estimating parameters for MACE is challenging since the structure of datasets and conventional use of multiple-trait models easily result in over-parameterized genetic covariance matrices. The number of parameters to be estimated can be reduced by taking into account only the leading principal components of the traits considered. For MACE, this is readily implemented in a random regression model. Methods This article compares two principal component approaches to estimate variance components for MACE using real datasets. The methods tested were a REML approach that directly estimates the genetic principal components (direct PC and the so-called bottom-up REML approach (bottom-up PC, in which traits are sequentially added to the analysis and the statistically significant genetic principal components are retained. Furthermore, this article evaluates the utility of the bottom-up PC approach to determine the appropriate rank of the (covariance matrix. Results Our study demonstrates the usefulness of both approaches and shows that they can be applied to large multi-country models considering all concerned countries simultaneously. These strategies can thus replace the current practice of estimating the covariance components required through a series of analyses involving selected subsets of traits. Our results support the importance of using the appropriate rank in the genetic (covariance matrix. Using too low a rank resulted in biased parameter estimates, whereas too high a rank did not result in

  14. Real-time protection of in-vessel components in ASDEX Upgrade

    International Nuclear Information System (INIS)

    Herrmann, A.; Drube, R.; Lunt, T.; Marne, P. de

    2011-01-01

    A video real time safety system (VRT) for protection of in-vessel components was fully implemented in the machine control system (CODAC) from the 2007 experimental campaign on. The VRT is based on video cameras in contrast to infrared systems. The visible wavelength range has a smaller measurement range but is a factor 5-10 less sensitive against changes of the transmission of the optical system and the target emissivity compared to infrared systems. Up to 12 analog video channels with multiple regions of interest (ROI) are processed and monitored on each video stream. At present two safety algorithms, to detect the fraction of overheating in a ROI and hot spot detection, respectively, are implemented. The integral algorithm is preferentially used for probe or limiter protection, the hot spot algorithm for divertor protection. The VRT system is realized with ReadHawk real time operating system on a multi core Linux computer.

  15. Real-time protection of in-vessel components in ASDEX Upgrade

    Energy Technology Data Exchange (ETDEWEB)

    Herrmann, A., E-mail: albrecht.herrmann@ipp.mpg.de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Drube, R.; Lunt, T.; Marne, P. de [Max-Planck-Institut fuer Plasmaphysik, EURATOM Association, D-85748 Garching (Germany)

    2011-10-15

    A video real time safety system (VRT) for protection of in-vessel components was fully implemented in the machine control system (CODAC) from the 2007 experimental campaign on. The VRT is based on video cameras in contrast to infrared systems. The visible wavelength range has a smaller measurement range but is a factor 5-10 less sensitive against changes of the transmission of the optical system and the target emissivity compared to infrared systems. Up to 12 analog video channels with multiple regions of interest (ROI) are processed and monitored on each video stream. At present two safety algorithms, to detect the fraction of overheating in a ROI and hot spot detection, respectively, are implemented. The integral algorithm is preferentially used for probe or limiter protection, the hot spot algorithm for divertor protection. The VRT system is realized with ReadHawk real time operating system on a multi core Linux computer.

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  17. 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.

  18. 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)

  19. 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)

  20. 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

  1. 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.)

  2. Overview and status of ITER internal components

    International Nuclear Information System (INIS)

    Merola, Mario; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-01-01

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine

  3. Overview and status of ITER internal components

    Energy Technology Data Exchange (ETDEWEB)

    Merola, Mario, E-mail: mario.merola@iter.org; Escourbiac, Frederic; Raffray, René; Chappuis, Philippe; Hirai, Takeshi; Martin, Alex

    2014-10-15

    Highlights: • Manufacturing technologies for the ITER internal components have been developed. • The Blanket System successfully went through its Final Design Review in April 2013. • The decision to start operation with a Divertor with a full-W armour has been taken. - Abstract: The internal components of ITER are one of the most design and technically challenging components of the ITER machine, and include the Blanket System and the Divertor. The Blanket System successfully went through its Final Design Review in April 2013 and now it is entering into the procurement phase. The design and qualification of the Divertor with a full-tungsten armour was successfully completed and this enabled the decision in November 2013 to start operation with this material option. This paper summarizes the engineering design, the R and D, the technology qualification and procurement status of the Blanket System and of the Divertor of the ITER machine.

  4. Gamma dose from activation of internal shields in IRIS reactor.

    Science.gov (United States)

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  5. Gamma dose from activation of internal shields in IRIS reactor

    International Nuclear Information System (INIS)

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  6. 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...

  7. 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

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

    International Nuclear Information System (INIS)

    Xiong Dunshi; He Shuyan; Liu Junjie; Yu Suyuan

    1993-01-01

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

  9. 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.

  10. 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

  11. Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

  12. 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

  13. 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...

  14. 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

  15. Four whole-istic aspects of schistosome granuloma biology: fractal arrangement, internal regulation, autopoietic component and closure

    Directory of Open Access Journals (Sweden)

    HL Lenzi

    2006-10-01

    Full Text Available This paper centers on some whole-istic organizational and functional aspects of hepatic Schistosoma mansoni granuloma, which is an extremely complex system. First, it structurally develops a collagenic topology, originated bidirectionally from an inward and outward assembly of growth units. Inward growth appears to be originated from myofibroblasts derived from small portal vessel around intravascular entrapped eggs, while outward growth arises from hepatic stellate cells. The auto-assembly of the growth units defines the three-dimensional scaffold of the schistosome granulomas. The granuloma surface irregularity and its border presented fractal dimension equal to 1.58. Second, it is internally regulated by intricate networks of immuneneuroendocrine stimuli orchestrated by leptin and leptin receptors, substance P and Vasoactive intestinal peptide. Third, it can reach the population of ± 40,000 cells and presents an autopoietic component evidenced by internal proliferation (Ki-67+ Cells, and by expression of c-Kit+ Cells, leptin and leptin receptor (Ob-R, granulocyte-colony stimulating factor (G-CSF-R, and erythropoietin (Epo-R receptors. Fourth, the granulomas cells are intimately connected by pan-cadherins, occludin and connexin-43, building a state of closing (granuloma closure. In conclusion, the granuloma is characterized by transitory stages in such a way that its organized structure emerges as a global property which is greater than the sum of actions of its individual cells and extracellular matrix components.

  16. Design improvements and R and D achievements for VV and in-vessel components towards ITER construction and implications for the R and D programme

    International Nuclear Information System (INIS)

    Ioki, K.

    2002-01-01

    Procurement specifications are now being finalised for ITER components whose construction is lengthy, yet which are needed early, such as the vacuum vessel. Although the basic concept of the vacuum vessel (VV) and in-vessel components of the ITER design has stayed the same as reported at the last conference, there have been several detailed design improvements resulting from efforts to raise reliability, to improve better maintainability and to save money. One of the most important achievements in the VV R and D is demonstration of the necessary assembly tolerances. Further development of advanced methods of cutting, welding and NDT for the VV have been continued in order to refine manufacturing and improve cost and technical performance. With regard to the related FW/blanket and divertor designs, the R and D has resulted in the development of suitable technologies. Prototypes of the FW panel, the blanket shield block and the divertor components have been successfully fabricated. This paper reviews the recent progress in the design as procurement nears. (author)

  17. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  18. 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)

  19. The application of external vibration monitoring to reactors with concrete pressure vessels

    International Nuclear Information System (INIS)

    Hammill, W.J.

    1979-01-01

    The application of external vibration monitoring techniques to advanced gas cooled reactors (AGR) which have concrete pressure vessels is considered. A monitoring system for a particular AGR coolant circuit structure is developed, whose primary objective is to detect impacting of two components, although the detection of forced vibration response is also considered. Experimental results from instrumented components in the reactor and data from rig tests on full size units have been used together with a mathematical model of some elements of the transmission path in order to establish its dynamic characteristics and relate internal component vibration to externally measured signals. The application of external vibration monitoring to the external detection of the forced vibration response of an internal reactor assembly and the remote monitoring of circulator sound output is discussed. (author)

  20. 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.

  1. Components of the LWR primary circuit. Pt. 2

    International Nuclear Information System (INIS)

    1984-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 0 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

  2. AIS as key component in modern vessel traffic management and information systems

    Energy Technology Data Exchange (ETDEWEB)

    Lamers, W. [DaimlerChrysler Aerospace AG (DASA), Ulm (Germany)

    1999-07-01

    The objective of this paper is to provide information in respect to universal shipborne identification system (UAIS) as main sensor in various vessel traffic applications. The presented paper will give general information concerning AIS functionality and the standardisation process. Based on experience from recent projects and various IALA working group activities, a typical future VTMIS architectures is also presented being based on AIS as key sensor. The required key performance of AIS associated with the HW components will be described. The results from European technology study Indris are presented and discussed. Finally, a summary and conclusion from the presented material will complete the technical paper. The elaboration of this presentation has been carried out as a joint task between Mr. Andre van Berg, MDS Suedafrika und Mr. Walter Lamers, DASA Ulm. (orig.)

  3. Resource allocation planning with international components

    Science.gov (United States)

    Burke, Gene; Durham, Ralph; Leppla, Frank; Porter, David

    1993-01-01

    Dumas, Briggs, Reid and Smith (1989) describe the need for identifying mutually acceptable methodologies for developing standard agreements for the exchange of tracking time or facility use among international components. One possible starting point is the current process used at the Jet Propulsion Laboratory (JPL) in planning the use of tracking resources. While there is a significant promise of better resource utilization by international cooperative agreements, there is a serious challenge to provide convenient user participation given the separate project and network locations. Coordination among users and facility providers will require a more decentralized communication process and a wider variety of automated planning tools to help users find potential exchanges. This paper provides a framework in which international cooperation in the utilization of ground based space communication systems can be facilitated.

  4. 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

  5. In-Vessel Coil Material Failure Rate Estimates for ITER Design Use

    Energy Technology Data Exchange (ETDEWEB)

    L. C. Cadwallader

    2013-01-01

    The ITER international project design teams are working to produce an engineering design for construction of this large tokamak fusion experiment. One of the design issues is ensuring proper control of the fusion plasma. In-vessel magnet coils may be needed for plasma control, especially the control of edge localized modes (ELMs) and plasma vertical stabilization (VS). These coils will be lifetime components that reside inside the ITER vacuum vessel behind the blanket modules. As such, their reliability is an important design issue since access will be time consuming if any type of repair were necessary. The following chapters give the research results and estimates of failure rates for the coil conductor and jacket materials to be used for the in-vessel coils. Copper and CuCrZr conductors, and stainless steel and Inconel jackets are examined.

  6. 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.)

  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. 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.

  9. 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.

  10. Method for the radiographic examination of the walls or components of an essentially closed vessel, and also the provision of means for the application of the method

    International Nuclear Information System (INIS)

    1978-01-01

    Method for the radiographic examination of the wall ports or supporting components of an essentially closed vessel, whereby one brings to the side of the vessel walls or supports under examination a radiation source and, to the opposite side, a radiation sensitive film, the film being irradiated by the source and thereafter developed, characterised in that one introduces into the inside of the vessel a hollow tube at a unique distance from the wall or support component, at least one end of the hollow tube being fed out and in which the hollow tube, during the period of the examination, the irradiation source or an irradiation sensitive film is introduced. (G.C.)

  11. Investigating the cooling ability of reactor vessel head injection in the Maanshan PWR using CFD simulation

    International Nuclear Information System (INIS)

    Tseng Yungshin; Lin Chihhung; Wan Jongrong; Shih Chunkuan; Tsai, F. Peter

    2011-01-01

    In order to reduce the crack growth rate on the welding of penetration pipe, Pressurized Water Reactor (PWR) of Maanshan nuclear power plant (NPP) uses vessel head injection to cool vessel lid and control rod driving components. The injection flow from the cold leg is drained by the pressure difference between cold leg and upper internal components. In this study, 10 million meshes model with 4 sub-models have been developed to simulate the thermal-hydraulic behavior by commercial CFD program FLUENT. The results indicate that the injection nozzles can provide good cooling ability to reduce the maximum temperature for lid on the vessel head. The maximum temperature of vessel lid is about 293.81degC. Based on the simulated temperature, ASME CODE N-729-1 was further used to recount the effective degradation years (EDY) and reinspection years (RIY) factors. It demonstrates that the EDY and RIY factors are still less than 1.0. Therefore, the re-inspection period for Maanshan PWR would not be significantly affected by the miner temperature difference. (author)

  12. 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.

  13. Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

    International Nuclear Information System (INIS)

    Loeb, A.; Sokcic-Kostic, M.; Eisenmann, B.; Prechtl, E.

    2007-01-01

    This paper presents the segmentation of the in 8 meter depth of water and for cutting through super alloyed moderator vessel and of the thermal shield of the MZFR stainless steel up to 130 mm wall thickness. Depending on the research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the MZFR reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m 3 . Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (author)

  14. 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

  15. A basic study on the ITER tritium storage vessel design and components

    International Nuclear Information System (INIS)

    Chung, H. S.; Ahn, D. H.; Kim, K. R.; Yim, S. P.; Paek, S. W.; Lee, M. S.; Lee, S. H.; Shim, M. H.

    2006-01-01

    The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder mixed with Cu spheres of less than one mm diameter and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements

  16. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

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

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

  17. 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.

  18. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    International Nuclear Information System (INIS)

    Joshi, Jaydeep; Yadav, Ashish; Gangadharan, Roopesh; Prasad, Rambilas; Ulahannan, Shino; Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun

    2015-01-01

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  19. Design of vacuum vessel for Indian Test Facility (INTF) for 100 keV neutral beams

    Energy Technology Data Exchange (ETDEWEB)

    Joshi, Jaydeep, E-mail: Jaydeep.joshi@iter-india.org [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Yadav, Ashish; Gangadharan, Roopesh [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India); Prasad, Rambilas [Madan Mohan Malaviya University of Technology, Gorakhpur, Uttar Pradesh 273001 (India); Ulahannan, Shino [Airframe Aerodesigns Pvt. Ltd., HAL Airport Exit Road, Old Airport Road, Bengaluru 17 (India); Rotti, Chandramouli; Bandyopadhyay, Mainak; Chakraborty, Arun [ITER-India, Institute for Plasma Research, A29, GIDC Electronics Estate, Gandhinagar 382016, Gujarat (India)

    2015-10-15

    Highlights: • Thickness calculation and optimization for the main shell, ducts, Dishends and top lid on the main shell. • Nozzle and flange design for the port openings. • Support structure design for the main shell and ducts. • FEA validation of the INTF vessel for operational, seismic and lifting condition. - Abstract: The Indian Test Facility (INTF) vacuum vessel is designed to install a full-scale test set-up of Diagnostic Neutral Beam (DNB) [1] for the qualification of beam parameters and the behavior of beam-line components prior to installation and operation in ITER. Vacuum vessel is designed in cylindrical shape having length of ∼9 m with diameter of ∼4.5 m and has a detachable top-lid for mounting as well as removal of internal components during installation and maintenance phases. The Vessel has hemispherical dish-ends with large openings for high-voltage bushing on one side and duct on another side. Vessel is provided with openings for hydraulic, cryo, gas-feed and diagnostics. Vessel duct is composed of three segments with length ranges from 3 m to 5 m with diameter of ∼1.5 m and one vessel at the end to house the second calorimeter. The objective of this paper is to present the design and analysis of vacuum vessel, with respect to its functional and operational requirements. The design calculations are done as per ASME-BPVC SectionVIII-Div.1 and subsequently Finite Element Analysis (FEM) method has been adopted to verify the design.

  20. 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

  1. 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)

  2. 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

  3. 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

  4. Discharge of Non-Reactive Fluids from Vessels

    Directory of Open Access Journals (Sweden)

    M. Castier

    Full Text Available Abstract This paper presents simulations of discharges from pressure vessels that consistently account for non-ideal fluid behavior in all the required thermodynamic properties and individually considers all the chemical components present. The underlying assumption is that phase equilibrium occurs instantaneously inside the vessel and, thus, the dynamics of the fluid in the vessel comprises a sequence of equilibrium states. The formulation leads to a system of differential-algebraic equations in which the component mass balances and the energy balance are ordinary differential equations. The algebraic equations account for the phase equilibrium conditions inside the vessel and at the discharge point, and for sound speed calculations. The simulator allows detailed predictions of the condition inside the vessel and at the discharge point as a function of time, including the flow rate and composition of the discharge. The paper presents conceptual applications of the simulator to predict the effect of leaks from vessels containing mixtures of light gases and/or hydrocarbons and comparisons to experimental data.

  5. 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.

  6. 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

  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. 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.)

  10. Aging management of PWR reactor internals in U.S. plants

    International Nuclear Information System (INIS)

    Amberge, K.J.; Demma, A.

    2015-01-01

    This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

  11. 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.

  12. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    This white paper provides an overview and status report of the thermal-hydraulic nuclear research and development, both experimental and computational, conducted predominantly at Argonne National Laboratory. Argonne from the early 1970s through the early 1990s was the Department of Energy's (DOE's) lead lab for thermal-hydraulic development of Liquid Metal Reactors (LMRs). During the 1970s and into the mid-1980s, Argonne conducted thermal-hydraulic studies and experiments on individual reactor components supporting the Experimental Breeder Reactor-II (EBR-II), Fast Flux Test Facility (FFTF), and the Clinch River Breeder Reactor (CRBR). From the mid-1980s and into the early 1990s, Argonne conducted studies on phenomena related to forced- and natural-convection thermal buoyancy in complete in-vessel models of the General Electric (GE) Prototype Reactor Inherently Safe Module (PRISM) and Rockwell International (RI) Sodium Advanced Fast Reactor (SAFR). These two reactor initiatives involved Argonne working closely with U.S. industry and DOE. This paper describes the very important impact of thermal hydraulics dominated by thermal buoyancy forces on reactor global operation and on the behavior/performance of individual components during postulated off-normal accident events with low flow. Utilizing Argonne's LMR expertise and design knowledge is vital to the further development of safe, reliable, and high-performance LMRs. Argonne believes there remains an important need for continued research and development on thermal-hydraulic design in support of DOE's and the international community's renewed thrust for developing and demonstrating the Global Nuclear Energy Partnership (GNEP) reactor(s) and the associated Argonne Liquid Metal-Advanced Burner Reactor (LM-ABR). This white paper highlights that further understanding is needed regarding reactor design under coolant low-flow events. These safety-related events are associated with the transition

  13. 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.

  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. 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.

  16. 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.

  17. 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.

  18. 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.

  19. The Development of Key Technologies in Applications of Vessels Connected to the Internet

    Directory of Open Access Journals (Sweden)

    Zhe Tian

    2017-10-01

    Full Text Available With the development of science and technology, traffic perception, communication, information processing, artificial intelligence and the shipping information system have become important in supporting the realization of intelligent shipping transportation. Against this background, the Internet of Vessels (IoV is proposed to integrate all these advanced technologies into a platform to meet the requirements of international and regional transportations. The purpose of this paper is to analyze how to benefit from the Internet of Vessels to improve the efficiency and safety of shipping, and promote the development of world transportation. In this paper, the IoV is introduced and its main architectures are outlined. Furthermore, the characteristics of the Internet of Vessels are described. Several important applications that illustrate the interaction of the Internet of Vessels’ components are proposed. Due to the development of the Internet of Vessels still being in its primary stage, challenges and prospects are identified and addressed. Finally, the main conclusions are drawn and future research priorities are provided for reference and as professional suggestions for future researchers in this field.

  20. 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)

  1. 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.)

  2. 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

  3. Seismic proving tests on the reliability for large components and equipment of nuclear power plants

    International Nuclear Information System (INIS)

    Ohno, Tokue; Tanaka, Nagatoshi

    1988-01-01

    Since Japan has destructive earthquakes frequently, the structural reliability for large components and equipment of nuclear power plants are rigorously required. They are designed using sophisticated seismic analyses and have not yet encountered a destructive earthquake. When nuclear power plants are planned, it is very important that the general public understand the structural reliability during and after an earthquake. Seismic Proving Tests have been planned by Ministry of International Trade and Industry (Miti) to comply with public requirement in Japan. A large-scale high-performance vibration table was constructed at Tasted Engineering Laboratory of Nuclear Power Engineering Test Center (NU PEC), in order to prove the structural reliability by vibrating the test model (of full scale or close to the actual size) in the condition of a destructive earthquake. As for the test models, the following four items were selected out of large components and equipment important to the safety: Reactor Containment Vessel; Primary Coolant Loop or Primary Loop Recirculation System; Reactor Pressure Vessel; and Reactor Core Internals. Here is described a brief of the vibration table, the test method and the results of the tests on PWR Reactor Containment Vessel and BWR Primary Loop Recirculation System (author)

  4. 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

  5. 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

  6. AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

    International Nuclear Information System (INIS)

    SUBUDHI, M.; MORANTE, R.; LEE, A.D.

    2002-01-01

    The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

  7. 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

  8. 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...

  9. Smooth muscle cell recruitment to lymphatic vessels requires PDGFB and impacts vessel size but not identity.

    Science.gov (United States)

    Wang, Yixin; Jin, Yi; Mäe, Maarja Andaloussi; Zhang, Yang; Ortsäter, Henrik; Betsholtz, Christer; Mäkinen, Taija; Jakobsson, Lars

    2017-10-01

    Tissue fluid drains through blind-ended lymphatic capillaries, via smooth muscle cell (SMC)-covered collecting vessels into venous circulation. Both defective SMC recruitment to collecting vessels and ectopic recruitment to lymphatic capillaries are thought to contribute to vessel failure, leading to lymphedema. However, mechanisms controlling lymphatic SMC recruitment and its role in vessel maturation are unknown. Here, we demonstrate that platelet-derived growth factor B (PDGFB) regulates lymphatic SMC recruitment in multiple vascular beds. PDGFB is selectively expressed by lymphatic endothelial cells (LECs) of collecting vessels. LEC-specific deletion of Pdgfb prevented SMC recruitment causing dilation and failure of pulsatile contraction of collecting vessels. However, vessel remodelling and identity were unaffected. Unexpectedly, Pdgfb overexpression in LECs did not induce SMC recruitment to capillaries. This was explained by the demonstrated requirement of PDGFB extracellular matrix (ECM) retention for lymphatic SMC recruitment, and the low presence of PDGFB-binding ECM components around lymphatic capillaries. These results demonstrate the requirement of LEC-autonomous PDGFB expression and retention for SMC recruitment to lymphatic vessels, and suggest an ECM-controlled checkpoint that prevents SMC investment of capillaries, which is a common feature in lymphedematous skin. © 2017. Published by The Company of Biologists Ltd.

  10. 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

  11. Components of the LWR primary circuit. Pt. 2. Design, construction and calculation. Draft

    International Nuclear Information System (INIS)

    1995-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400 deg 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

  12. In-vessel remote maintenance of the Compact Ignition Tokamak

    International Nuclear Information System (INIS)

    Tabor, M.A.; Hager, E.R.; Creedon, R.L.; Fisher, M.V.; Atkin, S.D.

    1987-01-01

    The Compact Ignition Tokamak (CIT) is the first deuterium-tritium (D-T) fusion device that will study the physics of an ignited plasma. The ability of the tokamak vacuum vessel to be maintained remotely while under vacuum has not been fully demonstrated on previous machines, and this ability will be critical to the efficient and safe operation of ignition devices. Although manned entry into the CIT vacuum vessel will be possible during the nonactivated stages of operation, remotely automated equipment will be used to assist in initial assembly of the vessel as well as to maintain all in-vessel components once the D-T burn is achieved. Remote maintenance and operation will be routinely required for replacement of thermal protection tiles, inspection of components, leak detection, and repair welding activities. Conceptual design to support these remote maintenance activities has been integrated with the conceptual design of the in-vessel components to provide a complete and practical remote maintenance system for CIT. The primary remote assembly and maintenance operations on CIT will be accomplished through two dedicated 37- x 100-cm ports on the main toroidal vessel. Each port contains a single articulated boom manipulator (ABM), which is capable of accessing half of the torus. The proposed ABM consists of a movable carriage assembly, telescoping two-part mast, and articulated link sections. 1 ref

  13. Fabrication of nuclear ship reactor MRX model and study on inspection and maintenance of components

    International Nuclear Information System (INIS)

    Kasahara, Yoshiyuki; Nakazawa, Toshio; Kusunoki, Tsuyoshi; Takahashi, Hiroki; Yoritsune, Tsutomu.

