WorldWideScience

Sample records for vessel materials research

  1. 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...

  2. Molten material-containing vessel

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko

    1998-01-01

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

  3. Computerized reactor pressure vessel materials information system

    International Nuclear Information System (INIS)

    Strosnider, J.; Monserrate, C.; Kenworthy, L.D.; Tether, C.D.

    1980-10-01

    A computerized information system for storage and retrieval of reactor pressure vessel materials data was established, as part of Task Action Plan A-11, Reactor Vessel Materials Toughness. Data stored in the system are necessary for evaluating the resistance of reactor pressure vessels to flaw-induced fracture. This report includes (1) a description of the information system; (2) guidance on accessing the system; and (3) a user's manual for the system

  4. Flexible Composite-Material Pressure Vessel

    Science.gov (United States)

    Brown, Glen; Haggard, Roy; Harris, Paul A.

    2003-01-01

    A proposed lightweight pressure vessel would be made of a composite of high-tenacity continuous fibers and a flexible matrix material. The flexibility of this pressure vessel would render it (1) compactly stowable for transport and (2) more able to withstand impacts, relative to lightweight pressure vessels made of rigid composite materials. The vessel would be designed as a structural shell wherein the fibers would be predominantly bias-oriented, the orientations being optimized to make the fibers bear the tensile loads in the structure. Such efficient use of tension-bearing fibers would minimize or eliminate the need for stitching and fill (weft) fibers for strength. The vessel could be fabricated by techniques adapted from filament winding of prior composite-material vessels, perhaps in conjunction with the use of dry film adhesives. In addition to the high-bias main-body substructure described above, the vessel would include a low-bias end substructure to complete coverage and react peak loads. Axial elements would be overlaid to contain damage and to control fiber orientation around side openings. Fiber ring structures would be used as interfaces for connection to ancillary hardware.

  5. Pressure Vessel Steel Research: Belgian Activities

    International Nuclear Information System (INIS)

    Van Walle, E.; Fabry, A.; Ait Abderrahim, H.; Chaouadi, R.; D'hondt, P.; Puzzolante, J.L.; Van de Velde, J.; Van Ransbeeck, T.; Gerard, R.

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly

  6. Pressure Vessel Steel Research: Belgian Activities

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E; Fabry, A; Ait Abderrahim, H; Chaouadi, R; D` hondt, P; Puzzolante, J L; Van de Velde, J; Van Ransbeeck, T [Centre d` Etude de l` Energie Nucleaire, Mol (Belgium); Gerard, R [TRACTEBEL, Brussels (Belgium)

    1994-03-01

    A review of the Belgian research activities on Nuclear Reactor Pressure Vessel Steels (RPVS) and on related Neutron Dosimetry Aspects is presented. Born out of the surveillance programmes of the Belgian nuclear power plants, this research has lead to the development of material saving techniques, like reconstitution and miniaturization, and to improved neutron dosimetry techniques. A physically- justified RPVS fracture toughness indexation methodology, supported by micro-mechanistic modelling, is based on the elaborate use of the instrumented Charpy impact signal. Computational tools for neutron dosimetry allow to reduce the uncertainties on surveillance capsule fluences significantly.

  7. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2001-01-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  8. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  9. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    Scheuer, A.; Gutsmiedl, E.

    1999-01-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256 deg. C and 250 deg. C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was take into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256 deg. C and 150 deg. C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to take into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼ 1x10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture

  10. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    Gutsmiedl, Erwin

    2001-01-01

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1·10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture criteria of

  11. Crack growth rates in vessel head penetration materials

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Blazquez, F.

    1994-01-01

    The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

  12. Multilayer Pressure Vessel Materials Testing and Analysis. Phase 1

    Science.gov (United States)

    Cardinal, Joseph W.; Popelar, Carl F.; Page, Richard A.

    2014-01-01

    To provide NASA a comprehensive suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for aging multilayer pressure vessels, Southwest Research Institute (R) (SwRI) was contracted in two phases to obtain relevant material property data from a representative vessel. This report describes Phase 1 of this effort which includes a preliminary material property assessment as well as a fractographic, fracture mechanics and fatigue crack growth analyses of an induced flaw in the outer shell of a representative multilayer vessel that was subjected to cyclic pressure test. SwRI performed this Phase 1 effort under contract to the Digital Wave Corporation in support of their contract to Jacobs ATOM for the NASA Ames Research Center.

  13. Multilayer Pressure Vessel Materials Testing and Analysis Phase 2

    Science.gov (United States)

    Popelar, Carl F.; Cardinal, Joseph W.

    2014-01-01

    To provide NASA with a suite of