    1997-10-01

    The MRX (Marine Reactor X) is an integral type small reactor adopting passive safety systems. As for an integral type reactor, primary system components are installed in the reactor vessel. It is therefor important to establish the appropriate procedure for construction, inspection and maintenance, dismauntling, etc., for all components in the reactor vessel as well as in the reactor containment, because inspection space is limited. To study these subjects, a one-fifth model of the MRX was fabricated and operation capabilities were studied. As a result of studies, the following results are obtained. (1) Manufacturing and installing problems of the reactor pressure vessel, the containment vessel and internal components are basically not abserved. (2) Heat transfer tube structures of the steam generator and the heat exchangers of emergency decay heat removal system and containment water cooler were not seen of any problem for fabrication. However, due consideration is required in the detailed design of supports of heat transfer tubes. (3) Further studies should be needed for designs of flange penetrations and leak countermeasures for pipes instrument cables. (4) Arrangements of equipments in the containment should be taken in consideration in detail because the space is narrow. (5) Further discussion is required for installation methods of instruments and cables. (author)

  14. Remote maintenance of in-vessel components in Tokamak Fusion Test Reactor

    International Nuclear Information System (INIS)

    Loesser, G.D.; Heitzenroeder, P.; Kungl, D.; Dylla, H.F.; Cerdan, G.

    1990-01-01

    The Tokamak Fusion Test Reactor (TFTR) will generate a total of 3 x 10 21 neutrons during its planned D-T operational period. A maintenance manipulator has been designed and tested to minimize personnel radiation during in-vessel maintenance activities. Its functions include visual inspection, first-wall tile replacement, cleaning, diagnostics calibrations and leak detection. To meet these objectives, the TFTR maintenance manipulator is required to be operable in the TFTR high vacuum environment, typically -8 torr, ( -6 Pa). Geometrically, the manipulator must extend 180 0 in either direction around the torus to assure complete coverage of the vessel first-wall. The manipulator consists of a movable carriage, and movable articulated link sections which are driven by electrical actuators. The boom has vertical load capacity of 455 kg and lateral load capacity of 46 kg. The boom can either be fitted with a general inspection arm or dextrous slave arms. The general inspection arm is designed to hold the leak detector and an inspection camera; it is capable of rotation along two axes and has a linkage system which permits motion normal to the vacuum vessel wall. All systems except the dextrous slave arms are operable in a vacuum. (author)

  15. Automated image segmentation and registration of vessel wall MRI for quantitative assessment of carotid artery vessel wall dimensions and plaque composition

    NARCIS (Netherlands)

    Klooster, Ronald van 't

    2014-01-01

    The main goal of this thesis was to develop methods for automated segmentation, registration and classification of the carotid artery vessel wall and plaque components using multi-sequence MR vessel wall images to assess atherosclerosis. First, a general introduction into atherosclerosis and

  16. BETHSY 9.1b Test Calculation with TRACE Using 3D Vessel Component

    International Nuclear Information System (INIS)

    Berar, O.; Prosek, A.

    2012-01-01

    Recently, several advanced multidimensional computational tools for simulating reactor system behaviour during real and hypothetical transient scenarios were developed. One of such advanced, best-estimate reactor systems codes is TRAC/RELAP Advanced Computational Engine (TRACE), developed by the U.S. Nuclear Regulatory Commission. The advanced TRACE comes with a graphical user interface called SNAP (Symbolic Nuclear Analysis Package). It is intended for pre- and post-processing, running codes, RELAP5 to TRACE input deck conversion, input deck database generation etc. The TRACE code is still not fully development and it will have all the capabilities of RELAP5. The purpose of the present study was therefore to assess the 3D capability of the TRACE on BETHSY 9.1b test. The TRACE input deck was semi-converted (using SNAP and manual corrections) from the RELAP5 input deck. The 3D fluid dynamics within reactor vessel was modelled and compared to 1D fluid dynamics. The 3D calculation was compared both to TRACE 1D calculation and RELAP5 calculation. Namely, the geometry used in TRACE is basically the same, what gives very good basis for the comparison of the codes. The only exception is 3D reactor vessel model in case of TRACE 3D calculation. The TRACE V5.0 Patch 1 and RELAP5/MOD3.3 Patch 4 were used for calculations. The BETHSY 9.1b test (International Standard Problem no. 27 or ISP-27) was 5.08 cm equivalent diameter cold leg break without high pressure safety injection and with delayed ultimate procedure. BETHSY facility was a 3-loop replica of a 900 MWe FRAMATOME pressurized water reactor. For better presentation of the calculated physical phenomena and processes, an animation model using SNAP was developed. In general, the TRACE 3D code calculation is in good agreement with the BETHSY 9.1b test. The TRACE 3D calculation results are as good as or better than the RELAP5 calculated results. Also, the TRACE 3D calculation is not significantly different from TRACE 1D

  17. 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

  18. 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

  19. 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

  20. 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)

  1. 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)

  2. 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...

  3. 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.

  4. Construction of the HTTR in-core components

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Jinza, K.; Miki, T.

    1996-01-01

    The reactor internals of HTTR consist of graphite and metallic core support structures and shielding blocks and are designed to support core elements and to shield neutron fluence. They also have functions to restrict by-pass flow for ensuring the core cooling performance and to maintain the temperature of metallic core support structures within their design limits. The detailed design of the HTTR core support structure was approved by the government through safety review, 1990-1991. Machining of all graphite components, which consist of about 150 large blocks, was finished in September 1994 successfully. Machining and fabricating of the metallic components were also finished in September. Prior to their installation in the reactor pressure vessel (RPV), the assembly test of actual reactor internals was performed at the works to confirm above mentioned functions. The assembly test was conducted by examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the RPV and reactor internals as well as under the core support plates with respect to structural integrity, and measuring by-pass flow rate through gaps between graphite components which may degrade core performance. The another purpose of the assembly test was to confirm the installation procedure of those components. All components were assembled at the works according to the planned procedure, and the tests were executed while assembling. As a result of the tests, measured level difference and gap width between reactor internals were negligible from core thermal and hydraulic performance point of view. Coolant flows uniformly in circumferential direction at any axial level in the RPV. By-pass flow rate was found to be suppressed sufficiently and far less than the design limit. (J.P.N.)

  5. LECOTELO - conceptual design, testings and realisation of the main vessel

    International Nuclear Information System (INIS)

    Ioan, M.; Hororoi, M.

    2013-01-01

    Lead Corrosion Testing Loop (LECOTELO) facility was conceived to assure all conditions requested by corrosion/erosion tests in pure hot lead for different materials. The main vessel will receive at least 36 different material samples; each of them must be swept on both sides by a lead flow at a very well known speed. Taking into account that the inner system of this vessel is rather complex, it is very important to know the behavior of the vessel at different speeds of the lead flow around the samples. After many simulations of different configurations of the inner components, it was obtained the best inner geometry of the flow which provides the minimum pressure loss between inlet and outlet vessel. Consequently, the design of vessel components was changed in accordance with these new results of simulations and in this moment they are in the manufacturing process. (authors)

  6. A Laser Metrology/Viewing System for ITER In-Vessel Inspection

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Menon, M.M.; Dagher, M.A.; Slotwinski, A.

    1997-10-01

    This paper identifies the requirements for a remotely operated precision laser ranging system for the International Thermonuclear Experimental Reactor. The inspection system is used for metrology and viewing, and must be capable of achieving submillimeter accuracy and operation in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field levels. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser-optic module linked through fiberoptics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic mast. Gamma irradiation up to 10 7 Gy was conducted on critical sensor components with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway

  7. A laser metrology/viewing system for ITER in-vessel inspection

    International Nuclear Information System (INIS)

    Spampinato, P.T.; Barry, R.E.; Chesser, J.B.; Herndon, J.N.; Menon, M.M.; Slotwinski, A.; Dagher, M.A.; Yuen, J.L.

    1998-01-01

    This paper identifies the requirements for the International Thermonuclear Experimental Reactor metrology and viewing system, and describes a remotely operated precision surface mapping system. A metrology system capable of achieving sub-millimeter accuracy must operate in a reactor vessel that has high gamma radiation, high vacuum, elevated temperature, and magnetic field. A coherent, frequency modulated laser radar system is under development to meet these requirements. The metrology/viewing sensor consists of a compact laser optics module linked through fiber optics to the laser source and imaging units, located outside the harsh environment. The deployment mechanism is a remotely operated telescopic-mast. Gamma irradiation to 10 7 Gy was conducted on critical sensor components at Oak Ridge National Laboratory, with no significant impact to data transmission, and analysis indicates that critical sensor components can operate in a magnetic field with certain design modifications. Plans for testing key components in a magnetic field are underway. (orig.)

  8. 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.

  9. 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)

  10. 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.)

  11. Components of the LWR primary circuit. Pt. 2. Komponenten des Primaerkreises von Leichtwasserreaktoren. T. 2

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This standard is to be applied to components made of metallic materials, operated at design temperatures of up to 673 K (400/sup 0/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.

  12. Nuclear reactors sited deep underground in steel containment vessels

    Energy Technology Data Exchange (ETDEWEB)

    Bourque, Robert [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, NM 87545 (United States)

    2006-07-01

    Although nuclear power plants are certainly very safe, they are not perceived as safe by the general populace. Also, there are concerns about overland transport of spent fuel rods and other irradiated components. It is hereby proposed that the nuclear components of nuclear power plants be placed in deep underground steel vessels with secondary coolant fed from them to turbines at or near the surface. All irradiated components, including spent fuel, would remain in the chamber indefinitely. This general concept was suggested by the late Edward Teller, generated some activity 20-25 years ago and appears to be recently reviving in interest. Previous work dealt with issues of geologic stability of underground, possibly reinforced, caverns. This paper presents another approach that makes siting independent of geology by placing the reactor components in a robust steel vessel capable of resisting full overburden pressure as well as pressures resulting from accident scenarios. Structural analysis of the two vessel concepts and approximate estimated costs are presented. This work clears the way for the extensive discussions required to evaluate the advantages of this concept. (author)

  13. Cracking at nozzle corners in the nuclear pressure vessel industry

    International Nuclear Information System (INIS)

    Smith, C.W.

    1986-01-01

    Cracks in nozzle corners at the pressure boundary of nuclear reactors have been frequently observed in service. These cracks tend to form with radial orientations with respect to the nozzle central axis and are believed to be initiated by thermal shock. However, their growth is believed to be primarily due to a steady plus a fluctuating internal pressure. Due to the impracticality of fracture testing of full-scale models, the Oak Ridge National Laboratory instituted the use of an intermediate test vessel (ITV) for use in fracture testing which had the same wall thickness and nozzle size as the prototype but significantly reduced overall length and diameter. In order to determine whether or not these ITVs could provide realistic data for full-scale reactor vessels, laboratory models of full-scale boiling water reactors and ITVs were constructed and tested. After briefly reviewing the laboratory testing and correlating results with service experience, results obtained will be used to draw some general conclusions regarding the stable growth of nonplanar cracks with curved crack fronts which are the most common precursors to fracture of pressure vessel components near junctures. Use of linear elastic fracture mechanics is made in determining stress-intensity distribution along the crack fronts

  14. 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

  15. Irradiation effects on plasma diagnostic components

    International Nuclear Information System (INIS)

    Nishitani, Takeo; Iida, Toshiyuki; Ikeda, Yujiro

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m 2 and 1 MWa/m 2 , respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  16. Irradiation effects on plasma diagnostic components

    Energy Technology Data Exchange (ETDEWEB)

    Nishitani, Takeo [ed.] [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Iida, Toshiyuki; Ikeda, Yujiro [and others

    1998-10-01

    One of the most important issues to develop the diagnostics for the experimental thermonuclear reactor such as ITER is the irradiation effects on the diagnostics components. Typical neutron flux and fluence on the first wall are 1 MW/m{sup 2} and 1 MWa/m{sup 2}, respectively for ITER. In such radiation condition, most of the present diagnostics could not survive so that those will be planed to be installed far from the vacuum vessel. However, some diagnostics sensors such as bolometers and magnetic probes still have to be install inside vessel. And many transmission components for lights, wave and electric signals are inevitable even inside vessel. As a part of this R and D program of the ITER Engineering Design Activities (EDA), we carried out the irradiation tests on the basic materials of the transmission components and in-vessel diagnostics sensors in order to identify radiation hardened materials that can be used for diagnostic systems. (J.P.N.)

  17. Reactor internals production under ideal conditions at Pensacola. [PWR type reactors

    Energy Technology Data Exchange (ETDEWEB)

    1975-12-01

    The Westinghouse factory at Pensacola, Florida, which specialises in the production of pressure vessel internal components for PWRs, is described. Its excellent manufacturing and inspection facilities, supported by careful attention to staff training and motivation, are responsible for the extremely high level of quality and continual improvement in productivity.

  18. Development of radiation hardness components for ITER remote maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Yagi, Toshiaki; Morita, Yousuke

    1998-04-01

    In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

  19. Development of radiation hardness components for ITER remote maintenance

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Oka, Kiyoshi; Ito, Akira; Yagi, Toshiaki; Morita, Yousuke

    1998-01-01

    In the ITER, in-vessel remote handling is required to assemble and maintain in-vessel components in DT operations. Since in-vessel remote handling systems must operate under intense gamma ray radiation exceeding 30 kGy/h, their components must have sufficiently high radiation hardness to allow maintenance long enough in ITER in-vessel environments. Thus, extensive radiation tests and quality improvement, including optimization of material compositions, have been conducted through the ITER R and D program to develop radiation hardness components that meet radiation doses from 10 to 100 MGy at 10 kGy/h. This paper presents the latest on radiation hardness component development conducted by the Japan Home Team as a contribution to the ITER. The remote handling components tested are categorized for use in robotic or viewing systems, or as common components. Radiation tests have been conducted on commercially available products for screening, on modified products, and on new products to improve the radiation hardness. (author)

  20. 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

  1. 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.

  2. The pressure vessel for the NSF tandem

    International Nuclear Information System (INIS)

    Jones, C.W.

    1979-04-01

    The pressure vessel is a major component of the 30 MV tandem Van de Graaff electrostatic accelerator to be used in nuclear structure research at Daresbury Laboratory. The accelerator will be capable of accelerating the full range of ions in the form of a beam. Acceleration takes place in a vertical evacuated tube (beam tube) by means of a high potential on a terminal at the central position, the terminal and beam tube assembly being supported by an insulated stack structure within the pressure vessel. Under operating conditions the vessel is filled with sulphur hexafluoride gas (SF 6 ) at high pressure which acts as an insulating medium between the centre terminal and the vessel wall. The vessel is situated inside a concrete tower which besides supporting the injector room above the vessel also acts as radiation shielding around the accelerator. The report covers: functional requirements; fundamental considerations with regard to the design and procurement; detail design; materials; manufacture; acceptance test; surface treatment; final leak test. (U.K.)

  3. Float level switch for a nuclear power plant containment vessel

    International Nuclear Information System (INIS)

    Powell, J.G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel. 1 figures

  4. Float level switch for a nuclear power plant containment vessel

    Science.gov (United States)

    Powell, James G.

    1993-01-01

    This invention is a float level switch used to sense rise or drop in water level in a containment vessel of a nuclear power plant during a loss of coolant accident. The essential components of the device are a guide tube, a reed switch inside the guide tube, a float containing a magnetic portion that activates a reed switch, and metal-sheathed, ceramic-insulated conductors connecting the reed switch to a monitoring system outside the containment vessel. Special materials and special sealing techniques prevent failure of components and allow the float level switch to be connected to a monitoring system outside the containment vessel.

  5. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  6. 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.

  7. Manipulator for testing a top-opened reactor pressure vessel

    International Nuclear Information System (INIS)

    Bauer, R.; Kastl, H.

    1991-01-01

    The design is described of a manipulator to be inserted into the inside of reactor pressure vessels opened at the top. The main components of the manipulator include a fixed column protruding into the pressure vessel and a support which is slidable on the column and carries the bearing component for the measuring, testing, inspection and repair instruments. The device includes a driving equipment for the support as well as the power supply for the sets accommodated on the support, with the aim to reduce the failure rate of the manipulator as a whole, shorten the time necessary for its assembling and thus the time of staying in the reactor pressure vessel and, at the same time, make its maintenance and operation easier. (Z.S.). 13 figs

  8. 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...

  9. A study of reactor vessel integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Hoon [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Kim, Jong Kyung; Shin, Chang Ho; Seo, Bo Kyun [Hanyang Univ., Seoul (Korea, Republic of)

    1999-02-15

    The fast neutron fluence at the Reactor Pressure Vessel(RPV) of KNGR designed for 60 years lifetime was calculated by full-scope Monte Carlo simulation for reactor vessel integrity assessment. KNGR core geometry was modeled on a three-dimensional representation of the one-sixteenth of the reactor in-vessel component. Each fuel assemblies were modeled explicitly, and each fuel pins were axially divided into 5 segments. The maximum flux of 4.3 x 10{sup 10} neutrons/cm{sup 2}. sec at the RPV was obtained by tallying neutrons crossing the beltline of inner surface of the RPV.

  10. Path Analysis of Acculturative Stress Components and Their Relationship with Depression Among International Students in China.

    Science.gov (United States)

    Liu, Yang; Chen, Xinguang; Li, Shiyue; Yu, Bin; Wang, Yan; Yan, Hong

    2016-12-01

    Acculturative stress prevents international students from adapting to the host culture, increasing their risk for depression. International students in China are a growing and at-risk population for acculturative stress and depression. With data from the International Student Health and Behaviour Survey (Yu et al., ) in China, seven acculturative stress components were detected in a previous study (Yu et al., ), including a central component (self-confidence), three distal components (value conflict, identity threat and rejection) and three proximal components (poor cultural competence, opportunity deprivation and homesickness). The current study extended the previous study to investigate the relationship between these components and depression with data also from International Student Health and Behaviour Survey. Participants were 567 students (59% male, 40.4% African, mean age = 22.75, SD = 4.11) recruited in Wuhan, China. The sample scored high on the Acculturative Stress Scale for International Students (M = 92.81, SD = 23.93) and Center for Epidemiologic Studies Short Depression Scale (M = 0.97, SD = 0.53). Acculturative stress was positively associated with depression; the association between the three distal stress components and depression was fully mediated through self-confidence, while the three proximal components had a direct effect and a self-confidence-mediated indirect effect. These findings extended the value of the previous study, highlighted the central role of self-confidence in understanding acculturative stress and depression and provided new data supporting more effective counselling for international students in China. Copyright © 2016 John Wiley & Sons, Ltd. Copyright © 2016 John Wiley & Sons, Ltd.

  11. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    International Nuclear Information System (INIS)

    Mozzillo, Rocco; Tarallo, Andrea; Marzullo, Domenico; Bachmann, Christian; Di Gironimo, Giuseppe; Mazzone, Giuseppe

    2016-01-01

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  12. Preliminary structural assessment of DEMO vacuum vessel against a vertical displacement event

    Energy Technology Data Exchange (ETDEWEB)

    Mozzillo, Rocco, E-mail: rocco.mozzillo@unina.it [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Tarallo, Andrea; Marzullo, Domenico [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Bachmann, Christian [EUROfusion PMU, Boltzmannstraße 2, 85748 Garching (Germany); Di Gironimo, Giuseppe [CREATE, University of Naples Federico II, DII, P.le Tecchio 80, 80125, Naples (Italy); Mazzone, Giuseppe [Unità Tecnica Fusione - ENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy)

    2016-11-15

    Highlights: • The paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel. • The Vacuum Vessel was checked against the VDE in combinations with the weight force of all components that the vessel shall bear. • Different configurations for the vacuum vessel supports are considered, showing that the best solution is VV supported at the lower port. • The analyses evaluated the “P damage” according to RCC-MRx code. - Abstract: This paper focuses on a preliminary structural analysis of the current concept design of DEMO vacuum vessel (VV). The VV structure is checked against a vertical load due to a Vertical Displacement Event in combination with the weight force of all components that the main vessel shall bear. Different configurations for the supports are considered. Results show that the greatest safety margins are reached when the tokamak is supported through the lower ports rather than the equatorial ports, though all analyzed configurations are compliant with RCC-MRx design rules.

  13. ITER plasma facing components

    International Nuclear Information System (INIS)

    Kuroda, T.; Vieider, G.; Akiba, M.

    1991-01-01

    This document summarizes results of the Conceptual Design Activities (1988-1990) for the International Thermonuclear Experimental Reactor (ITER) project, namely those that pertain to the plasma facing components of the reactor vessel, of which the main components are the first wall and the divertor plates. After an introduction and an executive summary, the principal functions of the plasma-facing components are delineated, i.e., (i) define the low-impurity region within which the plasma is produced, (ii) absorb the electromagnetic radiation and charged-particle flux from the plasma, and (iii) protect the blanket/shield components from the plasma. A list of critical design issues for the divertor plates and the first wall is given, followed by discussions of the divertor plate design (including the issues of material selection, erosion lifetime, design concepts, thermal and mechanical analysis, operating limits and overall lifetime, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, and advanced divertor concepts) and the first wall design (armor material and design, erosion lifetime, overall design concepts, thermal and mechanical analysis, lifetime and operating limits, tritium inventory, baking and conditioning, safety analysis, manufacture and testing, an alternative first wall design, and the limiters used instead of the divertor plates during start-up). Refs, figs and tabs

  14. 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

  15. 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

  16. DSTAR: A comprehensive tokamak resistive disruption model for vacuum vessel components

    International Nuclear Information System (INIS)

    Merrill, B.J.; Jardin, S.C.

    1987-01-01

    A computer code, DSTAR, has recently been developed to quantify the surface erosion and induced forces than can occur during major tokamak plasma disruptions. A disruption analysis has been performed for the TFCX fusion device. The limiters and inboard first wall were assumed to be clad with beryllium. Disruption simulations were performed with and without these structures present, to determine their electromagnetic influence. The results with structure show that the ablated wall material is transported poloidally, as well as radially, in the plasma causing the outermost regions of the plasma to cool. The plasma moves downward and deforms while maintaining contact with the lower limiter. This motion maintains the peak impurity radiant source directly above the exposed surface. For the disruption simulation without the vacuum vessel included, the plasma moves radially along the lower limiter until it contacts the inboard wall, causing ablation of this surface as well. The conclusion is drawn that disruption simulations that do not include both the thermal and electromagnetic response of the vaccum vessel will not result in an accurate prediction. (orig.)

  17. Distribution of the In-Vessel Diagnostics in ITER Tokamak

    Energy Technology Data Exchange (ETDEWEB)

    González, Jorge, E-mail: Jorge.Gonzalez@iter.org [Rüecker Lypsa, Carretera del Prat, 65, Cornellá de Llobregat (Spain); Clough, Matthew; Martin, Alex; Woods, Nick; Suarez, Alejandro [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France); Martinez, Gonzalo [Technical University Of Catalonia (UPC), Barcelona-Tech, Barcelona (Spain); Stefan, Gicquel; Yunxing, Ma [ITER Organization, Route de Vinon sur Verdon-CS 90 046 13067 Saint Paul Lez Durance (France)

    2017-01-15

    The ITER In-Vessel Diagnostics have been distributed around the In-Vessel shell to understand burning plasma physics and assist in machine operation. Each diagnostics component has its own requirements, constraints, and even exclusion among them for the highly complex In-Vessel environment. The size of the plasma, the requirement to be able to align the blanket system to the magnetic centre of the machine, the cooling requirements of the blanket system and the size of the pressure vessel itself all add to the difficulties of integrating these systems into the remaining space available. The available space for the cables inside the special trays (in-Vessel looms) is another constraint to allocate In-Vessel electrical sensors. Besides this, there are issues with the Assembly sequences and surface & volumetric neutron heating considerations that have imposed several additional restrictions.

  18. 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

  19. International cooperation in production inspections

    International Nuclear Information System (INIS)

    Limousin, S.

    2009-01-01

    Nuclear pressure equipment, like the reactor pressure vessel or steam generators, are manufactured in many countries all around the world. As only few reactors were built in the 90's, most of the nuclear safety authorities have lost part of their know how in component manufacturing oversight. For these two reasons, vendor inspection is a key area for international cooperation. On the one hand, ASN has bilateral relationships with several countries (USA, Finland, China...) to fulfill specific purposes. On the other hand, ASN participates in international groups like the MDEP ( Multinational Design Evaluation Program). A MDEP working group dedicated to vendor inspection cooperation enables exchanges of informations (inspection program plan, inspection findings...) among the regulators. Join inspections are organized. International cooperation could lead in the long term to an harmonization of regulatory practices. (author)

  20. FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

    Directory of Open Access Journals (Sweden)

    Thiollay Nicolas

    2016-01-01

    Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

  1. 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.

  2. 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)

  3. Analysis of irradiation creep and the structural integrity of fusion in-vessel components

    International Nuclear Information System (INIS)

    Karditsas, Panayiotis J.

    2000-01-01

    This paper presents a brief review of the irradiation creep mechanism, analyses of the effect on the performance and behaviour of fusion in-vessel components, and discusses procedures for the estimation of in-service time (or lifetime) of components under combined creep-fatigue. The irradiation creep models and proposed theories are examined and analysed to produce a creep law relevant to fusion conditions. The necessary material data, constitutive equations and other parameters needed for estimation of in-service time from the combination of creep and fatigue damage are identified. Wherever possible, design curves are proposed for stress and strain. Time dependent non-linear elastoplastic example calculations are performed, for a typical first wall structure under power plant loading conditions, assuming austenitic and martensitic steel as structural materials, including material irradiation creep. The results of calculations for the stress and strain history of the first wall are used together with the proposed cumulative damage expressions derived in this study to estimate the in-service time, including the effects of stress relaxation due to creep, reduction of ductility (or fracture strain) and helium-to-displacement-damage ratio. The calculations give a displacement damage of ∼70 dpa for the 316 austenitic steel and ∼110-130 dpa for the martensitic steel. Provided there are no power transients, for a design strain of 0.5%, the in-service time is estimated to be ∼3 years for the 316 steel case (at 2.2 MW/m 2 wall load) and the high wall loading martensitic steel (5.0 MW/m 2 case), and ∼5.3 years for the martensitic steel at lower wall load (2.2 MW/m 2 case). The difficulty in defending these results lies in the uncertainty arising from the limited database and experience of the material properties, especially the creep constitutive law, when exposed to fusion environments

  4. 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)

  5. 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)

  6. 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

  7. Comparison of implant component fractures in external and internal type: A 12-year retrospective study.

    Science.gov (United States)

    Yi, Yuseung; Koak, Jai-Young; Kim, Seong-Kyun; Lee, Shin-Jae; Heo, Seong-Joo

    2018-04-01

    The aim of this study was to compare the fracture of implant component behavior of external and internal type of implants to suggest directions for successful implant treatment. Data were collected from the clinical records of all patients who received WARANTEC implants at Seoul National University Dental Hospital from February 2002 to January 2014 for 12 years. Total number of implants was 1,289 and an average of 3.2 implants was installed per patient. Information about abutment connection type, implant locations, platform sizes was collected with presence of implant component fractures and their managements. SPSS statistics software (version 24.0, IBM) was used for the statistical analysis. Overall fracture was significantly more frequent in internal type. The most frequently fractured component was abutment in internal type implants, and screw fracture occurred most frequently in external type. Analyzing by fractured components, screw fracture was the most frequent in the maxillary anterior region and the most abutment fracture occurred in the maxillary posterior region and screw fractures occurred more frequently in NP (narrow platform) and abutment fractures occurred more frequently in RP (regular platform). In external type, screw fracture occurred most frequently, especially in the maxillary anterior region, and in internal type, abutment fracture occurred frequently in the posterior region. placement of an external type implant rather than an internal type is recommended for the posterior region where abutment fractures frequently occur.

  8. 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

  9. 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

  10. Assessment and Management of ageing of major nuclear power plant components important to safety: PWR pressure vessels

    International Nuclear Information System (INIS)

    1999-10-01

    ageing management and economic planning. The target audience of the reports consists of technical experts from NPPs and from regulatory, plant design, manufacturing and technical support organizations dealing with specific plant components addressed in the reports. The NPP component addressed in the present publication is the PWR pressure vessel

  11. International feedback experience on the cutting of reactor internal components

    International Nuclear Information System (INIS)

    Boucau, J.

    2014-01-01

    Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

  12. Problems in manufacturing and transport of pressure vessels of integral reactors

    International Nuclear Information System (INIS)

    Kralovec, J.

    1997-01-01

    Integral water-cooled reactors are typical with eliminating large-diameter primary pipes and placing primary components, i.e. steam generators and pressurizers in reactor vessels. This arrangement leads to reactor pressure vessels of large dimensions: diameters, heights and thick walls and subsequently to great weights. Thus, even medium power units have pressure vessels which are on the very limit of present manufacturing capabilities. Principal manufacturing and inspection operations as well as pertinent equipment are concerned: welding, cladding, heat treatment, machining, shop-handling, non-destructive testing, hydraulic pressure tests etc. Tile transport of such a large and heavy component makes a problem which effects its design as well as the selection of the plant site. Railway, road and ship are possible ways of transport each of them having its advantages and limitations. Specific features and limits of the manufacture and transport of large pressure vessels are discussed in the paper. (author)

  13. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  14. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  15. 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

  16. Confinement Vessel Assay System: Calibration and Certification Report

    Energy Technology Data Exchange (ETDEWEB)

    Frame, Katherine C. [Los Alamos National Laboratory; Bourne, Mark M. [Los Alamos National Laboratory; Crooks, William J. [Los Alamos National Laboratory; Evans, Louise [Los Alamos National Laboratory; Gomez, Cipriano [Retired CMR-OPS: OPERATIONS; Mayo, Douglas R. [Los Alamos National Laboratory; Miko, David K. [Los Alamos National Laboratory; Salazar, William R. [Los Alamos National Laboratory; Stange, Sy [Los Alamos National Laboratory; Vigil, Georgiana M. [Los Alamos National Laboratory

    2012-07-17

    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 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of 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 {sup 239}Pu equivalent in a vessel for safeguards termination. The system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  17. Confinement Vessel Assay System: Calibration and Certification Report

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Gomez, Cipriano; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; 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 to 2 inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the vessels. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of 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 system was calibrated in three different mass regions (low, medium, and high) to cover the entire plutonium mass range that will be assayed. The low mass calibration and medium mass calibration were verified for material positioned in the center of an empty vessel. The systematic uncertainty due to position bias was estimated using an MCNPX model to simulate the response of the system to material localized at various points along the inner surface of the vessel. The background component due to cosmic ray spallation was determined by performing measurements of an empty vessel and comparing to measurements in the same location with no vessel present. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements of CVs before and after cleanout.

  18. 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

  19. 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

  20. Design standard issues for ITER in-vessel components

    International Nuclear Information System (INIS)

    Majumdar, S.

    1994-01-01

    Unique requirements that must be addressed by a structural design code for the ITER have been summarized. Existing codes such as ASME Section III, or the French RCC-MR were developed primarily for fission reactor out-of-core components and are not directly applicable to the ITER. They may be used either as a guide for developing a design code for the ITER or as interim standards. However, new rules will be needed for handling the irradiation-induced embrittlement problems faced by the ITER blanket components. Design standards developed in the past for the design of fission reactor core components in the United States can be used as guides in this area

  1. Some aspects of reactor pressure vessel integrity

    International Nuclear Information System (INIS)

    Korosec, D.; Vojvodic, G.J.

    1996-01-01

    Reactor pressure vessel of the pressurized water reactor nuclear power plant is the subject of extreme interest due to the fact that presents the pressure boundary of the reactor coolant system, which is under extreme thermal, mechanical and irradiation effects. Reactor pressure vessel by itself prevents the release of fission products to the environment. Design, construction and in-service inspection of such component is governed by strict ASME rules and other forms of administrative control. The reactor pressure vessel in nuclear power plant Kriko is designed and constructed in accordance with related ASME rules. The in-service inspection program includes all requests presented in ASME Code section XI. In the present article all major requests for the periodic inspections of reactor pressure vessel and fracture mechanics analysis are discussed. Detailed and strict fulfillment of all prescribed provisions guarantee the appropriate level of nuclear safety. (author)

  2. Implication of irradiation effects on materials data for the design of near core components

    International Nuclear Information System (INIS)

    Dietz, W.; Breitling, H.

    1995-01-01

    For LWR's strict regulations exist for the consideration of irradiation in the design and surveillance of the reactor pressure vessel in the various codes (ASME, RCC-M, KTA) but less for near core components. For FBR's no firm rules exist either for the vessel nor the reactor internals. In this paper the German design practices for the loop type SNR-300 will be presented, and also some information from the surveillance programme of the KNK-reactor. Austenitic stainless steels have been mainly selected for the near core components. For some special applications Ni-alloys and a stabilized 2 1/4 Cr 1 Mo-alloy were specified. Considerations of the irradiation effects on material properties will be made for the various temperature and fluence levels around the core. The surveillance programmes will be described. Both, the consideration of irradiation effects in the elastic and inelastic analysis and the surveillance programmes had been a part of the licensing process for SNR-300. (author). 8 figs, 4 tabs

  3. Vessel network detection using contour evolution and color components

    Energy Technology Data Exchange (ETDEWEB)

    Ushizima, Daniela; Medeiros, Fatima; Cuadros, Jorge; Martins, Charles

    2011-06-22

    Automated retinal screening relies on vasculature segmentation before the identification of other anatomical structures of the retina. Vasculature extraction can also be input to image quality ranking, neovascularization detection and image registration, among other applications. There is an extensive literature related to this problem, often excluding the inherent heterogeneity of ophthalmic clinical images. The contribution of this paper relies on an algorithm using front propagation to segment the vessel network. The algorithm includes a penalty in the wait queue on the fast marching heap to minimize leakage of the evolving interface. The method requires no manual labeling, a minimum number of parameters and it is capable of segmenting color ocular fundus images in real scenarios, where multi-ethnicity and brightness variations are parts of the problem.

  4. In-service inspection robot for PFBR main vessel- concept

    Energy Technology Data Exchange (ETDEWEB)

    Rajendran, S; Ramakumar, M S [Bhabha Atomic Research Centre, Mumbai (India). Div. of Remote Handling and Robotics

    1994-12-31

    In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs.

  5. In-service inspection robot for PFBR main vessel- concept

    International Nuclear Information System (INIS)

    Rajendran, S.; Ramakumar, M.S.

    1994-01-01

    In-service inspection (ISI) of critical components in a nuclear reactor is one of the foremost and important tasks which reveals the state of health of the system, thereby ensuring the safety of the plant, personnel and environment. Prototype Fast Breeder Reactor (PFBR) is designed as a pool type reactor. A safety vessel is provided in the design which envelopes the main reactor vessel. The ISI of the main vessel is mandatory and will be carried out by a robot which will operate on this annular gap. The design of the robot is such that it can crawl around the vessel and into the gap at the bottom of the vessel relying on friction grip. The mobile robot will carry a CCTV camera and the inspection technique packages into the interspace, position and orient these to carry out the ISI of the main vessel. The paper discusses about the design features of the robot including the gripping mechanism and the crawling sequence to perform ISI of the reactor vessel. 3 figs

  6. A Study on Components of Internal Control-Based Administrative System in Secondary Schools

    Science.gov (United States)

    Montri, Paitoon; Sirisuth, Chaiyuth; Lammana, Preeda

    2015-01-01

    The aim of this study was to study the components of the internal control-based administrative system in secondary schools, and make a Confirmatory Factor Analysis (CFA) to confirm the goodness of fit of empirical data and component model that resulted from the CFA. The study consisted of three steps: 1) studying of principles, ideas, and theories…

  7. Study on in-vessel thermohydraulics phenomena of sodium-cooled fast reactors. 1. Numerical investigation for the rationalization of hydrodynamics in the upper plenum

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamaguchi, Akira

    2002-02-01

    A large-scale sodium-cooled fast breeder reactor in feasibility studies on commercialized fast reactors has a tendency of consideration of thorough simplified and compacted system designs to realize drastic economical improvements. Therefore, special attention should be paid to thermohydraulic designs for a gas entrainment behavior from free surfaces, a flow-induced vibration of in-vessel components, a thermal shock for various structures due to high-speed coolant flows, nonsymmetrical coolant flows, etc. in the reactor vessel. In-vessel thermohydraulic analyses were carried out using a multi-dimensional code AQUA to understand the thermohydraulic characteristics in the upper plenum, and to investigate suitable in-vessel structure for the elimination of gas entrainment possibility. From the analysis, the following results were obtained. (1) It is difficult to rationale in-vessel flow patterns through adjustments of porous ratio and pressure loss for a hold down plate and baffle plates installed in an upper core structure. (2) Dummy plug insertion to a slit of the upper core structure is one of effective measures to stabilize in-vessel flow patterns. (3) Flow guide devices such as a baffle ring and a partial inner barrel are also effective measures to eliminate impinging jet to a dipped plate (D/P) and to reduce horizontal flow velocity components at free surface. (4) Installations of labyrinth structures to a R/V - D/P gap is successful for decreasing of free surface horizontal flows. (5) Gap closing of an in-vessel fuel pot and two cold trap components has the effects of reductions for free surface horizontal flows and for the difference of free surface levels. Following future investigations are important preventive measures against the gas entrainment from the free surface. (1) Flattening of spatial axial velocity distributions at the R/V - D/P gap. (2) Alleviation measures of vortex concentration at free surface. (3) Separation measures of 3-dimensional vortex

  8. 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.

  9. Development of a Remote Handling Robot for the Maintenance of an ITER-Like D-Shaped Vessel

    Directory of Open Access Journals (Sweden)

    Peihua Chen

    2014-01-01

    Full Text Available Robotic operation is one of the major challenges in the remote maintenance of ITER vacuum vessel (VV and future fusion reactors as inner operations of Tokamak have to be done by robots due to the internal adverse conditions. This paper introduces a novel remote handling robot (RHR for the maintenance of ITER-like D-shaped vessel. The modular designed RHR, which is an important part of the remote handling system for ITER, consists of three parts: an omnidirectional transfer vehicle (OTV, a planar articulated arm (PAA, and an articulated teleoperated manipulator (ATM. The task of RHR is to carry processing tools, such as the viewing system, leakage detector, and electric screwdriver, to inspect and maintain the components installed inside the D-shaped vessel. The kinematics of the OTV, as well as the kinematic analyses of the PAA and ATM, is studied in this paper. Because of its special length and heavy payload, the dynamics of the PAA is also investigated through a dynamic simulation system based on robot technology middleware (RTM. The results of the path planning, workspace simulations, and dynamic simulation indicate that the RHR has good mobility together with satisfying kinematic and dynamic performances and can well accomplish its maintenance tasks in the ITER-like D-shaped vessel.

  10. Assessment and management of ageing of major nuclear power plant components important to safety: BWR pressure vessels

    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. This report addresses the reactor pressure vessel (RPV) in BWRs. Maintaining the structural integrity of this RPV throughout NPP service life in spite of several ageing mechanisms is essential for plant safety

  11. Variability of mechanical properties of nuclear pressure vessel steels

    International Nuclear Information System (INIS)

    Petrequin, P.; Soulat, P.

    1980-01-01

    Causes of variability of mechanical properties nuclear pressure vessel steels are reviewed and discussed. The effects of product shape and size, processing history and heat treatment are investigated. Some quantitative informations are given on the scatter of mechanical properties of typical pressure vessel components. The necessity of using recommended or standardized properties for comparing mechanical properties before and after irradiation in pin pointed. (orig.) [de

  12. High-Temperature Gas-cooled Reactor steam-cycle/cogeneration lead plant reactor vessel: system design description

    International Nuclear Information System (INIS)

    1983-01-01

    The Reactor Vessel System contains the primary coolant inventory within a gas-tight pressure boundary, and provides the necessary flow paths and overpressure protection for this pressure boundary. The Reactor Vessel System also houses the components of the Reactor System, the Heat Transport System, and the Auxiliary Heat Removal System. The scope of the Reactor Vessel System includes the prestressed concrete reactor vessel (PCRV) structure with its reinforcing steel and prestressing components; liners, penetrations, closures, and cooling water tubes attached to the concrete side of the liner; the thermal barrier (insulation) on the primary coolant side of the liner; instrumentation for structural monitoring; and a pressure relief system. Specifications are presented

  13. 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

  14. 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.

  15. Acceptance test for graphite components and construction status of HTTR

    International Nuclear Information System (INIS)

    Iyoku, T.; Ishihara, M.; Maruyama, S.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    In March, 1991, the Japan Atomic Energy Research Institute (JAERI) started to constructed the High Temperature engineering Test Reactor(HTTR) which is a 30-MW(thermal) helium gas-cooled reactor with a core composed of prismatic graphite blocks piled on the core support graphite structures. Two types of graphite materials are used in the HTTR. One is the garde IG-110, isotropic fine grain graphite, another is the grade PGX, medium-to-fine grained molded graphite. These materials were selected on the basis of the appropriate properties required by the HTTR reactor design. Industry-wide standards for an acceptance test of graphite materials used as main components of a nuclear reactor had not been established. The acceptance standard for graphite components of the HTTR, therefore, was drafted by JAERI and reviewed by specialists outside JAERI. The acceptance standard consists of the material testing, non-destructive examination such as the ultrasonic and eddy current testings, dimensional and visual inspections and assembly test. Ultrasonic and eddy current testings are applied to graphite logs to detect an internal flaw and to graphite components to detect a surface flaw, respectively. The assembly test is performed at the works, prior to their installation in the reactor pressure vessel, to examine fabricating precision of each component and alignment of piled-up structures. The graphite components of the HTTR had been tested on the basis of the acceptance standard. It was confirmed that the graphite manufacturing process was well controlled and high quality graphite components were provided to the HTTR. All graphite components except for the fuel graphite blocks are to be installed in the reactor pressure vessel of the HTTR in September 1995. The paper describes the construction status of the HTTR focusing on the graphite components. The acceptance test results are also presented in this paper. (author). Figs

  16. 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.

  17. 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.

  18. Fatigue and fracture mechanics in pressure vessels and piping. PVP-Volume 304

    International Nuclear Information System (INIS)

    Mehta, H.S.; Wilkowski, G.; Takezono, S.; Bloom, J.; Yoon, K.; Aoki, S.; Rahman, S.; Nakamura, T.; Brust, F.; Yoshimura, S.

    1995-01-01

    Fracture mechanics and fatigue evaluations are an important part of the structural integrity analyses to assure safe operation of pressure vessels and piping components during their service life. The paper presented in this volume illustrate the application of fatigue and fracture mechanics techniques to assess the structural integrity of a wide variety of Pressure Vessels and Piping components. The papers are organized in six sections: (1) fatigue and fracture--vessels; (2) fatigue and fracture--piping; (3) fatigue and fracture--material property evaluations; (4) constraint effects in fracture mechanics; (5) probabilistic fracture mechanics analyses; and (6) user's experience with failure assessment diagrams. Separate abstracts were prepared for most of the papers in this book

  19. PWR internals segmentation and packaging experience in the U.S

    International Nuclear Information System (INIS)

    Kreitman, P.J.

    2008-01-01

    Seven commercial nuclear power plants of the Pressurized Water Reactor (PWR) design have been permanently shut down in the US to date. Six of these plants have been decommissioned using dismantling methods. The remaining plant, Indian Point 1, located in Buchanan, NY, has been placed in Safe Storage. Of the six dismantled plants, five underwent extensive segmentation, separation, and packaging of their internal components to allow removal and disposal of the reactor vessel assembly. Only the Trojan Plant in Portland, Oregon, was able to ship the reactor vessel assembly with its internals intact to the final disposal site near Richland, Washington. An important part of the planning for the upcoming decommissioning of similar plants in Europe such as Chooz A in France and Zorita in Spain will be to study the lessons learned from the US efforts and apply the best practices to future projects. This paper will chronicle the evolution of the reactor internals segmentation and packaging process to date, including the planning, methodology, equipment, waste management, and packaging strategy. (author)

  20. Stresses and displacements in vessels due to loads imposed by single and multiple piping attachments

    International Nuclear Information System (INIS)

    Wong, F.M.G.; Craft, W.J.; East, G.H.

    1985-01-01

    The Fourier solution for thin shell equations models pressure vessels as continuous simply connected surfaces with local loads. The technique allows placement of tractions with combinations of radial, shear, and axial components. Unlike Bijlaard, the solution in this paper includes loads placed at any position along the cylinder. Stiffness and the enhanced load-carrying capacity that internal pressure gives to thin vessels can be simulated. Numerical convergence problems are reduced by an improved displacement-load algorithm, and by use of load sites that allow the circular functions to be compactly grouped. A variety of loading distributions may be analyzed including large and small nozzles near and away from centerlines. Both rectangular and circular attachments are simulated. Through superposition, multiple attachments with their own loads may be examined. The attachments to the vessel may be either rigid or soft. A comparison to analytical results from Bijlaard shows excellent agreement. Comparisons with experimental tests on an API-650 nozzle on a storage tank are in good agreement. Variations between experimental and calculated results are primarily caused by assuming a simply supported base in the calculation, whereas in the experimental test, the base is more nearly fixed

  1. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    Energy Technology Data Exchange (ETDEWEB)

    Zaccaria, Pierluigi, E-mail: pierluigi.zaccaria@igi.cnr.it [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco [Consorzio RFX (CNR, ENEA, INFN, Università di Padova, Acciaierie Venete S.p.A.), Padova (Italy); Masiello, Antonio [Fusion for Energy F4E, Barcelona (Spain); Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario [Ettore Zanon S.p.A., Schio (VI) (Italy)

    2015-10-15

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  2. Manufacturing, assembly and tests of SPIDER Vacuum Vessel to develop and test a prototype of ITER neutral beam ion source

    International Nuclear Information System (INIS)

    Zaccaria, Pierluigi; Valente, Matteo; Rigato, Wladi; Dal Bello, Samuele; Marcuzzi, Diego; Agostini, Fabio Degli; Rossetto, Federico; Tollin, Marco; Masiello, Antonio; Corniani, Giorgio; Badalocchi, Matteo; Bettero, Riccardo; Rizzetto, Dario

    2015-01-01

    Highlights: • The SPIDER experiment aims to qualify and optimize the ion source for ITER injectors. • The large SPIDER Vacuum Vessel was built and it is under testing at the supplier. • The main working and assembly steps for production are presented in the paper. - Abstract: The SPIDER experiment (Source for the Production of Ions of Deuterium Extracted from an RF plasma) aims to qualify and optimize the full size prototype of the negative ion source foreseen for MITICA (full size ITER injector prototype) and the ITER Heating and Current Drive Injectors. Both SPIDER and MITICA experiments are presently under construction at Consorzio RFX in Padova (I), with the financial support from IO (ITER Organization), Fusion for Energy, Italian research institutions and contributions from Japan and India Domestic Agencies. The vacuum vessel hosting the SPIDER in-vessel components (Beam Source and calorimeters) has been manufactured, assembled and tested during the last two years 2013–2014. The cylindrical vessel, about 6 m long and 4 m in diameter, is composed of two cylindrical modules and two torispherical lids at the ends. All the parts are made by AISI 304 L stainless steel. The possibility of opening/closing the vessel for monitoring, maintenance or modifications of internal components is guaranteed by bolted junctions and suitable movable support structures running on rails fixed to the building floor. A large number of ports, about one hundred, are present on the vessel walls for diagnostic and service purposes. The main working steps for construction and specific technological issues encountered and solved for production are presented in the paper. Assembly sequences and tests on site are furthermore described in detail, highlighting all the criteria and requirements for correct positioning and testing of performances.

  3. UK regulatory aspects of prestressed concrete pressure vessels for gas-cooled reactor nuclear power stations

    International Nuclear Information System (INIS)

    Watson, P.S.

    1990-01-01

    Safety assessment principles for nuclear power plants and for nuclear chemical plants demand application of best proven techniques, recognised standards, adequacy margins, inspection and maintenance of all the components including prestressed concrete pressure vessels. In service inspection of prestressed concrete pressure vessels includes: concrete surface examination; anchorage inspection; tendon load check; tendon material examination; foundation settlement and tilt; log-term deformation; vessel temperature excursions; coolant loss; top cap deflection. Hartlepool and Heysham 1 power plants prestress shortfall problem is discussed. Main recommendations can be summarised as follows: at all pressure vessel stations prestress systems should be calibrated in a manner which results in all load bearing components being loaded in a representative manner; at all pressure vessel stations load measurements during calibration should be verified by a redundant and diverse system

  4. 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.)

  5. 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

  6. 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 containing the heat source, an outer shell enclosing the primary pressure vessel and acting as a secondary means of containment for this vessel against outside projectiles. Multiple auxiliary equipment points are arranged outside the outer shell which comprises a part of a lower wall around the primary pressure vessel, an annular part integrated in the lower wall and extending outwards as from this wall and an upper part integrated in the annular part and extending above this annular part and above the primary pressure vessel. The annular part and the primary pressure vessel are formed with vertical penetrations which can be closed communicating respectively with the auxiliary equipment points and with inside the pressure vessel whilst handling gear is provided in the upper part for vertically raising reactor components through these penetrations and for transporting them over the annular part and over the primary pressure vessel [fr

  7. Design and rescue scenario of common repair equipment for in-vessel components in ITER hot cell

    International Nuclear Information System (INIS)

    Kakudate, Satoshi; Takeda, Nobukazu; Nakahira, Masataka; Shibanuma, Kiyoshi

    2006-06-01

    Transportation of the in-vessel components to be repaired in the ITER hot cell is carried by two kinds of transporters, i.e., overhead cranes and floor vehicles. The access area for repair operations in the hot cell is duplicated by these transporters. Clear sharing of the respective roles of these transporters with the minimum duplication is therefore useful for rationalization. The overhead cranes, which are independently installed in the respective cells in the hot sell, cannot pass through the components to be repaired between cells, i.e., receiving cell and refurbishment cell as an example. If the floor vehicle with simple mechanisms can cover the inaccessible area for the overhead cranes, a global transporter system in the hot cell will be simplified and the reliability will be increased. Based on this strategy, the overhead crane and floor vehicle concepts are newly proposed. The overhead crane has an adapter for change of the end-effectors, which can be easily changed, to grasp many kinds of components to be repaired. The floor vehicle, which is equipped with wheel mechanisms for transportation, is just to pass through the components between cells with only straight (linear) motion on the floor. The simple wheel mechanism can solve the spread of the dust, which is the critical issue of the original air bearing mechanism for traveling in the 2001 FDR design. Rescue scenarios and procedures in the hot cell are also studied in this report. The proposed rescue crane has major two functions for rescue operations of the hot cell facility, i.e., one for the overhead crane and the other for refurbishment equipment such as workstation for divertor repair. The rescue of the faulty overhead crane is carried out using the rescue tool installed on the rescue crane or directly traveled by pushing/pulling by the rescue crane after docking on the faulty overhead crane. For the rescue of the workstation, the rescue crane consists of a telescopic manipulator (maximum length

  8. In-vessel core melt retention by RPV external cooling for high power PWR. MAAP 4 analysis on a LBLOCA scenario without SI

    International Nuclear Information System (INIS)

    Cognet, C.; Gandrille, P.

    1999-01-01

    In-, ex-vessel reflooding or both simultaneously can be envisaged as Accident Management Measures to stop a Severe Accident (SA) in vessel. This paper addresses the possibility of in-vessel core melt retention by RPV external flooding for a high power PWR (4250 MWth). The reactor vessel is assumed to have no lower head penetration and thermal insulation is neglected. The effects of external cooling of high power density debris, where the margin for such a strategy is low, are investigated with the MAAP4 code. MAAP4 code is used to verify the system capability to flood the reactor pit and to predict simultaneously the corium relocation into the lower head with the thermal and mechanical response of the RPV in transient conditions. The corium pool cooling and holding in the RPV lower head is analysed. Attention is paid to the internal heat exchanges between corium components. This paper focuses particularly the heat transfer between oxidic and metallic phases as well as between the molten metallic phase and the RPV wall of utmost importance for challenging the RPV integrity in vicinity of the metallic phase. The metal segregation has a decisive influence upon the attack of the vessel wall due to a very strong peaking of the lateral flux ('focusing effect'). Thus, the dynamics of the formation of the metallic layer characterized by a growing inventory of steel, both from a partial vessel ablation and the degradation of internals steel structures by the radiative heat flux from the debris, is displayed. The analysed sequence is a surge line rupture near the hot leg (LBLOCA) leading to the fastest accident progression

  9. RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

    International Nuclear Information System (INIS)

    2007-01-01

    The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

  10. Helium leak testing of large pressure vessels or subassemblies

    International Nuclear Information System (INIS)

    Hopkins, J.S.; Valania, J.J.

    1977-01-01

    Specifications for pressure-vessel components [such as the intermediate heat exchangers (IHX)] for service in the liquid metal fast breeder reactor facilities require helium leak testing of pressure boundaries to very exacting standards. The experience of Foster Wheeler Energy Corporation (FWEC) in successfully leak-testing the IHX shells and bundle assemblies now installed in the Fast Flux Test Facility at Richland, WA is described. Vessels of a somewhat smaller size for the closed loop heat exchanger system in the Fast Flux Test Facility have also been fabricated and helium leak tested for integrity of the pressure boundary by FWEC. Specifications on future components call for helium leak testing of the tube to tubesheet welds of the intermediate heat exchangers

  11. Simulation of the removal of NET internal components with dynamic modeling software

    International Nuclear Information System (INIS)

    Becquet, M.; Crutzen, Y.R.; Farfaletti-Casali, F.

    1989-01-01

    The replacement of the internal plasma-facing components (first-wall and blanket segments) for maintenance or at the end of their lifetime is an important aspect of the design of the next European torus (NET) and of the remote handling procedures. The first phase of development of the design software tool INVDYN (inverse dynamics) is presented, which will allow optimization of the movements of the internal segments during replacement, taking into account inertial effects and structural deformations. A first analysis of the removal of one NET internal segment provides, for a defined trajectory, the required generalized forces that must be applied on the crane system

  12. 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.

  13. 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

  14. 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

  15. 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

  16. Vacuum vessels for the LHC magnets arrive at CERN

    CERN Multimedia

    2001-01-01

    The first batch of pre-series vacuum vessels for the LHC dipole magnets has just been delivered to CERN. The vessels are components of the cryostats and will provide the thermal insulation for the superconducting magnets. The first batch of vacuum vessels for the LHC dipole magnets with the team taking part at CERN in ordering and installing them. Left to right : Claude Hauviller, Monique Dupont, Lloyd Williams, Franck Gavin, Alain Jacob, Christophe Vuitton, Davide Bozzini, Laure Sandri, Mikael Sjoholm and André de Saever. In 2006 all that will be seen of the LHC superconducting dipoles in the LHC tunnel will be a line of over 1230 blue cylindrical vacuum vessels. Ten vessels, each weighing 4 tonnes, are already at CERN. On 6 July the first batch of pre-series vessels reached the Lab-oratory from the firm SIMIC Spa whose works are near Savona in north-western Italy. Despite appearances, these 15-metre long, 1-metre diameter blue tubes are much more sophisticated than sections of a run-of-the-mill...

  17. 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...

  18. Firefighter's compressed air breathing system pressure vessel development program

    Science.gov (United States)

    Beck, E. J.

    1974-01-01

    The research to design, fabricate, test, and deliver a pressure vessel for the main component in an improved high-performance firefighter's breathing system is reported. The principal physical and performance characteristics of the vessel which were required are: (1) maximum weight of 9.0 lb; (2) maximum operating pressure of 4500 psig (charge pressure of 4000 psig); (3) minimum contained volume of 280 in. 3; (4) proof pressure of 6750 psig; (5) minimum burst pressure of 9000 psig following operational and service life; and (6) a minimum service life of 15 years. The vessel developed to fulfill the requirements described was completely sucessful, i.e., every category of performence was satisfied. The average weight of the vessel was found to be about 8.3 lb, well below the 9.0 lb specification requirement.

  19. Commissioning result of the KSTAR in-vessel cryo-pump

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y. B.; Lee, H. J.; Park, Y.M. [National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2013-12-15

    KSTAR in-vessel cryo-pump has been installed in the vacuum vessel top and bottom side with up-down symmetry for the better plasma density control in the D-shape H-mode. The cryogenic helium lines of the in-vessel cryo-pump are located at the vertical positions from the vacuum vessel torus center 2,000 mm. The inductive electrical potential has been optimized to reduce risk of electrical breakdown during plasma disruption. In-vessel cryo-pump consists of three parts of coaxial circular shape components; cryo-panel, thermal shield and particle shield. The cryo-panel is cooled down to below 4.5 K. The cryo-panel and thermal shields were made by Inconel 625 tube for higher mechanical strength. The thermal shields and their cooling tubes were annealed in air environment to improve the thermal radiation emissivity on the surface. Surface of cryo-panel was electro-polished to minimize the thermal radiation heat load. The in-vessel cryo-pump was pre-assembled on a test bed in 180 degree segment base. The leak test was carried out after the thermal shock between room temperature to LN2 one before installing them into vacuum vessel. Two segments were welded together in the vacuum vessel and final leak test was performed after the thermal shock. Commissioning of the in-vessel cryo-pump was carried out using a temporary liquid helium supply system.

  20. Status report: Intergranular stress corrosion cracking of BWR core shrouds and other internal components

    International Nuclear Information System (INIS)

    1996-03-01

    On July 25, 1994, the US Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 94-03 to obtain information needed to assess compliance with regulatory requirements regarding the structural integrity of core shrouds in domestic boiling water reactors (BWRs). This report begins with a brief description of the safety significance of intergranular stress corrosion cracking (IGSCC) as it relates to the design and function of BWR core shrouds and other internal components. It then presents a brief history of shroud cracking events both in the US and abroad, followed by an indepth summary of the industry actions to address the issue of IGSCC in BWR core shrouds and other internal components. This report summarizes the staff's basis for issuing GL 94-03, as well as the staff's assessment of plant-specific responses to GL 94-03. The staff is continually evaluating the licensee inspection programs and the results from examinations of BWR core shrouds and other internal components. This report is representative of submittals to and evaluations by the staff as of September 30, 1995. An update of this report will be issued at a later date

  1. Status of the ITER vacuum vessel construction

    Energy Technology Data Exchange (ETDEWEB)

    Choi, C.H.; Sborchia, C.; Ioki, K.; Giraud, B.; Utin, Yu.; Sa, J.W. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Wang, X., E-mail: xiaoyuwww@gmail.com [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Teissier, P.; Martinez, J.M.; Le Barbier, R.; Jun, C.; Dani, S.; Barabash, V.; Vertongen, P.; Alekseev, A. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Jucker, P.; Bayon, A. [F4E, c/ Josep Pla, n. 2, Torres Diagonal Litoral, Edificio B3, E-08019 Barcelona (Spain); Pathak, H.; Raval, J. [ITER-India, IPR, A-29, Electronics Estate, GIDC, Sector-25, Gandhinagar 382025 (India); Ahn, H.J. [ITER Korea, National Fusion Research Institute, Daejeon (Korea, Republic of); and others

    2014-10-15

    Highlights: • Final design of the ITER vacuum vessel (VV). • Procurement of the ITER VV. • Manufacturing results of real scale mock-ups. • Manufacturing status of the VV in domestic agencies. - Abstract: The ITER vacuum vessel (VV) is under manufacturing by four domestic agencies after completion of engineering designs that have been approved by the Agreed Notified Body (ANB). Manufacturing designs of the VV have been being completed, component by component, by accommodating requirements of the RCC-MR 2007 edition. Manufacturing of the VV first sector has been started in February 2012 in Korea and in-wall shielding in May 2013 in India. EU will start manufacturing of its first sector from September 2013 and Russia the upper port by the end of 2013. All DAs have manufactured several mock-ups including real-size ones to justify/qualify and establish manufacturing techniques and procedures.

  2. 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

  3. A wide angle view imaging diagnostic with all reflective, in-vessel optics at JET

    Energy Technology Data Exchange (ETDEWEB)

    Clever, M. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, 52425 Jülich (Germany); Arnoux, G.; Balshaw, N. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Garcia-Sanchez, P. [Laboratorio Nacional de Fusion, Asociacion EURATOM-CIEMAT, Madrid (Spain); Patel, K. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Sergienko, G. [Institute of Energy and Climate Research – Plasma Physics, Forschungszentrum Jülich GmbH, Association EURATOM-FZJ, 52425 Jülich (Germany); Soler, D. [Winlight System, 135 rue Benjamin Franklin, ZA Saint Martin, F-84120 Pertuis (France); Stamp, M.F.; Williams, J.; Zastrow, K.-D. [Euratom/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2013-10-15

    Highlights: ► A new wide angle view camera system has been installed at JET. ► The system helps to protect the ITER-like wall plasma facing components from damage. ► The coverage of the vessel by camera observation systems was increased. ► The system comprises an in-vessel part with parabolic and flat mirrors. ► The required image quality for plasma monitoring and wall protection was delivered. -- Abstract: A new wide angle view camera system has been installed at JET in preparation for the ITER-like wall campaigns. It considerably increases the coverage of the vessel by camera observation systems and thereby helps to protect the – compared to carbon – more fragile plasma facing components from damage. The system comprises an in-vessel part with parabolic and flat mirrors and an ex-vessel part with beam splitters, lenses and cameras. The system delivered the image quality required for plasma monitoring and wall protection.

  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. A model for structural analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A. de.

    1987-01-01

    Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

  6. An efficient modeling of fine air-gaps in tokamak in-vessel components for electromagnetic analyses

    International Nuclear Information System (INIS)

    Oh, Dong Keun; Pak, Sunil; Jhang, Hogun

    2012-01-01

    Highlights: ► A simple and efficient modeling technique is introduced to avoid undesirable massive air mesh which is usually encountered at the modeling of fine structures in tokamak in-vessel component. ► This modeling method is based on the decoupled nodes at the boundary element mocking the air gaps. ► We demonstrated the viability and efficacy, comparing this method with brute force modeling of air-gaps and effective resistivity approximation instead of detail modeling. ► Application of the method to the ITER machine was successfully carried out without sacrificing computational resources and speed. - Abstract: A simple and efficient modeling technique is presented for a proper analysis of complicated eddy current flows in conducting structures with fine air gaps. It is based on the idea of replacing a slit with the decoupled boundary of finite elements. The viability and efficacy of the technique is demonstrated in a simple problem. Application of the method to electromagnetic load analyses during plasma disruptions in ITER has been successfully carried out without sacrificing computational resources and speed. This shows the proposed method is applicable to a practical system with complicated geometrical structures.

  7. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

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

  8. Expanding plasma jet in a vacuum vessel

    International Nuclear Information System (INIS)

    Chutov, Yu.I.; Kravchenko, A.Yu.; Yakovetskij, V.S.

    1998-01-01

    The paper deals with numerical calculations of parameters of a supersonic quasi-neutral argon plasma jet expanding into a cylindrical vacuum vessel and interacting with its inner surface. A modified method of large particles was used, the complex set of hydrodynamic equations being broken into simpler components, each of which describes a separate physical process. Spatial distributions of the main parameters of the argon plasma jet were simulated at various times after the jet entering the vacuum vessel, the parameters being the jet velocity field, the full plasma pressure, the electron temperature, the temperature of heavy particles, and the degree of ionization. The results show a significant effect of plasma jet interaction on the plasma parameters. The jet interaction with the vessel walls may result e.g. in excitation of shock waves and rotational plasma motions. (J.U.)

  9. Design improvements and R and D achievements for VV and In-vessel components towards ITER construction

    International Nuclear Information System (INIS)

    Ioki, Kimihiro; Barabaschi, P.; Barabash, V.

    2003-01-01

    There have been several detailed vacuum vessel (VV) design improvements, such as elimination of the inboard triangular support, separate interspace between inner and outer shells for independent leak detection of field joints and revised VV support system to gain a more comfortable margin in the structural performance. The blanket design has been updated; an inter-modular key instead of two prismatic keys and a co-axial inlet-outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R and D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and non destructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. With regard to the R and D for the FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block and the divertor components have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods. (author)

  10. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. Life extension of components with high cumulative fatigue usage

    International Nuclear Information System (INIS)

    Server, W.L.

    1985-01-01

    The current ASME Boiler and Pressure Vessel Code fatigue design approach has significant margins of safety as evidenced by fatigue data on full-scale vessels. In order to extend qualification (life) of components which have reached the Code design usage limit of unity, improved criteria are needed which address crack initiation and propagation separately such that safe operation of these components is ensured. The fatigue initiation phase is composed of two processes: initial microcracking of internal particles or accumulation of local strain (cyclic slip) creating discontinuities which form microcracks, and the growth of the microcracks in a noncontinuum manner. The microcracks which initiate will eventually grow to a size in which continuum mechanics apply, and fracture mechanics concepts have been employed. The later propagation to failure of a component is also composed of two parts, continuum crack growth in a stable manner and eventual unstable fracture of the remaining ligament of material. This paper reviews the current status of technology in assessing initiation and propagation relative to the current design Code and suggests areas of improvement to cover extended life of high usage factor components. To illustrate some of these considerations, a case study for a small manufacturing defect was reviewed. A realistic component was analyzed to investigate the interrelationship between the ASME Code Section III design life and crack propagation behavior of a small manufacturing defect. A pressurized water reactor (PWR) primary coolant system was used in the analysis, and the terminal end of the hot-leg pipe at the safe end weld was selected since usage factors as high as 0.95 had been reported. The particular plant chosen was Zion-1 because the necessary information on loading, including thermal transients, was available in the open literature. 11 refs., 1 fig., 1 tab

  12. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    1999-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  13. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2001-01-01

    This paper reviews the design and performance of the in-vessel components of ITER as developed for the EDA Final Design Report (FDR). The double-wall vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g., the most intense VDE's and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature differences. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected and these are accommodated by HHF technology developed during the EDA. Disruptions and VDE's can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowables for all postulated disruption and seismic events. (author)

  14. An experimental system for the study of ultrasound exposure of isolated blood vessels

    NARCIS (Netherlands)

    Tokarczyk, Anna; Rivens, Ian; van Bavel, E.; Symonds-Tayler, Richard; ter Haar, Gail

    2013-01-01

    An experimental system designed for the study of the effects of diagnostic or therapeutic ultrasound exposure on isolated blood vessels in the presence or absence of intraluminal contrast agent is described. The system comprised several components. A microscope was used to monitor vessel size (and

  15. 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

  16. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  17. 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

  18. 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.

  19. 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...

  20. Structural analysis of NPP components and structures

    International Nuclear Information System (INIS)

    Saarenheimo, A.; Keinaenen, H.; Talja, H.

    1998-01-01

    Capabilities for effective structural integrity assessment have been created and extended in several important cases. In the paper presented applications deal with pressurised thermal shock loading, PTS, and severe dynamic loading cases of containment, reinforced concrete structures and piping components. Hydrogen combustion within the containment is considered in some severe accident scenarios. Can a steel containment withstand the postulated hydrogen detonation loads and still maintain its integrity? This is the topic of Chapter 2. The following Chapter 3 deals with a reinforced concrete floor subjected to jet impingement caused by a postulated rupture of a near-by high-energy pipe and Chapter 4 deals with dynamic loading resistance of the pipe lines under postulated pressure transients due to water hammer. The reliability of the structural integrity analysing methods and capabilities which have been developed for application in NPP component assessment, shall be evaluated and verified. The resources available within the RATU2 programme alone cannot allow performing of the large scale experiments needed for that purpose. Thus, the verification of the PTS analysis capabilities has been conducted by participation in international co-operative programmes. Participation to the European Network for Evaluating Steel Components (NESC) is the topic of a parallel paper in this symposium. The results obtained in two other international programmes are summarised in Chapters 5 and 6 of this paper, where PTS tests with a model vessel and benchmark assessment of a RPV nozzle integrity are described. (author)

  1. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.-P.

    1995-01-01

    Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

  2. Synergistic failure of BWR internals

    International Nuclear Information System (INIS)

    Ware, A. G.; Chang, T.Y.

    1999-01-01

    Boiling Water Reactor (BWR) core shrouds and other reactor internals important to safety are experiencing intergranular stress corrosion cracking (IGSCC). The United States Nuclear Regulatory Commission has followed the problem, and as part of its investigations, contracted with the Idaho National Engineering and Environmental Laboratory to conduct a risk assessment. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, with specific consideration given to potential cascading and common mode effects. An initial phase has been completed in which background material was gathered and evaluated, and potential accident sequences were identified. A second phase is underway to perform a simplified, quantitative probabilistic risk assessment on a representative high-power BWR/4. Results of the initial study conducted on the jet pumps show that any cascading failures would not result in a significant increase in the core damage frequency. The methodology is currently being extended to other major reactor internals components

  3. Large component regulatory relief in the United States

    International Nuclear Information System (INIS)

    Williams, J.L.; Boyle, R.W.

    2004-01-01

    The dismantling of retired nuclear power and test facilities required the transport of reactor vessels, reactor vessel heads, pressurizers, reactor coolant pumps, and steam generators; whereas the replacement of degraded components to continue operations has been generally limited to pressurized water reactor steam generators, and more recently, reactor heads. These components are quite large and massive, measuring up to 20 feet in diameter and 70 feet in length and weighing 50 to 600 tons. Issue Several issues arose, due to the implementation of the 1985 IAEA transportations regulations in the United States, on the practical matters of how to characterize the components and comply with the transportation regulations. The concept of a Surface Contaminated Object (SCO) was new to many in the U.S., and the large components were not readily amenable to transportation under the regulations. It was apparent that most of the components should be considered as SCO. However, it was not certain that the contamination limits for inaccessible areas could be met due to non-uniform contamination deposition; nor could the interior areas be readily surveyed without on-site dismantlement of the large component. Additionally, while the components were generally substantial in design and construction due to their use as pressure vessels under other codes, the current regulations required packages that met tests such as for stacking and free drop that would pose severe engineering challenges, prohibitive costs, or logistical difficulties during transport, due to the size and weight of the components being transport

  4. Assembly & Metrology of First Wall Components of SST-1

    Science.gov (United States)

    Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal; Patel, Hiteshkumar; Paravastu, Yuvakiran; Jaiswal, Snehal; Chauhan, Pradeep; Babu, Gattu Ramesh; A, Arun Prakash; Bhavsar, Dhaval; Raval, Dilip C.; Khan, Ziauddin; Pradhan, Subrata

    2017-04-01

    First Wall Components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring & port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 undergoing a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects & procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel.

  5. 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)

  6. Development of ITER in-vessel viewing and metrology systems

    Energy Technology Data Exchange (ETDEWEB)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation ({approx}30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  7. Development of ITER in-vessel viewing and metrology systems

    International Nuclear Information System (INIS)

    Obara, Kenjiro; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The ITER in-vessel viewing system is vital for detecting and locating damage to in-vessel components such as the blankets and divertors and in monitoring and assisting in-vessel maintenance. This system must be able to operate at high temperature (200degC) under intense gamma radiation (∼30 kGy/h) in a high vacuum or 1 bar inert gas. A periscope viewing system was chosen as a reference due to its clear, wide view and a fiberscope viewing system chosen as a backup for viewing in narrow confines. According to the ITER R and D program, both systems and a metrology system are being developed through the joint efforts of Japan, the U.S., and RF Home Teams. This paper outlines design and technology development mainly on periscope in-vessel viewing and laser metrology contributed by the Japan Home Team. (author)

  8. 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.

  9. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  10. Results of assembly test of HTTR reactor internals

    International Nuclear Information System (INIS)

    Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

    1996-01-01

    The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

  11. 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

  12. Comparison between 3D eddy current patterns in tokamak in-vessel components generated by disruptions

    International Nuclear Information System (INIS)

    Sakellaris, J.; Crutzen, Y.

    1996-01-01

    During plasma disruption events in Tokamaks, a large amount of magnetic energy is associated to the transfer of plasma current into eddy currents in the passive structures. In the ITER program two design concepts have been proposed. One approach (ITER CDA design) is based on copper stabilization loops (i.e., twin loops) attached to box-shaped blanket segments, electrically and mechanically separated along the toroidal direction. For another design concept (ITER EDA design) based on lower plasma elongation there is no need for specific stabilization loops. The passive stabilization is obtained by toroidally continuous components (i.e., the plasma facing wall of the blanket segments allows a continuity along the toroidal direction). Consequently, toroidal currents flow, when electromagnetic transients occur. Electromagnetic loads appear in the blanket structures in case of plasma disruptions and/or vertical displacement events either for the ITER CDA design concept or for the ITER EDA design concept. In this paper the influence of the in-vessel design configuration concepts--insulated segments or electrically continuous structures--in terms of magnetic shielding and electric insulation on the magnitude and the flow pattern of the eddy currents is investigated. This investigation will allow a performance evaluation of the two proposed design concepts

  13. 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

  14. 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...

  15. 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.

  16. Internal examination of CAGR's

    International Nuclear Information System (INIS)

    Walton, P.J.; Langley, J.P.; Simons, C.R.; Hart, J.D.

    1975-01-01

    During selected reactor maintenance periods it will be necessary to conduct examinations of reactor internal components to gain assurance of their continuing integrity throughout the design life. Equipment for remote examination is being developed that will measure the distortions of the graphite moderator bricks and extract samples from them for analysis. Other remotely operated equipment will allow closed circuit TV viewing inside the reactors: the equipment provides both wide field and detailed inspection facilities within a single compact unit. Manipulators are being developed to ring the viewing units to the best vantage points. Reeling and handling gear will be operated from outside the vessels and these provide full containment at all times. (author)

  17. Boiling-Water Reactor internals aging degradation study

    International Nuclear Information System (INIS)

    Luk, K.H.

    1993-09-01

    This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

  18. Plasma facing components design of KT-2 tokamak

    International Nuclear Information System (INIS)

    In, Sang Ryul; Yoon, Byung Joo; Song, Woo Soeb; Xu, Chao Yin

    1997-04-01

    The vacuum vessel of KT-2 tokamak is protected from high thermal loads by various kinds of plasma facing components (PFC): outer and inner divertors, neutral baffle, inboard limiter, poloidal limiter, movable limiter and passive plate, installed on the inner wall of the vessel. In this report the pre-engineering design of the plasma facing components, including design requirements and function, structures of PFC assemblies, configuration of cooling systems, calculations of some mechanical and hydraulic parameters, is presented. Pumping systems for the movable limiter and the divertor are also discussed briefly. (author). 49 figs

  19. Basic Components of Vascular Connective Tissue and Extracellular Matrix.

    Science.gov (United States)

    Halper, Jaroslava

    2018-01-01

    Though the composition of the three layers constituting the blood vessel wall varies among the different types of blood vessels, and some layers may even be missing in capillaries, certain basic components, and properties are shared by all blood vessels, though each histologically distinct layer contains a unique complement of extracellular components, growth factors and cytokines, and cell types as well. The structure and composition of vessel layers informs and is informed by the function of the particular blood vessel. The adaptation of the composition and the resulting function of the extracellular matrix (ECM) to changes in circulation/blood flow and a variety of other extravascular stimuli can be characterized as remodeling spearheaded by vascular cells. There is a surprising amount of cell traffic among the three layers. It starts with endothelial cell mediated transmigration of inflammatory cells from the bloodstream into the subendothelium, and then into tissue adjoining the blood vessel. Smooth muscle cells and a variety of adventitial cells reside in tunica media and tunica externa, respectively. The latter cells are a mixture of progenitor/stem cells, fibroblasts, myofibroblasts, pericytes, macrophages, and dendritic cells and respond to endothelial injury by transdifferentiation as they travel into the two inner layers, intima and media for corrective mission in the ECM composition. This chapter addresses the role of various vascular cell types and ECM components synthesized by them in maintenance of normal structure and in their contribution to major pathological processes, such as atherosclerosis, organ fibrosis, and diabetic retinopathy. © 2018 Elsevier Inc. All rights reserved.

  20. Sealing analysis for nuclear vessel of PWR

    International Nuclear Information System (INIS)

    Qu, J.; Dou, Y.

    1987-01-01

    Although design by analysis of pressure vessel has become a requirement in all codes for more than 20 years, sealing design for nuclear components is still too complicated and there are yet no criteria about this aspect, even though in the well-known ASME Boiler and Pressure Vessel Code. Thus it is of significance to undertake researches of transient sealing tests and analysis for nuclear vessel. Since 1960s great progress has been made in analytic computer program, which takes flange as a rigid ring. Actually, however, there are elastic or elastoplastic contacts on flange mating surface. Chen (1979) gave a mixed finite element method, using a condensing flexible matrix skill, to solve two-body contact problem. On the basis of axisymmetric stress and thermal analysis of finite element method and on accepting Chen's (1979) idea of mixed finite element method, we have developed a computer program for sealing analysis, named SMEC, which considers bolt loading changes and temperature effects. (orig./GL)

  1. Design improvements and R and D achievements for VV and in-vessel components towards ITER construction

    International Nuclear Information System (INIS)

    Ioki, K.; Barabaschi, P.; Barabash, V.

    2003-01-01

    During the preparation of the procurement specifications for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain a more comfortable structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inlet outlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R and D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and nondestructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods. (author)

  2. Stress analysis on a PWR pressure vessel support structure

    International Nuclear Information System (INIS)

    Cruz, J.R.B.; Mattar Neto, M.; Jesus Miranda, C.A. de.

    1992-01-01

    The paper presents the stress analysis of a research PWR vessel support structure. Different geometries and thermal boundary conditions are evaluated. The finite element analysis is performed using ANSYS program. The ASME Section III criteria are applied for the stress verification and the following points are discussed: stress classification and linearization; jurisdictional boundary between ASME Subsection NB (Class 1 Components) and Subsection NF (Component Supports). (author)

  3. ITER in-vessel system design and performance

    Science.gov (United States)

    Parker, R. R.

    2000-03-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events.

  4. ITER in-vessel system design and performance

    International Nuclear Information System (INIS)

    Parker, R.R.

    2000-01-01

    The article reviews the design and performance of the in-vessel components of ITER as developed for the Engineering Design Activities (EDA) Final Design Report. The double walled vacuum vessel is the first confinement boundary and is designed to maintain its integrity under all normal and off-normal conditions, e.g. the most intense vertical displacement events (VDEs) and seismic events. The shielding blanket consists of modules connected to a toroidal backplate by flexible connectors which allow differential displacements due to temperature non-uniformities. Breeding blanket modules replace the shield modules for the Enhanced Performance Phase. The divertor concept is based on a cassette structure which is convenient for remote installation and removal. High heat flux (HHF) components are mechanically attached and can be removed and replaced in the hot cell. Operation of the divertor is based on achieving partially detached plasma conditions along and near the separatrix. Nominal heat loads of 5-10 MW/m 2 are expected on the target. These are accommodated by HHF technology developed during the EDA. Disruptions and VDEs can lead to melting of the first wall armour but no damage to the underlying structure. Stresses in the main structural components remain within allowable ranges for all postulated disruption and seismic events. (author)

  5. Compact insert design for cryogenic pressure vessels

    Energy Technology Data Exchange (ETDEWEB)

    Aceves, Salvador M.; Ledesma-Orozco, Elias Rigoberto; Espinosa-Loza, Francisco; Petitpas, Guillaume; Switzer, Vernon A.

    2017-06-14

    A pressure vessel apparatus for cryogenic capable storage of hydrogen or other cryogenic gases at high pressure includes an insert with a parallel inlet duct, a perpendicular inlet duct connected to the parallel inlet. The perpendicular inlet duct and the parallel inlet duct connect the interior cavity with the external components. The insert also includes a parallel outlet duct and a perpendicular outlet duct connected to the parallel outlet duct. The perpendicular outlet duct and the parallel outlet duct connect the interior cavity with the external components.

  6. 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.)

  7. 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.)

  8. Simulant melt experiments on performance of the in-vessel core catcher

    International Nuclear Information System (INIS)

    Kyoung-Ho Kang; Rae-Joon Park; Sang-Baik Kim; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

    2005-01-01

    Full text of publication follows: 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 uniform gap of 5 mm or 10 mm to the inner surface of the lower head vessel. As a performance test of the in-vessel core catcher, the effects of base steel and internal coating materials and gap thickness between the core catcher and the lower head vessel were examined in this study. In the LAVA-GAP-2 and LAVA-GAP-3 tests, the base steel was carbon steel and the gap thickness was 10 mm. On the other hand, in the LAVA-GAP-4 and LAVA-GAP-5 tests, the base steel was stainless steel and the gap thickness was 5 mm. Actual composition of the coating material for the LAVA-GAP-4 test was 92% of ZrO 2 - 8% of Y 2 O 3 including 95% of Ni - 5% of Al bond coat same as the LAVA-GAP-3 test. In these tests, the thickness of ZrO 2 internal coating was 0.5 mm. To examine the effects of the coating material, in-vessel core catcher with a 0.6 mm-thick ZrO 2 coating without bond coat was used in the LAVA-GAP-5 test. This report summarizes the experimental results and the post metallurgical inspection results of the LAVA-GAP-4 and LAVA-GAP- 5 tests. In the LAVA-GAP-4 and LAVA-GAP-5 tests, the core catcher was failed and it was stuck to the inner surface of the lower head vessel. LAVA-GAP-4 and LAVA-GAP-5 test results imply that 5 mm thick gap is rather small for sufficient water ingression and steam venting through the gap. In case of small gap size, water is boiled off and steam increases pressure inside the gap and so water can not ingress into the gap at the initial heat up stage. Metallurgical inspections on the test specimens indicate that the internal coating layer might melt totally and dispersed in the base steel and the solidified iron melt and so the detection frequencies of Zr and O are trivial all

  9. 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

  10. Development of the Sixty Watt Heat-Source hardware components

    International Nuclear Information System (INIS)

    McNeil, D.C.; Wyder, W.C.

    1995-01-01

    The Sixty Watt Heat Source is a nonvented heat source designed to provide 60 thermal watts of power. The unit incorporates a plutonium-238 fuel pellet encapsulated in a hot isostatically pressed General Purpose Heat Source (GPHS) iridium clad vent set. A molybdenum liner sleeve and support components isolate the fueled iridium clad from the T-111 strength member. This strength member serves as the pressure vessel and fulfills the impact and hydrostatic strength requirements. The shell is manufactured from Hastelloy S which prevents the internal components from being oxidized. Conventional drawing operations were used to simplify processing and utilize existing equipment. The deep drawing reqirements for the molybdenum, T-111, and Hastelloy S were developed from past heat source hardware fabrication experiences. This resulted in multiple step drawing processes with intermediate heat treatments between forming steps. The molybdenum processing included warm forming operations. This paper describes the fabrication of these components and the multiple draw tooling developed to produce hardware to the desired specifications. copyright 1995 American Institute of Physics

  11. 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...

  12. In vivo and in vitro methods to study platelet adhesion to the components of the vessel wall

    International Nuclear Information System (INIS)

    Cazenave, J.-P.

    1979-01-01

    The methods that are used to measure platelet adhesion can be divided in five groups: methods that use an aggregometer to measure platelet adhesion to collagen in the presence of EDTA; methods that use binding of radiolabeled collagen, affinity chromatography, or gel filtration; the morphometric method of Baumgartner that measures platelet interaction with the subendothelium of an aorta exposed to flow in an annular perfusion chamber; the quantitative isotopic measurement of platelet adhesion to collagen-coated surfaces and to subendothelium with the rotating probe device of Cazenave; and in vivo platelet adhesion to the subendothelium measured by the morphometric method or with platelets radiolabeled with 51 Cr or 111 In. With these methods it has been possible to study the factors (Ca 2+ ; VIII: von Willebrand factor; hemodynamic factors: red cells, shear rate; components of the vessel wall) governing platelet adhesion to subendothelium and to collagen. It has also been possible to screen and study drugs inhibiting platelet adhesion, which is the first step in the formation of a thrombus at the site of vascular injury [fr

  13. Investigation of residual stresses in thick-walled vessels with combination of autofrettage and wire-winding

    International Nuclear Information System (INIS)

    Sedighi, M.; Jabbari, A.H.

    2013-01-01

    Wire-winding and autofrettage processes can be used to introduce beneficial residual stress in the cylinder of thick-walled pressure vessels. In both techniques, internal residual compressive stress will increase internal pressure capacity, improve fatigue life and reduce fatigue crack initiation. The purpose of this paper is to analyze the effects of wire-winding on an autofrettaged thick-walled vessel. Direct method which is a modified Variable Material Properties (VMP) method has been used in order to calculate residual stresses in an autofrettaged vessel. Since wire-winding is done after autofrettage process, the tangent and/or Young's modulus could be changed. For this reason, a new wire-winding method based on Direct Method is introduced. The obtained results for wire-wound autofrettaged vessels are validated by finite element method. The results show that by using this approach, the residual hoop stresses in a wire-wound autofrettaged vessel have a more desirable distribution in the cylinder. -- Highlights: • Combination of autofrettage and wire-winding in pressure vessels has been presented. • A new method based on Direct method is presented for wire-winding process. • Residual hoop stresses are compared in vessels cylinders for different cases. • The residual hoop stress has a more desirable stress distribution. • The benefits of the combined vessel are highlighted in comparison with single cases

  14. 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

  15. 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

  16. The FENIX [Fusion ENgineering International EXperimental] test facility

    International Nuclear Information System (INIS)

    Slack, D.S.; Patrick, R.E.; Chaplin, M.R.; Miller, J.R.; Shen, S.S.; Summers, L.T.; Kerns, J.A.

    1989-01-01

    The Fusion ENgineering International EXperimental Magnet Facility (FENIX), under construction at Lawrence Livermore National Laboratory (LLNL), is a significant step forward in meeting the testing requirements necessary for the development of superconductor for large-scale, superconducting magnets. A 14-T, transverse field over a test volume of 150 x 60 x 150 mm in length will be capable of testing conductors the size of the International Thermonuclear Experimental Reactor (ITER). Proposed conductors for ITER measure ∼35 mm on one side and will operate at currents of up to 40 kA at fields of ∼14 T. The testing of conductors and associated components, such as joints, will require large-bore, high-field magnet facilities. FENIX is being constructed using the existing A 2o and A 2i magnets from the idle MFTF. The east and west A 2 pairs will be mounted together to form a split-pair solenoid. The pairs of magnets will be installed in a 4.0-m cryostat vessel located in the HFTF building at LLNL. Each magnet is enclosed in its own cryostat, the existing 4.0-m vessel serving only as a vacuum chamber. 4 refs., 8 figs

  17. Assembly and metrology of first wall components of SST-1

    International Nuclear Information System (INIS)

    Parekh, Tejas; Santra, Prosenjit; Biswas, Prabal

    2015-01-01

    First Wall components (FWC) of SST-1 tokamak, which are in the immediate vicinity of plasma comprises of limiters, divertors, baffles, passive stabilizers are designed to operate long duration (1000 s) discharges of elongated plasma. All FWC consists of a copper alloy heat sink modules with SS cooling tubes brazed onto it, graphite tiles acting as armour material facing the plasma, and are mounted to the vacuum vessels with suitable Inconel support structures at ring and port locations. The FWC are very recently assembled and commissioned successfully inside the vacuum vessel of SST-1 under going a meticulous planning of assembly sequence, quality checks at every stage of the assembly process. This paper will present the metrology aspects and procedure of each FWC, both outside the vacuum vessel, and inside the vessel, assembly tolerances, tools, equipment and jig/fixtures, used at each stage of assembly, starting from location of support bases on vessel rings, fixing of copper modules on support structures, around 3800 graphite tile mounting on 136 copper modules with proper tightening torques, till final toroidal and poloidal geometry of the in-vessel components are obtained within acceptable limits, also ensuring electrical continuity of passive stabilizers to form a closed saddle loop, electrical isolation of passive stabilizers from vacuum vessel. (author)

  18. Evaluation and silicon nitride internal combustion engine components. Final report, Phase I

    Energy Technology Data Exchange (ETDEWEB)

    Voldrich, W. [Allied-Signal Aerospace Co., Torrance, CA (United States). Garrett Ceramic Components Div.

    1992-04-01

    The feasibility of silicon nitride (Si{sub 3}N{sub 4}) use in internal combustion engines was studied by testing three different components for wear resistance and lower reciprocating mass. The information obtained from these preliminary spin rig and engine tests indicates several design changes are necessary to survive high-stress engine applications. The three silicon nitride components tested were valve spring retainers, tappet rollers, and fuel pump push rod ends. Garrett Ceramic Components` gas-pressure sinterable Si{sub 3}N{sub 4} (GS-44) was used to fabricate the above components. Components were final machined from densified blanks that had been green formed by isostatic pressing of GS-44 granules. Spin rig testing of the valve spring retainers indicated that these Si{sub 3}N{sub 4} components could survive at high RPM levels (9,500) when teamed with silicon nitride valves and lower spring tension than standard titanium components. Silicon nitride tappet rollers showed no wear on roller O.D. or I.D. surfaces, steel axles and lifters; however, due to the uncrowned design of these particular rollers the cam lobes indicated wear after spin rig testing. Fuel pump push rod ends were successful at reducing wear on the cam lobe and rod end when tested on spin rigs and in real-world race applications.

  19. Electromagnetic and structural analyses of the vacuum vessel and plasma facing components for EAST

    International Nuclear Information System (INIS)

    Xu, Weiwei; Liu, Xufeng; Song, Yuntao; Li, Jun; Lu, Mingxuan

    2013-01-01

    Highlights: • The electromagnetic and structural responses of VV and PFCs for EAST are analyzed. • A detailed finite element model of the VV including PFCs is established. • The two most dangerous scenarios, major disruptions and downward VDEs are considered. • The distribution patterns of eddy currents, EMFs and torques on PFCs are analyzed. -- Abstract: During plasma disruptions, time-varying eddy currents are induced in the vacuum vessel (VV) and Plasma Facing Components (PFCs) of EAST. Additionally, halo currents flow partly through these structures during the vertical displacement events (VDEs). Under the high magnetic field circumstances, the resulting electromagnetic forces (EMFs) and torques are large. In this paper, eddy currents and EMFs on EAST VV, PFCs and their supports are calculated by analytical and numerical methods. ANSYS software is employed to evaluate eddy currents on VV, PFCs and their structural responses. To learn the electromagnetic and structural response of the whole structure more accurately, a detailed finite element model is established. The two most dangerous scenarios, major disruptions and downward VDEs, are examined. It is found that distribution patterns of eddy currents for various PFCs differ greatly, therefore resulting in different EMFs and torques. It can be seen that for certain PFCs the transient reaction force are severe. Results obtained here may set up a preliminary foundation for the future dynamic response research of EAST VV and PFCs which will provide a theoretical basis for the future engineering design of tokamak devices

  20. 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)

  1. Assessing the port to port risk of vessel movements vectoring non-indigenous marine species within and across domestic Australian borders.

    Science.gov (United States)

    Campbell, Marnie L; Hewitt, Chad L

    2011-07-01

    Biofouling of vessels is implicated as a high risk transfer mechanism of non-indigenous marine species (NIMS). Biofouling on international vessels is managed through stringent border control policies, however, domestic biofouling transfers are managed under different policies and legislative arrangements as they cross internal borders. As comprehensive guidelines are developed and increased compliance of international vessels with 'clean hull' expectations increase, vessel movements from port to port will become the focus of biosecurity management. A semi-quantitative port to port biofouling risk assessment is presented that evaluates the presence of known NIMS in the source port and determines the likelihood of transfer based on the NIMS association with biofouling and environmental match between source and receiving ports. This risk assessment method was used to assess the risk profile of a single dredge vessel during three anticipated voyages within Australia, resulting in negligible to low risk outcomes. This finding is contrasted with expectations in the literature, specifically those that suggest slow moving vessels pose a high to extreme risk of transferring NIMS species.

  2. Thermography for detection of scaling in slurry lines and process vessels

    International Nuclear Information System (INIS)

    Capolingua, Adam; Petrik, Andrew

    2006-01-01

    A major problem in many of today's refineries and mineral processing plants is internal scale build-up within slurry lines and process vessels. Consequences of such an internal scale build-up within lines and vessels include machinery damage, flow restrictions, blockages and localised pipe wear. These problems lead to a loss of production, increased maintenance costs, impinge on worker safety, increase environmental hazards and inevitably reduces profit for the organisation of concern. Hence, the application of an efficient and accurate non-intrusive detection method for locating internal scale within kilometres of lines and numerous process vessels is imperative to reduce maintenance costs and limit production losses. Thermography has been found to be a very useful NDT technique for applications where there is a differential between the ambient and internal product temperatures. The 'insulating' effect of the internal scaling results in a reduced external temperature over the associated area. These temperature differentials can be efficiently detected via a thermographic scan. While this technique is relatively straightforward, the interpretation of the thermographic images usually requires reasonable skill and experience to assess the true extent of each problem detected. In some cases, the true location and extent of scaling within the slurry lines may not be thermally obvious due to the nature of the internal scaling. In such cases, the use of other complementary methods to effectively 'listen'in to the lines has proved to be a valuable procedure. In particular a technology that is typically used in vibration monitoring to assess bearing and gear degradation has been successfully applied in conjunction with thermography to assess lines with localised or dislodged scale. This paper presents a number of case studies where thermography was either applied independently or in conjunction with other measurement techniques, to detect and assess different internal

  3. An experimental system for the study of ultrasound exposure of isolated blood vessels

    International Nuclear Information System (INIS)

    Tokarczyk, Anna; Rivens, Ian; Symonds-Tayler, Richard; Ter Haar, Gail; Van Bavel, E

    2013-01-01

    An experimental system designed for the study of the effects of diagnostic or therapeutic ultrasound exposure on isolated blood vessels in the presence or absence of intraluminal contrast agent is described. The system comprised several components. A microscope was used to monitor vessel size (and thus vessel functionality), and potential leakage of intraluminal 70 kDa FITC-dextran fluorescence marker. A vessel chamber allowed the mounting of an isolated vessel whilst maintaining its viability, with pressure regulation for the control of intraluminal pressure and induction of flow for the infusion of contrast microbubbles. A fibre-optic hydrophone sensor mounted on the vessel chamber using a micromanipulator allowed pre-exposure targeting of the vessel to within 150 µm, and monitoring of acoustic cavitation emissions during exposures. Acoustic cavitation was also detected using changes in the ultrasound drive voltage and by detection of audible emissions using a submerged microphone. The suitability of this system for studying effects in the isolated vessel model has been demonstrated using a pilot study of 6 sham exposed and 18 high intensity focused ultrasound exposed vessels, with or without intraluminal contrast agent (SonoVue) within the vessels. (paper)

  4. [Verification of ballast water exchange for international ships anchored in Xiamen Port by CDOM tracer].

    Science.gov (United States)

    Li, Chao; Zhang, Yan-po; Guo, Wei-dong; Zhu, Yue; Xu, Jing; Deng, Xun

    2010-09-01

    Fluorescence excitation-emission matrix (EEM) and absorption spectroscopy were applied to study the optical properties of 29 CDOM samples collected from different ballast tanks of nine international route vessels anchored in Xiamen Port between October 2007 and April 2008. The purpose was to examine the feasibility of these spectral properties as a tracer to verify if these vessels follow the mid-ocean ballast water exchange (BWE) regulation. Using parallel factor analysis, four fluorescent components were identified, including two humic-like components (C1: 245, 300/386 nm; C2: 250, 345/458 nm) and two protein-like components (C3: 220, 275/306 nm; C4: 235, 290/345 nm), of which C2 component was the suitable fluorescence verification indicator. The vertical distribution of all fluorescent components in ballast tank was nearly similar indicating that profile-mixing sampling was preferable. Combined use of C2 component, spectral slope ratio (SR) of absorption spectroscopy and salinity may provide reasonable verification if BWE carried out by these nine ships. The results suggested that the combined use of multiple parameters (fluorescence, absorption and salinity) would be much reliable to determine the origin of ballast water, and to provide the technical guarantee for fast examination of ballast water exchange in Chinese ports.

  5. Positioning means for circumferentially locating inspection apparatus in a nuclear reactor vessel

    International Nuclear Information System (INIS)

    Burns, D.C.

    1979-01-01

    Positioning means for locating inspection apparatus used to volumetrically examine a nuclear reactor vessel is disclosed. The positioning means is provided with a support ring having an annular key positioned longitudinally about its periphery. Three support legs are attached to the support ring by brackets adapted to fit the annular key. The support ring also carries three guide stud bushings which are movably mounted by clamps adapted to engage the support ring key. Prior to lowering the inspection apparatus into the vessel, the guide stud bushings are each moved to a point of alignment with one of three guide studs extending upwardly from the vessel. After alignment has been verified, the guide stud bushings are clamped in position. The inspection apparatus is lowered towards its fully seated position within the vessel and is coarsely circumferentially positioned with by the engagement of the guide studs within the guide stud bushings. A fine degree of circumferential positioning is achieved by providing a specially configured shoe for one of the support legs. With the core barrel internals in, the special shoe is adapted to key onto a core barrel pin the exact location of which is known. With the core barrel internals removed, the special shoe is adapted to place a locating key into a notch in a vessel flange, the location of which is known. As the inspection apparatus is lowered into its fully seated position, exact circumferential positioning with respect to the vessel is achieved. The other support legs rest on an inner circumferential flange so that no portion of the inspection apparatus touches or threatens the vessel's top flange. 19 claims

  6. Iter in vessel viewing system design and assessment activities

    Energy Technology Data Exchange (ETDEWEB)

    Neri, C., E-mail: carlo.neri@enea.it [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy); Costa, P.; Ferri De Collibus, M.; Florean, M.; Mugnaini, G.; Pillon, M.; Pollastrone, F.; Rossi, P. [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2011-10-15

    The In Vessel Viewing System (IVVS) is fundamental remote handling equipment, which will be used to make a survey of the status of the blanket first wall and divertor plasma facing components. A prototype of a laser In Vessel Viewing and ranging System was developed and tested at ENEA laboratories in Frascati under EFDA task agreements, it is able to perform sub-millimetric bi-dimensional and three-dimensional images inside ITER during maintenance procedure allowing the evaluation of the state and damages of the in-vessel surface. The present prototype has been designed to operate under room conditions and starting from springtime 2009 a Grant with F4E is in progress for the design and the assessment of the IVVS system for ITER, keeping in account all the environmental conditions and constraints.

  7. French administrative practice and design codes for nuclear vessels

    International Nuclear Information System (INIS)

    Roche, R.L.

    1987-07-01

    French regulations on boilers and pressure vessels have prevailed for a very long time, the first measure having been promulgated on 29 October 1823. Restraining the attention to nuclear pressure vessels it must be pointed out regulations and enforcement by public authorities are more stringent than they are for conventional pressure vessels. The first part of this paper will be devoted to regulations with a special attention to the decree of 26 February 1974 and to the practice of public authorities in this field with special attention given to the Bureau de Controle de la Construction Nucleaire (BCCN = Bureau of Inspection of Nuclear Design and Manufacturing). The second part of this paper will deal with the French construction codes for nuclear components RCC-M (water reactors) and RCC-MR (elevated temperature design)

  8. Advanced Approach of Reactor Pressure Vessel In-service Inspection

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Pajnic, M.

    2006-01-01

    The most important task of every utility operating a nuclear power plant is the continuously keeping of the desired safety and reliability level. This is achieved by the performance of numerous inspections of the components, equipment and system of the nuclear power plant in operation and in particular during the scheduled maintenance periods at re-fueling time. Periodic non-destructive in-service inspections provide most relevant criteria of the integrity of primary circuit pressure components. The task is to reliably detect defects and realistically size and characterize them. One of most important and the most extensive examination is a reactor pressure vessel in-service inspection. That inspection demand high standards of technology and quality and continual innovation in the field of non-destructive testing (NDT) advanced technology as well as regarding reactor pressure vessel tool and control systems. A remote underwater contact ultrasonic technique is employed for the examination of the defined sections (reactor welds), whence eddy current method is applied for clad surface examinations. Visual inspection is used for examination of the vessel inner surface. The movement of probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with NDT systems. The successful performance is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen NDT techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state of the art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. The advanced approach as presented in this paper offer more flexibility of application (non-destructive tests, local grinding action as well as taking of boat samples

  9. Conceptual design finalisation of the ITER In-Vessel Viewing and Metrology System (IVVS)

    Energy Technology Data Exchange (ETDEWEB)

    Dubus, Gregory, E-mail: gregory.dubus@f4e.europa.eu [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Puiu, Adrian; Damiani, Carlo; Van Uffelen, Marco; Lo Bue, Alessandro; Izquierdo, Jesus; Semeraro, Luigi [Fusion for Energy, c/ Josep Pla, n°2 - Torres Diagonal Litoral - Edificio B3, 08019 Barcelona (Spain); Martins, Jean-Pierre; Palmer, Jim [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R and D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis.

  10. 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

  11. 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.

  12. 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

  13. 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

  14. PWR reactor vessel in-service-inspection according to RSEM

    International Nuclear Information System (INIS)

    Algarotti, Marc; Dubois, Philippe; Hernandez, Luc; Landez, Jean Paul

    2006-01-01

    Nuclear services experience Framatome ANP (an AREVA and Siemens company) has designed and constructed 86 Pressurized Water Reactors (PWR) around the world including the three units lately commissioned at Ling Ao in the People's Republic of China and ANGRA 2 in Brazil; the company provided general and specialized outage services supporting numerous outages. Along with the American and German subsidiaries, Framatome ANP Inc. and Framatome ANP GmbH, Framatome ANP is among the world leading nuclear services providers, having experience of over 500 PWR outages on 4 continents, with current involvement in more than 50 PWR outages per year. Framatome ANP's experience in the examinations of reactor components began in the 1970's. Since then, each unit (American, French and German companies) developed automated NDT inspection systems and carried out pre-service and ISI (In-Service Inspections) using a large range of NDT techniques to comply with each utility expectations. These techniques have been validated by the utilities and the safety authorities of the countries where they were implemented. Notably Framatome ANP is fully qualified to provide full scope ISI services to satisfy ASME Section XI requirements, through automated NDE tasks including nozzle inspections, reactor vessel head inspections, steam generator inspections, pressurizer inspections and RPV (Reactor Pressure Vessel) inspections. Intercontrole (Framatome ANP subsidiary dedicated in supporting ISI) is one of the leading NDT companies in the world. Its main activity is devoted to the inspection of the reactor primary circuit in French and foreign PWR Nuclear Power Plants: the reactor vessel, the steam generators, the pressurizer, the reactor internals and reactor coolant system piping. NDT methods mastered by Intercontrole range from ultrasonic testing to eddy current and gamma ray examinations, as well as dye penetrant testing, acoustic monitoring and leak testing. To comply with the high requirements of

  15. 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)

  16. 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

  17. PFMC-16. 16th international conference on plasma-facing materials and components for fusion applications. Abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2017-07-01

    The performances of fusion devices and of future fusion power plants strongly depend on the plasma-facing materials and components. Resistance to heat and particle loads, compatibility in plasma operations, thermo-mechanical properties, as well as the response to neutron irradiation are critical parameters which need to be understood and tailored from atomistic to component levels. The 16th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues.

  18. 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

  19. Automatic segmentation of blood vessels from retinal fundus images ...

    Indian Academy of Sciences (India)

    The retinal blood vessels were segmented through color space conversion and color channel extraction, image pre-processing, Gabor filtering, image postprocessing, feature construction through application of principal component analysis, k-means clustering and first level classification using Naïve–Bayes classification ...

  20. Analysis of toroidal vacuum vessels for use in demonstration sized tokamak reactors

    International Nuclear Information System (INIS)

    Culbert, M.E.

    1978-07-01

    The vacuum vessel component of the tokamak fusion reactor is the subject of this study. The main objective of this paper was to provide guidance for the structural design of a thin wall externally pressurized toroidal vacuum vessel. The analyses are based on the available state-of-the-art analytical methods. The shortcomings of these analytical methods necessitated approximations and assumptions to be made throughout the study. A principal result of the study has been the identification of a viable vacuum vessel design for the Demonstration Tokamak Hybrid Reactor (DTHR) and The Next Step (TNS) Reactor

  1. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae

    2016-01-01

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study

  2. Evaluation for In-Vessel Retention Capabilities with In-Vessel Injection and External Reactor Vessel Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Seong; Ryu, In Chul; Moon, Young Tae [KEPCO Engineering and Construction Co. Ltd., Deajeon (Korea, Republic of)

    2016-10-15

    If the accident has not progressed to the point of substantial changes in the core geometry, establishing adequate cooling is as straightforward as re-establishing flow through the reactor core. However, if the accident has progressed to the point where the core geometry is substantially altered as a result of material melting and relocation, as was the case in the TMI-2 accident, the means of cooling the debris are not as straightforward. From this time on, the reactor core was either completely or nearly covered by water, with high pressure injection flow initiated shortly after three hours into the accident. However, the core debris was not coolable in this configuration and a substantial quantity of molten core material drained into the bypass region, with approximately twenty metric tons of molten debris draining into the reactor pressure vessel (RPV) lower head. Hence, the core configuration developed at approximately three hours into the accident was not coolable, even submerged in water. The purpose of this paper is to evaluate in-vessel retention capabilities with in-vessel injection (IVI) and external reactor vessel cooling (ERVC) available in a reactor application by using the integrated severe accident analysis code. The MAAP5 models were improved to facilitate evaluation of the in-vessel retention capability of APR1400. In-vessel retention capabilities have been analyzed for the APR1400 using the MAAP5.03 code. The results show that in-vessel retention is feasible when in-vessel injection is initiated within a relatively short time frame under the simulation condition used in the present study.

  3. Nuclear power plant component protection

    International Nuclear Information System (INIS)

    Michel, E.; Ruf, R.; Dorner, H.

    1976-01-01

    Described is a nuclear power plant installation which includes a concrete biological shield forming a pit in which a reactor pressure vessel is positioned. A steam generator on the outside of the shield is connected with the pressure vessel via coolant pipe lines which extend through the shield, the coolant circulation being provided by a coolant pump which is also on the outside of the shield. To protect these components on the outside of the shield and which are of mainly or substantially cylindrical shape, semicylindrical concrete segments are interfitted around them to form complete outer cylinders which are retained against outward separation radially from the components, by rings of high tensile steel which may be interspaced so closely that they provide, in effect, an outer steel cylinder. The invention is particularly applicable to pressurized-water coolant reactor installations

  4. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    Science.gov (United States)

    Kumar, B. Ramesh; Gangradey, R.

    2012-11-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  5. Thick SS316 materials TIG welding development activities towards advanced fusion reactor vacuum vessel applications

    International Nuclear Information System (INIS)

    Kumar, B Ramesh; Gangradey, R

    2012-01-01

    Advanced fusion reactors like ITER and up coming Indian DEMO devices are having challenges in terms of their materials design and fabrication procedures. The operation of these devices is having various loads like structural, thermo-mechanical and neutron irradiation effects on major systems like vacuum vessel, divertor, magnets and blanket modules. The concept of double wall vacuum vessel (VV) is proposed in view of protecting of major reactor subsystems like super conducting magnets, diagnostic systems and other critical components from high energy 14 MeV neutrons generated from fusion plasma produced by D-T reactions. The double walled vacuum vessel is used in combination with pressurized water circulation and some special grade borated steel blocks to shield these high energy neutrons effectively. The fabrication of sub components in VV are mainly used with high thickness SS materials in range of 20 mm- 60 mm of various grades based on the required protocols. The structural components of double wall vacuum vessel uses various parts like shields, ribs, shells and diagnostic vacuum ports. These components are to be developed with various welding techniques like TIG welding, Narrow gap TIG welding, Laser welding, Hybrid TIG laser welding, Electron beam welding based on requirement. In the present paper the samples of 20 mm and 40 mm thick SS 316 materials are developed with TIG welding process and their mechanical properties characterization with Tensile, Bend tests and Impact tests are carried out. In addition Vickers hardness tests and microstructural properties of Base metal, Heat Affected Zone (HAZ) and Weld Zone are done. TIG welding application with high thick SS materials in connection with vacuum vessel requirements and involved criticalities towards welding process are highlighted.

  6. 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.)

  7. Project management techniques used in the European Vacuum Vessel sectors procurement for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Losasso, Marcello, E-mail: marcello.losasso@f4e.europa.eu [Fusion for Energy (F4E), Barcelona (Spain); Ortiz de Zuniga, Maria; Jones, Lawrence; Bayon, Angel; Arbogast, Jean-Francois; Caixas, Joan; Fernandez, Jose; Galvan, Stefano; Jover, Teresa [Fusion for Energy (F4E), Barcelona (Spain); Ioki, Kimihiro [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lewczanin, Michal; Mico, Gonzalo; Pacheco, Jose Miguel [Fusion for Energy (F4E), Barcelona (Spain); Preble, Joseph [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Stamos, Vassilis; Trentea, Alexandru [Fusion for Energy (F4E), Barcelona (Spain)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer File name contains the directory tree structure with a string of three-letter acronyms, thereby enabling parent directory location when confronted with orphan files. Black-Right-Pointing-Pointer The management of the procurement procedure was carried out in an efficient and timely manner, achieving precisely the contract placement date foreseen at the start of the process. Black-Right-Pointing-Pointer The contract start-up has been effectively implemented and a flexible project management system has been put in place for an efficient monitoring of the contract. - Abstract: The contract for the seven European Sectors of the ITER Vacuum Vessel (VV) was placed at the end of 2010 with a consortium of three Italian companies. The task of placing and the initial take-off of this large and complex contract, one of the largest placed by F4E, the European Domestic Agency for ITER, is described. A stringent quality controlled system with a bespoke Vacuum Vessel Project Lifecycle Management system to control the information flow, based on ENOVIA SmarTeam, was developed to handle the storage and approval of Documentation including links to the F4E Vacuum Vessel system and ITER International Organization System interfaces. The VV Sector design and manufacturing schedule is based on Primavera software, which is cost loaded thus allowing F4E to carry out performance measurement with respect to its payments and commitments. This schedule is then integrated into the overall Vacuum Vessel schedule, which includes ancillary activities such as instruments, preliminary design and analysis. The VV Sector Risk Management included three separate risk analyses from F4E and the bidders, utilizing two different methodologies. These efforts will lead to an efficient and effective implementation of this contract, vital to the success of the ITER machine, since the Vacuum Vessel is the biggest single work package of Europe's contribution to ITER and

  8. Proactive life extension of pressure vessels

    Science.gov (United States)

    Mager, Lloyd

    1998-03-01

    place while our vessels are in service. As the inspection takes place we are able to view a real time image of detected discontinuities on a video monitor. The B-scan ultrasonic technique is allowing us to perform fast accurate examinations covering up to 95% of the surface area of each pressure vessel. Receiving data on 95% of a pressure vessel provides us with a lot of useful information. We use this data to determine the condition of each pressure vessel. Once the condition is known the vessels are classed by risk. The risk level is then managed by making decisions related to repair, operating parameters, accepting and monitoring or replacement of the equipment. Inspection schedules are set at maximum intervals and reinspection is minimized for the vessels that are not at risk. The remaining life of each pressure vessel is determined, mechanical integrity is proven and regulatory requirements are met. Abbott Laboratories is taking this proactive approach because we understand that our process equipment is a critical element for successful operation. A run to failure practice would never allow Abbott Laboratories to achieve the corporation's objective of being the world's leading health care company. Nondestructive state of the art technology and the understanding of its capabilities and limitations are key components of a proactive program for life extension of pressure vessels. 26

  9. Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1981-01-01

    Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

  10. 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.)

  11. 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

  12. Semi-automated retinal vessel analysis in nonmydriatic fundus photography.

    Science.gov (United States)

    Schuster, Alexander Karl-Georg; Fischer, Joachim Ernst; Vossmerbaeumer, Urs

    2014-02-01

    Funduscopic assessment of the retinal vessels may be used to assess the health status of microcirculation and as a component in the evaluation of cardiovascular risk factors. Typically, the evaluation is restricted to morphological appreciation without strict quantification. Our purpose was to develop and validate a software tool for semi-automated quantitative analysis of retinal vasculature in nonmydriatic fundus photography. matlab software was used to develop a semi-automated image recognition and analysis tool for the determination of the arterial-venous (A/V) ratio in the central vessel equivalent on 45° digital fundus photographs. Validity and reproducibility of the results were ascertained using nonmydriatic photographs of 50 eyes from 25 subjects recorded from a 3DOCT device (Topcon Corp.). Two hundred and thirty-three eyes of 121 healthy subjects were evaluated to define normative values. A software tool was developed using image thresholds for vessel recognition and vessel width calculation in a semi-automated three-step procedure: vessel recognition on the photograph and artery/vein designation, width measurement and calculation of central retinal vessel equivalents. Mean vessel recognition rate was 78%, vessel class designation rate 75% and reproducibility between 0.78 and 0.91. Mean A/V ratio was 0.84. Application on a healthy norm cohort showed high congruence with prior published manual methods. Processing time per image was one minute. Quantitative geometrical assessment of the retinal vasculature may be performed in a semi-automated manner using dedicated software tools. Yielding reproducible numerical data within a short time leap, this may contribute additional value to mere morphological estimates in the clinical evaluation of fundus photographs. © 2013 Acta Ophthalmologica Scandinavica Foundation. Published by John Wiley & Sons Ltd.

  13. 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

  14. Preliminary Performance Analysis Program Development for Safety System with Safeguard Vessel

    International Nuclear Information System (INIS)

    Kang, Han-Ok; Lee, Jun; Park, Cheon-Tae; Yoon, Ju-Hyeon; Park, Keun-Bae

    2007-01-01

    SMART is an advanced modular integral type pressurized water reactor for a seawater desalination and an electricity production. Major components of the reactor coolant system such as the pressurizer, Reactor Coolant Pump (RCP), and steam generators are located inside the reactor vessel. The SMART can fundamentally eliminate the possibility of large break loss of coolant accidents (LBLOCAs), improve the natural circulation capability, and better accommodate and thus enhance a resistance to a wide range of transients and accidents. The safety goals of the SMART are enhanced through highly reliable safety systems such as the passive residual heat removal system (PRHRS) and the safeguard vessel coupled with the passive safety injection feature. The safeguard vessel is a steel-made, leak-tight pressure vessel housing the RPV, SIT, and the associated valves and pipelines. A primary function of the safeguard vessel is to confine any radioactive release from the primary circuit within the vessel under DBAs related to loss of the integrity of the primary system. A preliminary performance analysis program for a safety system using the safeguard vessel is developed in this study. The developed program is composed of several subroutines for the reactor coolant system, passive safety injection system, safeguard vessel including the pressure suppression pool, and PRHRS. A small break loss of coolant accident at the upper part of a reactor is analyzed and the results are discussed

  15. 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

  16. Nonlinear response of vessel walls due to short-time thermomechanical loading

    International Nuclear Information System (INIS)

    Pfeiffer, P.A.; Kulak, R.F.

    1994-01-01

    Maintaining structural integrity of the reactor pressure vessel (RPV) during a postulated core melt accident is an important safety consideration in the design of the vessel. This study addresses the failure predictions of the vessel due to thermal and pressure loadings fro the molten core debris depositing on the lower head of the vessel. Different loading combinations were considered based on the dead load, yield stress assumptions, material response and internal pressurization. The analyses considered only short term failure (quasi static) modes, long term failure modes were not considered. Short term failure modes include plastic instabilities of the structure and failure due to exceeding the failure strain. Long term failure odes would be caused by creep rupture that leads to plastic instability of the structure. Due to the sort time durations analyzed, creep was not considered in the analyses presented

  17. Ageing study of Cirus reactor vessel expansion bellow

    International Nuclear Information System (INIS)

    Ramana, W.V.; Dutta, B.K.; Kushwaha, H.S.; Sahu, A.K.; Bhatnagar, A.; Pant, R.C.

    1994-01-01

    Expansion bellow of Cirus reactor vessel is a comparatively weak component which is joined to top tube sheet and shell by helium tight lap weld. This has been subjected to thermal stress caused by high temperature during reactor operation and thermal shock due to trip or shutdown. Therefore a finite element analysis was carried out to assess thermal stresses and fatigue life of the component. It was found that the fluctuating stress in the bellow is far less than its endurance limit. (author). 2 tabs., 3 figs

  18. 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...

  19. Customizable engineered blood vessels using 3D printed inserts.

    Science.gov (United States)

    Pinnock, Cameron B; Meier, Elizabeth M; Joshi, Neeraj N; Wu, Bin; Lam, Mai T

    2016-04-15

    Current techniques for tissue engineering blood vessels are not customizable for vascular size variation and vessel wall thickness. These critical parameters vary widely between the different arteries in the human body, and the ability to engineer vessels of varying sizes could increase capabilities for disease modeling and treatment options. We present an innovative method for producing customizable, tissue engineered, self-organizing vascular constructs by replicating a major structural component of blood vessels - the smooth muscle layer, or tunica media. We utilize a unique system combining 3D printed plate inserts to control construct size and shape, and cell sheets supported by a temporary fibrin hydrogel to encourage cellular self-organization into a tubular form resembling a natural artery. To form the vascular construct, 3D printed inserts are adhered to tissue culture plates, fibrin hydrogel is deposited around the inserts, and human aortic smooth muscle cells are then seeded atop the fibrin hydrogel. The gel, aided by the innate contractile properties of the smooth muscle cells, aggregates towards the center post insert, creating a tissue ring of smooth muscle cells. These rings are then stacked into the final tubular construct. Our methodology is robust, easily repeatable and allows for customization of cellular composition, vessel wall thickness, and length of the vessel construct merely by varying the size of the 3D printed inserts. This platform has potential for facilitating more accurate modeling of vascular pathology, serving as a drug discovery tool, or for vessel repair in disease treatment. Copyright © 2015 Elsevier Inc. All rights reserved.

  20. 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.

  1. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  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. PFMC14. 14th international conference on plasma-facing materials and components for fusion applications. Book of abstracts

    International Nuclear Information System (INIS)

    2013-01-01

    The performance of fusion devices and of a future fusion power plant critically depends on the plasma facing materials and components. Resistance to local heat and particle loads, thermo-mechanical properties, as well as the response to neutron damage of the selected materials are critical parameters which need to be understood and tailored from atomistic to component levels. The 14th International Conference on Plasma-Facing Materials and Components for Fusion Applications addresses these issues. Among the topics of the joint conference recent developments and research results in the following fields are addressed: - Tungsten and tungsten alloys - Low-Z materials - Mixed materials - Erosion, redeposition and fuel retention - Materials under extreme thermal loads - Technology and testing of plasma-facing components - Neutron effects in plasma-facing materials - Advanced characterization of materials and components. Selected international speakers present overview lectures and treat detailed aspects of the given topics. Contributed papers to the subjects of the meeting are solicited for oral and poster presentations.

  4. Examination of VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    Matokovic, A.; Picek, E.; Markulin, K.

    2008-01-01

    The increasing demand of a higher level of safety in the operation of the nuclear power plants requires the utilisation of more precise automated equipment to perform in-service inspections. That has been achieved by technological advances in computer technology, in robotics, in examination probe technology with the development of the advanced inspection technique and has also been due to the considerable and varied experience gained in the performance of such inspections. In-service inspection of reactor pressure vessel, especially Russian-designed WWER-1000 presents one of the most important and extensive examination of nuclear power plants primary circuit components. Such examination demand high standards of inspection technology, quality and continual innovation in the field of non-destructive testing advanced technology. A remote underwater contact ultrasonic technique is employed for the examination of the base metal of vessel and reactor welds, whence eddy current method is applied for clad surface examinations. Visual testing is used for examination of the vessel interior. The movement of inspection probes and data positioning are assured by using new reactor pressure vessel tool concept that is fully integrated with inspection systems. The successful performance of reactor pressure vessel is attributed thorough pre-outage planning, training and successful performance demonstration qualification of chosen non-destructive techniques on the specimens with artificial and/or real defects. Furthermore, use of advanced approach of inspection through implementation the state-of-the-art examination equipment significantly reduced the inspection time, radiation exposure to examination personnel, shortening nuclear power plant outage and cutting the total inspection costs. This paper presents advanced approach in the reactor pressure vessel in-service inspections and it is especially developed for WWER-1000 nuclear power plants.(author)

  5. On the state of acoustic emission analysis in pressure vessel and model vessel testing

    International Nuclear Information System (INIS)

    Morgner, W.; Theis, K.; Henke, F.; Imhof, D.

    1985-01-01

    In the GDR acoustic emission analysis is being applied primarily in connection with hydraulic pressure testing of vessels in chemical industry. It is, however, also used for testing and monitoring of equipment and components in other branches of industry. The state-of-the-art is presented with regard to equipment needed, training of personnel, licensing of testing methods and appropriate testing procedures. In particular, the evaluation of the sum curves and amplitude distributions is explained, using rupture tests of two oxygen cylinders and a compressed-air bottle as examples. (author)

  6. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  7. Dismantling large components at the Jose-Cabrera NPP (CNJC) in Spain

    International Nuclear Information System (INIS)

    Santiago, Juan Luis

    2012-01-01

    Located in central Spain, near Madrid, the Jose-Cabrera NPP (also known as Zorita) is the first PWR to be dismantled in Spain. The unit is a one-loop Westinghouse PWR, with a capacity of 150 MW. The plant was shut down in 1996 and ENRESA (Empresa Nacional de Residuos Radioactivos) has decided its prompt decommissioning, starting in 2010. In preparation for decommissioning, a full system decontamination (FSD) of the whole reactor cooling system (including the reactor vessel in the flow path) was carried out in 2006-7. The large components to be dismantled include: the reactor pressure vessel (RPV) and the internals; the vessel head; the SG; the pressurizer and the surge line; the reactor coolant pump, and the primary loop piping The objective of the project is not only to ensure the safe and efficient dismantling of those large components, but also to gain experience and to learn lessons to be applied during the future decommissioning and dismantling of the remaining six operating PWRs in Spain, whose operational lives are currently planned to end between 2021 and 2028. ENRESA has defined a waste-management policy for decommissioning activities, which includes Waste-management routes and optimisation. A case study describes the results obtained by ENRESA in the specific case of CNJC large components (including the RPV) dismantling project: Removal and conditioning of large components as a single piece is not considered a viable option. Segmentation therefore is required and 2 options have been analysed: large pieces for disposal in a large container and small pieces for disposal in approved concrete packages (CE-2a and the smaller CE-2b). The use of the CE-2b package is a feasible option and is easy to implement as a logical extension from the CE-2a. The use of the CE-2b package results in an important reduction in the total volume of final waste packages and does not require, in itself, any changes in the current waste handling and kinematics. The large size

  8. Contracts used for the charter or lease of pleasure vessels in pleasure navigation : an Italian perspective

    Directory of Open Access Journals (Sweden)

    Elena Orrù

    2018-02-01

    Full Text Available The Italian Navigation Code has transposed the practices developed at international level, in particular in international contracts for the ‘’locazione’’ and ‘’noleggio’’ of ships, distinguishing between the ship lease, from the one side, and the charter, from the other. The latter, in particular, consists of voyage charter and time charter. However, the Italian discipline differs in several respects from the contract types developed at international level. As for pleasure vessels, a specific regime lacked until the Law of 11 February 1971, No 50. The great development of this sector (which was previously considered limited to the use of pleasure vessels only for personal purposes, in particular of the entrepreneurial use of these vessels, furthered the draft and enactment, in 2005, of the Pleasure Navigation Code (Law of 18 July 2005, No 171, providing for a more comprehensive regime, however still not covering all the issues and aspects of pleasure navigation. The Code provides for a special regime of the contracts for the lease and charter of pleasure vessels: this article provides a review of the regime of these contracts provided by the Italian Pleasure Navigation Code, with regard also to its relationship with the Navigation Code and the Civil Code. The Code’s provisions are also examined with reference to standard contracts developed at the international level.

  9. 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...

  10. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  11. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  12. 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

  13. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

    2001-03-15

    This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

  14. Design and construction of Alborz tokamak vacuum vessel system

    International Nuclear Information System (INIS)

    Mardani, M.; Amrollahi, R.; Koohestani, S.

    2012-01-01

    Highlights: ► The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. ► As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. ► A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. ► Structural analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. - Abstract: The Alborz tokamak is a D-shape cross section tokamak that is under construction in Amirkabir University of Technology. At the heart of the tokamak is the vacuum vessel and limiter which collectively are referred to as the vacuum vessel system. As one of the key components for the device, the vacuum vessel can provide ultra-high vacuum and clean environment for the plasma operation. The VV systems need upper and lower vertical ports, horizontal ports and oblique ports for diagnostics, vacuum pumping, gas puffing, and maintenance accesses. A limiter is a solid surface which defines the edge of the plasma and designed to protect the wall from the plasma, localizes the plasma–surface interaction and localizes the particle recycling. Basic structure analyses were confirmed by FEM model for dead weight, vacuum pressure and plasma disruptions loads. Stresses at general part of the VV body are lower than the structure material allowable stress (117 MPa) and this analysis show that the maximum stresses occur near the gravity support, and is about 98 MPa.

  15. Development of standard components for remote handling

    International Nuclear Information System (INIS)

    Taguchi, Kou; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira

    1998-01-01

    The core of Fusion Experimental Reactor consists of various components such as superconducting magnets and forced-cooled in-vessel components, which are remotely maintained due to intense of gamma radiation. Mechanical connectors such as cooling pipe connections, insulation joints and electrical connectors are commonly used for maintenance of these components and have to be standardized in terms of remote handling. This paper describes these mechanical connectors developed as the standard component compatible with remote handling and tolerable for radiation. (author)

  16. Development of standard components for remote handling

    Energy Technology Data Exchange (ETDEWEB)

    Taguchi, Kou; Kakudate, Satoshi; Nakahira, Masataka; Ito, Akira [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-04-01

    The core of Fusion Experimental Reactor consists of various components such as superconducting magnets and forced-cooled in-vessel components, which are remotely maintained due to intense of gamma radiation. Mechanical connectors such as cooling pipe connections, insulation joints and electrical connectors are commonly used for maintenance of these components and have to be standardized in terms of remote handling. This paper describes these mechanical connectors developed as the standard component compatible with remote handling and tolerable for radiation. (author)

  17. 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...

  18. A patient with Moyamoya-like vessels after radiation therapy for a tumor in the basal ganglia

    International Nuclear Information System (INIS)

    Ishiyama, Koichi; Tomura, Noriaki; Kato, Koki; Takahashi, Satoshi; Watarai, Jiro; Sasajima, Toshio; Mizoi, Kazuo

    2001-01-01

    A patient with Moyamoya-like vessels after radiation therapy for treatment of a tumor in the basal ganglia is reported. He was diagnosed as Down syndrome at birth. He had a tumor in the left basal ganglionic region at 12 years of the age. The tumor increased in size at age 14. He underwent cerebral angiography, which did not show a stenosis nor occlusion of the internal carotid artery, anterior cerebral artery, nor the middle cerebral artery. He received radiation therapy with a total dose of 56 Gy. He presented a dressing apraxia at age 19. MRI showed cerebral infarction in the left temporo-occipital region. Right internal carotid angiography revealed a severe stenosis of the internal carotid artery and anterior cerebral artery as well as a severe stenosis of the middle cerebral artery on the right side. Moyamoya-like vessels were seen in the basal ganglionic region. Left internal carotid angiography also showed a stenosis of the internal carotid artery and anterior cerebral artery as well as a severe stenosis of the middle cerebral artery on the left side. Moyamoya-like vessels were seen in the basal ganglionic region. Leptomeningeal anastomose and transdural anastomose were bilaterally seen. These arterial occlusion and stenotic phenomenon corresponded to a previous radiation field. These Moyamoya-like vessels with arterial stenosis and occlusion were thought to be due to radiation-induced vasculopathy, because a previous cerebral angiography showed a normal caliber of cerebral arteries. This patient showed that patients with radiation therapy in their early childhood should be carefully observed considering the possibility of the phenomenon. (author)

  19. Reliability-based dynamic positioning of floating vessels with riser and mooring system

    DEFF Research Database (Denmark)

    Fang, Shaoji; Leira, Bernt J.; Blanke, Mogens

    2011-01-01

    To maintain safety of a floating vessel with associated slender components such as risers and mooring line, the vessel is normally kept within a limited region. To specify a safe position in that region, this paper suggests a new position chasing algorithm with the consideration of both riser ang...... to their criticality. An optimal position set-point is produced by minimization of the value of the cost function. Numerical simulations show the effectiveness of the proposed algorithm....

  20. 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

  1. Internal corium catcher of a nuclear reactor

    International Nuclear Information System (INIS)

    Anatolii S Vlasov; Vladimir N Mineev; Aleksandr S Sidorov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: A corium catcher is one of the main devices of a nuclear reactor that provides corium melt and fission products retention within a containment during severe accidents. Several studies and design developments have shown that corium retention within a reactor vessel can be attained with a moderate capacity of the latter (up to 600 - 650 MW el.). With a higher reactor capacity external corium catchers are applied both at Russian (VVER-1000) and European (EPR) reactors. In the external catcher of a VVER-1000 reactor, most technological problems are solved due to using sacrificial material. They are as follows: (a) endo-thermal interaction of corium and sacrificial material reduces a level of the temperatures in the final melt pool; (b) solution in the melt of a great amount of the sacrificial material reduces the specific heat release density and the heat flux density at the boundaries of a melt; (c) due to changing of the oxide-component density an inverse stratification of the metallic and oxide components of the corium takes place, thus excluding heat-flux focusing in the zone of the metallic layer and making it possible to supply water on the free surface of the corium without a danger of incipience of the vapor explosion; (d) final oxidation of zirconium occurs without hydrogen generation. The above principles have been realized in the external catcher of the VVER- 1000 reactor at Tyanvan NPS that is presently under construction in China. Successfully solving of the problems concerning to the external catcher makes it possible to return on the new conceptual and technological basis to the idea of retention of the corium melt inside the vessel of a nuclear reactor of large capacity, that is, to provide the reactor vessel to play a role of an internal catcher. For this purpose, a reactor vessel is elongated by approximately two meters. In the lower part of the vessel, on elliptical bottom, pieces of sacrificial material are arranged

  2. 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

  3. 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)

  4. Simulant melt experiments on performance of the in-vessel core catcher

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Kim, Sang-Baik; Suh, K.Y.; Cheung, F.B.; Rempe, J.L.

    2007-01-01

    In order to enhance the feasibility of in-vessel retention (IVR) of molten core material during a severe accident for high-power reactors, an in-vessel core catcher (IVCC) was designed and evaluated as part of a joint United States-Korean International Nuclear Energy Research Initiative (INERI). The proposed IVCC is expected to increase the thermal margin for success of IVR by providing an 'engineered gap' for heat transfer from materials that relocate during a severe accident and potentially serving as a sacrificial material under a severe accident. In this study, LAVA-GAP experiments were performed to investigate the thermal and mechanical performance of the IVCC using the alumina melt as simulant. The LAVA-GAP experiments aim to examine the feasibility and sustainability of the IVCC under the various test conditions using 1/8th scale hemispherical test sections. As a feasibility test of the proposed IVCC in this INERI project, the effects of IVCC base steel materials, internal coating materials, and gap size between the IVCC and the vessel lower head were examined. The test results indicated that the internally coated IVCC has high thermal performance compared with the uncoated IVCC. In terms of integrity of the base steel, carbon steel is superior to stainless steel and the effect of bond coat is found to be trivial for the tests performed in this study. The thermal load is mitigated via boiling heat removal in the gap between the IVCC and the vessel lower head. The current test results imply that gaps less than 10 mm are not enough to guarantee effective cooling induced by water ingression and steam venting there through. Selection of endurable material and pertinent gap size is needed to implement the proposed IVCC concept into advanced reactor designs

  5. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    Energy Technology Data Exchange (ETDEWEB)

    Houry, M., E-mail: Michael.houry@cea.fr [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H. [CEA-IRFM, F-13108 Saint-Paul-Lez-Durance (France); Kammerer, N.; Measson, Y. [CEA, LIST, F-92265 Fontenay-aux-Roses (France); Carrel, F.; Schoepff, V. [CEA, LIST, F-91191 Gif-sur-Yvette (France)

    2011-10-15

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  6. Diagnostics carried by a light multipurpose deployer for vacuum vessel interventions

    International Nuclear Information System (INIS)

    Houry, M.; Gargiulo, L.; Balorin, C.; Bruno, V.; Keller, D.; Roche, H.; Kammerer, N.; Measson, Y.; Carrel, F.; Schoepff, V.

    2011-01-01

    ITER will greatly rely on remote-handling operations to accomplish its scientific missions. Robotic systems will also be required to operate inside vacuum vessels in order to limit or replace human access, to intervene quickly between experimental sessions for in-vessel inspections and measurements, and to preserve the machine conditioning and thus improve machine availability. In this prospect, a multipurpose carrier prototype called Articulated Inspection Arm (AIA) was developed by CEA laboratories within the European work program. With an embedded camera, it successfully demonstrated close inspection feasibility inside Tore Supra tokamak. The AIA robot was designed for mini-invasive operations with interchangeable diagnostics to be plugged at its head. This covers various applications for the safety, the operation and the scientific mission (in-vessel inspection, plasma diagnostics calibrations or inner components analysis and treatments). This paper presents recent analysis and results obtain with diagnostics developed by CEA for in-vessel remote-handling intervention.

  7. Slurry Ice as a Cooling System on 30 GT Fishing Vessel

    Directory of Open Access Journals (Sweden)

    Alam Baheramsyah

    2017-06-01

    Full Text Available Indonesia is the largest archipelago country in the world that has a sea area that is very spacious. Indonesian sea area is 5.8 million square kilometers and a coastline of 95 181 km has huge potential in the fisheries sector. In line with the need to further improve on the quality of the fish catch. One way to preserve fish is to use a slurry of ice. Slurry ice proved more effective preserving fishery products instead of using ice cubes. Ice slurry cooling system was designed and applied to the fishing vessel 30 GT. The cooling system uses a simple vapor compression system consists of five major components consisting of evaporator, condenser, compressor, and two pumps.In designing this system determined the type of refrigerant used in advance which type of refrigerant R-507a. Then do the design or selection of its main components. The design is only done on the evaporator. As for the other major components such as condensers, compressors, and pumps election in accordance with the specification of the power needed. After that dialakukan depiction of each system component. Then subsequently designing the laying of ice slurry cooling system components on a fishing vessel 30 GT.            Through calculations using simple thermodynamic equations obtained cooling load on this system amounted to 32.06 kW. Condenser with a power of 40 kW. Compressor with power 12 kW. Pump with capacity 10 m3 / h. With memepertimbangkan space left on the ship in the ice slurry system design on the main deck of the ship to the efficient use of space on board. The power requirements of the generator vessel increases due to the addition of ice slurry system components therefore do replacement generator into the generator with a power of 100 kW and penambahn fuel tank to 6,000 L.

  8. Proceedings International Workshop on Formal Engineering approaches to Software Components and Architectures

    OpenAIRE

    Kofroň, Jan; Tumova, Jana

    2017-01-01

    These are the proceedings of the 14th International Workshop on Formal Engineering approaches to Software Components and Architectures (FESCA). The workshop was held on April 22, 2017 in Uppsala (Sweden) as a satellite event to the European Joint Conference on Theory and Practice of Software (ETAPS'17). The aim of the FESCA workshop is to bring together junior researchers from formal methods, software engineering, and industry interested in the development and application of formal modelling ...

  9. 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)

  10. Eddy currents in a nonperiodic vacuum vessel induced by axisymmetric plasma motion

    International Nuclear Information System (INIS)

    DeLucia, J.

    1985-12-01

    A method is described for calculating the two-dimensional trajectory of a vertically or horizontally unstable axisymmetric tokamak plasma in the presence of a resistive vacuum vessel. The vessel is not assumed to have toroidal symmetry. The plasma is represented by a current-filament loop that is free to move vertically and to change its major radius. Its position is evolved in time self-consistently with the vacuum vessel eddy currents. The plasma current, internal inductance, and poloidal beta can be specified functions of time so that eddy currents resulting from a disruption can be modeled. The vacuum vessel is represented by a set of current-filaments whose positions and orientations are chosen to model the dominant eddy current paths. Although the specific application is to TFTR, the present model is of general applicability. 7 refs., 4 figs., 2 tabs

  11. Bounding the conservatism in flaw-related variables for pressure vessel integrity analyses

    International Nuclear Information System (INIS)

    Foulds, J.R.; Kennedy, E.L.

    1993-01-01

    The fracture mechanics-based integrity analysis of a pressure vessel, whether performed deterministically or probabilistically, requires use of one or more flaw-related input variables, such as flaw size, number of flaws, flaw location, and flaw type. The specific values of these variables are generally selected with the intent to ensure conservative predictions of vessel integrity. These selected values, however, are largely independent of vessel-specific inspection results, or are, at best, deduced by ''conservative'' interpretation of vessel-specific inspection results without adequate consideration of the pertinent inspection system performance (reliability). In either case, the conservatism associated with the flaw-related variables chosen for analysis remains examination (NDE) technology and the recently formulated ASME Code procedures for qualifying NDE system capability and performance (as applied to selected nuclear power plant components) now provides a systematic means of bounding the conservatism in flaw-related input variables for pressure vessel integrity analyses. This is essentially achieved by establishing probabilistic (risk)-based limits on the assigned variable values, dependent upon the vessel inspection results and on the inspection system unreliability. Described herein is this probabilistic method and its potential application to: (i) defining a vessel-specific ''reference'' flaw for calculating pressure-temperature limit curves in the deterministic evaluation of pressurized water reactor (PWR) reactor vessels, and (ii) limiting the flaw distribution input to a PWR reactor vessel-specific, probabilistic integrity analysis for pressurized thermal shock loads

  12. Repair and replacement of reactor internals for plant life extension

    International Nuclear Information System (INIS)

    Graae, T.

    1998-01-01

    Recent experience from early Swedish BWRs corroborate that all components in a nuclear power plant can be repaired or replaced with new ones. Oskarshamn 1 has gone through a thorough refurbishment project. A number of internals were repaired or replaced including the core shroud support which was welded to the bottom of the reactor pressure vessel. The project verifies that it is fully possible to carry out complicated inspection and repair work inside a nuclear pressure vessel which has been in operation for more than 20 years. Along with increased capacity factor, operating nuclear power plants get the financial conditions needed for extensive repair and modernization projects. Large power output leads to short pay-back times for the investments. The FENIX project at Oskarshamn 1 is such a project. There are utilities whose policy is to keep their plants in as-new condition for an unlimited length of time. (orig.)

  13. Vessel Biofouling Prevention and Management Options Report

    Science.gov (United States)

    2015-03-01

    microorganism -containing biofilm (slime layer) which occurs on a marine surface under the right conditions. Biofilm growth can progress to a point where... biodegradable components, or non-biocidal alternatives; vessels that reside or operate >30 days in copper impaired ports or harbors shall consider alternatives...except for nontoxic, biodegradable types) or chemicals in the water is prohibited, as would be the creation of a cloud or plume generated by paint

  14. 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)

  15. 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.

  16. 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%.

  17. 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.)

  18. 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)

  19. Conceptual design of EAST flexible in-vessel inspection system

    International Nuclear Information System (INIS)

    Peng, X.B.; Song, Y.T.; Li, C.C.; Lei, M.Z.; Li, G.

    2010-01-01

    Remote handling technology, especially the flexible in-vessel inspection system (FIVIS) without breaking the working condition of the vacuum vessel, has been identified as one major challenge on the maintenance for the future tokamak fusion reactor. The FIVIS introduced here is specially developed for EAST superconducting tokamak that has actively cooled plasma facing components (PFCs). It aims flexible close-up inspection of EAST PFCs to help the understanding of operation issues that could occur in the vacuum vessel. This paper resumes the preliminary work of the FIVIS project, including the requirement analysis and the development of the conceptual design. The FIVIS consists out of a long reach multi-articulated manipulator and a process tool. The manipulator has a modular design for its subsystems and can reach all areas of the first wall in the distance of 15 mm and in the range of ±90 o along toroidal direction. It will be folded and hidden in the designated horizontal port during plasma discharge period.

  20. 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.

  1. 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

  2. Electro-mechanical connection system for ITER in-vessel magnetic sensors

    Energy Technology Data Exchange (ETDEWEB)

    Rizzolo, Andrea; Brombin, Matteo; Gonzalez, Winder [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Marconato, Nicolò, E-mail: nicolo.marconato@igi.cnr.it [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Peruzzo, Simone [Consorzio RFX, Corso Stati Uniti, 4, 35127 Padova (Italy); Arshad, Shakeib [Fusion for Energy, C/Josep Pla, 2, 08019 Barcelona (Spain); Ma, Yunxing; Vayakis, George [ITER Organization, Route de Vinon-sur-Verdon, 13067 St Paul Lez Durance (France); Williams, Adrian [Oxford Technologies Ltd, 7 Nuffield Way, Abingdon, Oxon, OX14 1RL (United Kingdom)

    2016-11-01

    Highlights: • Latest status of the ITER “Generic In-Vessel Magnetic Platform” design activity. • Integration within the ITER In-Vessel configuration model. • Geometry optimization based on thermo-mechanical and magnetic field 3D calculation. • Assessment of the remote handling maintenance compatibility. - Abstract: This paper presents the preliminary design of the “In-Vessel Magnetic platform”, which is a subsystem of the magnetic diagnostics formed by all the components necessary for guaranteeing the thermo-mechanical interface of the actual magnetic sensors with the vacuum vessel (VV), their protection and the electrical connection to the in-vessel wiring for the transmission of the detected signal with a minimum level of noise. The design has been developed in order to comply with different functional requirements: the mechanical attachment to the VV; the electrical connection to the in-vessel wiring; efficient heat transfer to the VV; the compatibility with Remote Handling (RH) system for replacement; the integration of metrology features for post-installation control; the Electro Magnetic Interference (EMI) shielding from Electron Cyclotron Heating (ECH) stray radiation without compromising the sensor pass band (15 kHz). Significant effort has been dedicated to develop the CAD model, integrated within the ITER In-Vessel configuration model, taking care of the geometrical compliance with the Blanket modules (modified in order to accommodate the magnetic sensors in suitable grooves) and the RH compatibility. Thorough thermo-mechanical and electro-magnetic Finite Element Method (FEM) analyses have been performed to assess the reliability of the system in standard and off-normal operating conditions for the low frequency magnetic sensors.

  3. Mural cell associated VEGF is required for organotypic vessel formation.

    Directory of Open Access Journals (Sweden)

    Lasse Evensen

    Full Text Available BACKGROUND: Blood vessels comprise endothelial cells, mural cells (pericytes/vascular smooth muscle cells and basement membrane. During angiogenesis, mural cells are recruited to sprouting endothelial cells and define a stabilizing context, comprising cell-cell contacts, secreted growth factors and extracellular matrix components, that drives vessel maturation and resistance to anti-angiogenic therapeutics. METHODS AND FINDINGS: To better understand the basis for mural cell regulation of angiogenesis, we conducted high content imaging analysis on a microtiter plate format in vitro organotypic blood vessel system comprising primary human endothelial cells co-cultured with primary human mural cells. We show that endothelial cells co-cultured with mural cells undergo an extensive series of phenotypic changes reflective of several facets of blood vessel formation and maturation: Loss of cell proliferation, pathfinding-like cell migration, branching morphogenesis, basement membrane extracellular matrix protein deposition, lumen formation, anastamosis and development of a stabilized capillary-like network. This phenotypic sequence required endothelial-mural cell-cell contact, mural cell-derived VEGF and endothelial VEGFR2 signaling. Inhibiting formation of adherens junctions or basement membrane structures abrogated network formation. Notably, inhibition of mural cell VEGF expression could not be rescued by exogenous VEGF. CONCLUSIONS: These results suggest a unique role for mural cell-associated VEGF in driving vessel formation and maturation.

  4. Prevention of catastrophic failure in pressure vessels and pipings

    International Nuclear Information System (INIS)

    Rintamaa, R.; Wallin, K.; Ikonen, K.; Toerroenen, K.; Talja, H.; Keinaenen, H.; Saarenheimo, A.; Nilsson, F.; Sarkimo, M.; Waestberg, S.; Debel, C.

    1989-01-01

    The fracture resistance and integrity of pressure-loaded components have been assessed in a Nordic research programme. Experiments were performed to validate the computational fracture assessment analysis. Two tests were also conducted on a large decommissioned pressure vessel from an oil refinery plant. Different fracture assessment methods were developed and subsequently applied to the tested components. Interlaboratory round robin programmes with the participation of several laboratories were arranged to examine elastic-plastic finit element calculations and fracture mechanics testing. The transferability of material parameters derived from small specimens with simple crack geometries to more realistic crack geometries in real components has been verified. (author)

  5. Intercrystalline internal adsorption in systems with limiting solubility of components

    International Nuclear Information System (INIS)

    Krysova, S.K.; Stepanova, V.A.; Mozgovoj, M.V.

    1979-01-01

    The decrease of the excessive energy of the intercrystalline boundary of ion by additions of transitional elements having unlimited solubility in iron has been studied. The data obtained agree with the results of an earlier work based on materials of a less higher initial purity. For the systems studied (Fe-V, Fe-Cr, Fe-Mn, Fe-Ni) the degree of the intercrystalline interval adsorption is independent of either the annealing temperature or the cooling method. This corresponds to the notion of the relationship between the intercrystalline internal adsorption and the cubic solubility of the addition in a given solvent. For pure ion, a weak temperature dependence of the excessive energy of the intercrystalline boundries was found in the lower section of the examined temperature range. The constants of cumulative recrystallization of the alloys studied in α-phase show a stepwise dependence upon the atomic number of the solute component, what indicates the relationship between the cumulative recrystallization and the intercrystalline internal adsorption. A monotonous decrease of the constant of cumulative recrystallization is observed for the same alloys in α-phase, on both sides of iron

  6. Mechanical properties of reactor pressure vessel steels studied by static and dynamic torsion tests

    International Nuclear Information System (INIS)

    Munier, A.; Maamouri, M.; Schaller, R.; Mercier, O.

    1993-01-01

    Internal friction measurements and torsional plastic deformation tests have been performed in reactor pressure vessel steels (unirradiated, irradiated and irradiated/annealed specimens). The results of these experiments have been interpreted with help of transmission electron microscopy observations (conventional and in situ). It is shown how the interactions between screw dislocations and obstacles (Peierls valleys, impurities and precipitates) could explain the low temperature hardening and the irradiation embrittlement of ferritic steels. In addition, it appears that the nondestructive internal friction technique could be used advantageously to follow the evolution of the material properties under irradiation, as for instance the irradiation embrittlement of the reactor pressure vessel steels. (orig.)

  7. 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.

  8. High temperature helium test rig with prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schmidl, H.

    1975-10-01

    The report gives a short description of the joint project prestressed concrete vessel-helium test station as there is the building up of the concrete structure, the system of instrumentation, the data processing, the development of the helium components as well as the testing programs. (author)

  9. Development of cold moderator vessel for the spallation neutron source. Flow field measurements and thermal hydraulic analyses in cold moderator vessel

    International Nuclear Information System (INIS)

    Aso, Tomokazu; Kaminaga, Masanori; Terada, Atsuhiko; Hino, Ryutaro

    2001-01-01

    The Japan Atomic Energy Research Institute is developing a several MW-scale spallation target system under the High-Intensity Accelerator Project. A cold moderator using supercritical hydrogen is one of the key components in the target system, which directly affects the neutronic performance both in intensity and resolution. Since a hydrogen temperature rise in the moderator vessel affects the neutronic performance, it is necessary to suppress the recirculation and stagnant flows which cause hot spots. In order to develop the conceptual design of the moderator structure in progress, the flow field was measured using a PIV (Particle Image Velocimetry) system under water flow conditions using a flat model that simulated a moderator vessel. From these results, the flow field such as recirculation flows, stagnant flows etc. was clarified. The hydraulic analytical results using the standard k-ε model agreed well with experimental results. Thermal-hydraulic analyses in the moderator vessel were carried out under liquid hydrogen conditions. Based on these results, we clarified the possibility of suppressing the local temperature rise within 3 K under 2 MW operating condition. (author)

  10. Cylindrical reinforced-concrete pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Vaessen, F.

    1975-01-01

    The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

  11. Draft I.E.C. standard for monitoring PWR internal structures; Projet de norme C.E.I. pour la surveillance des structures internes des REP

    Energy Technology Data Exchange (ETDEWEB)

    Trenty, A.

    1994-06-01

    EDF has proposed to the International Electrotechnical Commission a draft standard for monitoring the vessel internal structures of PWRs. The standard applies to systems used for monitoring the vibratory behavior of the internal structures of PWRs (core barrel, thermal shield, fuel assemblies) on the basis of neutron fluctuations observed outside the vessel as well as of vessel vibrations. It covers the systems characteristics and the monitoring procedures. It should facilitate standardization of monitoring and comparisons on an international level. This paper presents the main features of the draft standard: -principles of measurement: correlation between movements of internals and ex core neutron noise on the one hand, forced vibrations of the vessel on the other hand; -sampling and conditioning of the signals; -monitoring equipment and in particular spectral analysis device; -functions of the monitoring software used for spectral analysis, peak detection and calculation of structure displacement; -studies preliminary to setting up the monitoring (calculation of internal vibratory modes, defect simulation on mockup, qualification on reactor during hot test...); -monitoring procedures (periodicity of analysis and what to do in case of anomaly); -documentation necessary to the monitoring. A diagnostic procedure is given as an example. The draft standard, written in 1994, will be presented in Frankfurt (Germany) in February 1995. (author). 1 annexe.

  12. Stresses in transition region of VVER-1000 reactor vessels

    International Nuclear Information System (INIS)

    Namgung, I.; Nguye, T.L.

    2014-01-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  13. Stresses in transition region of VVER-1000 reactor vessels

    Energy Technology Data Exchange (ETDEWEB)

    Namgung, I. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Nguye, T.L. [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); National Research Inst. of Mechanical Engineering, Hanoi City, Vietnam (China)

    2014-07-01

    Most of the western PWR reactor's bottom head is hemi-spherical shape, however for Russian designed VVER family of reactor it is ellipsoidal shape. The transition region from shell side to ellipsoidal head and transition top flange to cylindrical shell develop higher stress concentration than western PWR reactor vessel. This region can be modeled as conical shell with varying thickness. The theoretical derivation of stress in the thick-walled conical cylinder with varying thickness was developed and shown in detail. The results is applied to VVER-1000 reactor vessel of which shell to bottom ellipsoidal shell and shell to upper flange were investigated for stress field. The theoretical calculations were also compared with FEM solutions. An axisymmetric 3D model of VVER-1000 reactor vessel (without closure head) FEM model was created and internal hydrostatic pressure boundary condition was applied. The stress results from FEM and theoretical calculation were compared, and the discrepancies and accuracies of the theoretical results were discussed. (author)

  14. 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.)

  15. Analyses for passive safety of fusion reactor during ex-vessel loss of coolant accident

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Maki, Koichi; Uda, Tatuhiko; Seki, Yasushi; Aoki, Isao; Kunugi, Tomoaki.

    1995-01-01

    Passive safety of nuclear fusion reactors during ex-vessel Loss-of-Coolant Accidents (LOCAs) in the divertor cooling system has been investigated using a hybrid code, which can treat the interaction of the plasma and plasma facing components (PFCs). The code has been modified to include the impurity emission from PFCs with a diffusion model at the edge plasma. We assumed an ex-vessel LOCA of the divertor cooling system during the ignited operation in International Thermonuclear Experimental Reactor (ITER), in which a carbon-copper brazed divertor plate was employed in the Conceptual Design Activity (CDA). When a double-ended break occurs at the cold leg of the divertor cooling system, the impurity density in the main plasma becomes about twice within 2s after the LOCA due to radiation enhanced sublimation of graphite PFCs. The copper cooling tube of the divertor begins to melt at about 3s after the LOCA, even though the plasma is passively shut down at about 4s due to the impurity accumulation. It is necessary to apply other PFC materials, which can shorten the time period for passive shutdown, or an active shutdown system to keep the reactor structures intact for such rapid transient accident. (author)

  16. 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

  17. 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)

  18. 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.

  19. Development of the Inspection and Diagnosis Technology for the NSSS Components Integrity

    International Nuclear Information System (INIS)

    Kim, Jae Hee; Eom, Heung Soup; Lee, Jae Cheol

    2007-02-01

    This project is to develop and demonstrate new technologies for a monitoring, inspection, diagnosis and evaluation of the safety related components in nuclear power plants. These technologies are required to detect the defects in the components of nuclear power plants and to prepare thoroughly against accidents. We studied on the four issues recently focused. Thus we developed an impact analysis model of the reactor and steam generator, and diagnosis software of the reactor internals. As an on-line monitoring technology using an ultrasonic guided wave, we developed a new method enhancing the S/N ratio of the weak signal based on time reversal technique. A network based remote inspection system and an inspection robot for reactor vessel head penetration was developed. We also performed a lifetime estimation and degradation analysis of the NPP cables through accelerated degradation tests. The technologies developed in this project are applied to the components of nuclear power plants. The applications include a localization of the NSSS integrity monitoring system, replacement of an in-service inspection by on-line monitoring, remote inspection of the major components of the plants, lifetime estimation of the degraded plant cables, and so on

  20. Development of the Inspection and Diagnosis Technology for the NSSS Components Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Hee; Eom, Heung Soup; Lee, Jae Cheol (and others)

    2007-02-15

    This project is to develop and demonstrate new technologies for a monitoring, inspection, diagnosis and evaluation of the safety related components in nuclear power plants. These technologies are required to detect the defects in the components of nuclear power plants and to prepare thoroughly against accidents. We studied on the four issues recently focused. Thus we developed an impact analysis model of the reactor and steam generator, and diagnosis software of the reactor internals. As an on-line monitoring technology using an ultrasonic guided wave, we developed a new method enhancing the S/N ratio of the weak signal based on time reversal technique. A network based remote inspection system and an inspection robot for reactor vessel head penetration was developed. We also performed a lifetime estimation and degradation analysis of the NPP cables through accelerated degradation tests. The technologies developed in this project are applied to the components of nuclear power plants. The applications include a localization of the NSSS integrity monitoring system, replacement of an in-service inspection by on-line monitoring, remote inspection of the major components of the plants, lifetime estimation of the degraded plant cables, and so on.

  1. 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.

  2. ITER nuclear components, preparing for the construction and R&D results

    Science.gov (United States)

    Ioki, K.; Akiba, M.; Barabaschi, P.; Barabash, V.; Chiocchio, S.; Daenner, W.; Elio, F.; Enoeda, M.; Ezato, K.; Federici, G.; Gervash, A.; Grebennikov, D.; Jones, L.; Kajiura, S.; Krylov, V.; Kuroda, T.; Lorenzetto, P.; Maruyama, S.; Merola, M.; Miki, N.; Morimoto, M.; Nakahira, M.; Ohmori, J.; Onozuka, M.; Rozov, V.; Sato, K.; Strebkov, Yu; Suzuki, S.; Tanchuk, V.; Tivey, R.; Utin, Yu

    2004-08-01

    Progress has been made in the preparation of the procurement specifications for key nuclear components of ITER. Detailed design of the vacuum vessel (VV) and in-vessel components is being performed to consider fabrication methods and non-destructive tests (NDT). R&D activities are being carried out on vacuum vessel UT inspection with waves launched at an angle of 20° or 30°, on flow distribution tests of a two-channel model, on fabrication and testing of FW mock-ups and panels, on the blanket flexible support as a complete system including the housing, on the blanket co-axial pipe connection with guard vacuum for leak detection, and on divertor vertical target prototypes. The results give confidence in the validity of the design and identify possibilities of attractive alternate fabrication methods.

  3. Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

    International Nuclear Information System (INIS)

    Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

    2013-06-01

    To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

  4. 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.)

  5. 49 CFR 173.220 - Internal combustion engines, self-propelled vehicles, mechanical equipment containing internal...

    Science.gov (United States)

    2010-10-01

    ... and vehicles with certain electronic equipment when transported by aircraft or vessel. When an... vehicles, mechanical equipment containing internal combustion engines, and battery powered vehicles or... Than Class 1 and Class 7 § 173.220 Internal combustion engines, self-propelled vehicles, mechanical...

  6. Revisiting the reactor pressure vessel for long-time operation

    International Nuclear Information System (INIS)

    Lapena, J.; Serrano, M.; Diego, G. de; Hernandez Mayoral, M.

    2013-01-01

    The reactor pressure vessel (RPV) is one of the key components of nuclear power plants, especially for long time operation. It is a non-replaceable component, at least with current technology. the structural integrity of the vessel is evaluated within called monitoring programs where the degradation of the mechanical properties due to neutron irradiation is determined. From the first designs of the RPVs and monitoring programs in the years 60-70 currently still in force, there have been major advances in the understanding of radiation damage and methods of evaluation. Thus, it is recommended the use of forgings instead of plates in the construction of the RPVs in order to reduce the number of welds, more sensitive to neutron irradiation, and using starting materials with less content of impurities, particularly copper. To evaluate the embrittlement of RPVs the Master Curve methodology is currently used, through the testing of the charpy specimens from the surveillance capsules, to determine the fracture toughness. This article summarizes the last activities of CIEMAT into the European research projects LONGIIFE and PERFORM60, about the knowledge of radiation damage in materials with low copper content, traditionally considered less sensitive to irradiation, and the use of the Master Curve in advanced surveillance programs. The activities related to the problems associated with the use of large forging, such as the appearance of hydrogen flakes in the vessel of Doel 3, and its implications, are also presented. (Author)

  7. PWR upper/lower internals shield

    Energy Technology Data Exchange (ETDEWEB)

    Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  8. 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...

  9. Design and structural analysis of support structure for ITER vacuum vessel

    International Nuclear Information System (INIS)

    Takeda, Nobukazu; Ohmori, Junji; Nakahira, Masataka; Shibanuma, Kiyoshi

    2004-01-01

    The International Thermonuclear Experimental Reactor (ITER) vacuum vessel (VV) is a safety component confining radioactive materials such as tritium and activated dust. An independent VV support structure with multiple flexible plates located at the bottom of VV lower port is proposed as a new concept, which is deferent from the current design, i.e., the VV support is directly connected to the toroidal coils (TF coils). This independent concept has two advantages comparing to the current one: (1) thermal load due to the temperature deference between VV and TF coils becomes lower and (2) the TF coils are categorized as non-safety components because of its independence from VV. Stress Analyses have been performed to assess the integrity of the VV support structure using a precisely modeled VV structure. As a result, (1) the maximum displacement of the VV corresponding to the relative displacement between VV and TF coils is found to be 15 mm, much less than the current design clearance of 100 mm, and (2) the stresses of the whole VV system including VV support are estimated to be less than the allowable ones defined by ASME Section III Subsection NF, respectively. Based on these assessments, the feasibility of the proposed independent VV support has been verified as a VV support. (author)

  10. In-service inspection program for the NCS-80 reactor pressure vessel

    International Nuclear Information System (INIS)

    Scharge, J.; Wehowsky, P.; Zeibig, H.

    1978-01-01

    The in-service inspection program of reactor pressure vessels is mainly based on the ultra-sonic method, visual checking of inner and outer surfaces as well as pressure and leak tests. The test procedure require a design of the pressure vessel suitable for the test methods and the possibility to remove the pressure vessel internals. For the outside inspection a gap of sufficient width is mandatory. The present status of the ultra-sonic method and of the inner and outer manipulators affords to conduct the in-service inspection program in form of automatic checkings. The in-service inspection program for NCS-80, the Nuclear Container-Ship design of 80,000 shp, is integrated in the refueling periods due to the request for a high availability of the ship and reactor plant

  11. Heat up and potential failure of BWR upper internals during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

  12. Proceedings of the 17th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    2003-01-01

    The conference was divided into the following divisions and subdivisions: DIVISION A: Plenary lectures and panel; DIVISION B: Computational mechanics (Structural and thermal analysis; High-non linear analysis, material behaviour; Vibration and fluid dynamics analysis); DIVISION C: Fuel and core structures (Fuel vibration and fretting; Fuel design and constitutive modelling; Fuel failure under operation and accident conditions; Fuel failure under operation and accident conditions; Components and material behaviour under irradiation; Integrity of fuel systems under transient conditions); DIVISION D: Aging, Life Extension and Licence Renewal (International Regulatory and Economic Perspectives; Utility perspectives, WWER technology; Fatigue, corrosion and crack issues; Component integrity; Aging assessment and monitoring; Containment and other structures); DIVISION F: Design methods and rules for components (International codes and standards; Tube, piping codes and standards; Analyses; Fatigue and life assessment; Creep; Bolted connections and gaskets); DIVISION G: Fracture mechanics (Reactor pressure vessel integrity; Dynamic loading; Fracture considerations for various applications; Failure assessment of Zr alloy; Pipe integrity; Integrity of welds; Failure of non-metallic materials; Leak before break (LBB); Corrosion aspects); DIVISION H: Concrete Containment and Other Structures (Concrete materials and performance; Tests of scale prestressed concrete containment vessel; Shear wall test and analysis; Structural analysis and containment design; Structural integrity and analysis); DIVISION J: Analysis and design for dynamic and extreme load (Vibration of shells and plates; Impact analysis; Piping vibration; Structural dynamics; Experimental and other topics); DIVISION K: Seismic analysis, design and qualification (General seismic issues; Ground motion and sitting; Soil-structure interaction; Seismic response of structures; Seismic re-evaluation; Seismic response and

  13. Inspection and repair of nuclear components

    International Nuclear Information System (INIS)

    Lahner, K.; Poetz, F.

    1993-01-01

    Despite careful design, manufacturing and operation, some of the important safety-relevant components show deterioration with time. Because of activation and contamination of these components, their inspection and repair has to be performed with manipulators. Some sophisticated manipulators are described, built by ABB Reaktor and used for inspection, maintenance and repair of PWR steam generators, fuel alignment pins, core baffle former bolts and reactor pressure vessel head penetrations. (Z.S.) 7 figs

  14. 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.)

  15. Current meter components and other data from XCP casts from VARIOUS SMALL VESSELS and other platforms from the North Atlantic Ocean as part of the OCEAN DUMPING and other projects from 01 December 1990 to 01 June 1991 (NODC Accession 9300076)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Current meter components and other data were collected from XCP casts from VARIOUS SMALL VESSELS and other platforms in the North Atlantic Ocean. Data were collected...

  16. Preventive protection device and method for bottom of reactor pressure vessel

    International Nuclear Information System (INIS)

    Hayashi, Eisaku; Kurosawa, Koichi; Furukawa, Hideyasu; Morinaka, Ren; Enomoto, Kunio; Otaka, Masahiro; Yoshikubo, Fujio; Chiba, Noboru; Sato, Kazunori.

    1995-01-01

    In a preventive protection device for improving stresses in reactor structural components by jetting highly pressurized water with cavitation bubbles from a jetting nozzle toward structural components in a reactor pressure vessel, a fixed structure to a CRD housing is provided with a rotational body attached to the structure, a multi joint arm and a jetting nozzle supported to the multi joint arm. The jetting nozzle is disposed at a position where the center of the jetting deviates from the center of the CRD housing. In addition, a monitoring camera is disposed for displaying the target for preventive protection. The state of stresses on a plurality of targets for preventive protection can be improved by the preventive protection device at a fixed position in the bottom of a reactor pressure vessel where housings stand densely, thereby enabling to attain the preventive protection operation easily and rapidly. (N.H.)

  17. 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

  18. Neutrophil-Mediated Delivery of Therapeutic Nanoparticles across Blood Vessel Barrier for Treatment of Inflammation and Infection.

    Science.gov (United States)

    Chu, Dafeng; Gao, Jin; Wang, Zhenjia

    2015-12-22

    Endothelial cells form a monolayer in lumen of blood vessels presenting a great barrier for delivery of therapeutic nanoparticles (NPs) into extravascular tissues where most diseases occur, such as inflammation disorders and infection. Here, we report a strategy for delivering therapeutic NPs across this blood vessel barrier by nanoparticle in situ hitchhiking activated neutrophils. Using intravital microscopy of TNF-α-induced inflammation of mouse cremaster venules and a mouse model of acute lung inflammation, we demonstrated that intravenously (iv) infused NPs made from denatured bovine serum albumin (BSA) were specifically internalized by activated neutrophils, and subsequently, the neutrophils containing NPs migrated across blood vessels into inflammatory tissues. When neutrophils were depleted using anti-Gr-1 in a mouse, the transport of albumin NPs across blood vessel walls was robustly abolished. Furthermore, it was found that albumin nanoparticle internalization did not affect neutrophil mobility and functions. Administration of drug-loaded albumin NPs markedly mitigated the lung inflammation induced by LPS (lipopolysaccharide) or infection by Pseudomonas aeruginosa. These results demonstrate the use of an albumin nanoparticle platform for in situ targeting of activated neutrophils for delivery of therapeutics across the blood vessel barriers into diseased sites. This study demonstrates our ability to hijack neutrophils to deliver nanoparticles to targeted diseased sites.

  19. Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

    Science.gov (United States)

    Krasikov, E.

    2015-04-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation. There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment. The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible.

  20. Radiation intensification of the reactor pressure vessels recovery by low temperature heat treatment (wet annealing)

    International Nuclear Information System (INIS)

    Krasikov, E

    2015-01-01

    As a main barrier against radioactivity outlet reactor pressure vessel (RPV) is a key component in terms of NPP safety. Therefore present-day demands in RPV reliability enhance have to be met by all possible actions for RPV in-service embrittlement mitigation. Annealing treatment is known to be the effective measure to restore the RPV metal properties deteriorated by neutron irradiation.There are two approaches to annealing. The first one is so-called «dry» high temperature (∼475°C) annealing. It allows obtaining practically complete recovery, but requires the removal of the reactor core and internals. External heat source (furnace) is required to carry out RPV heat treatment.The alternative approach is to anneal RPV at a maximum coolant temperature which can be obtained using the reactor core or primary circuit pumps while operating within the RPV design limits. This low temperature «wet» annealing, although it cannot be expected to produce complete recovery, is more attractive from the practical point of view especially in cases when the removal of the internals is impossible. (paper)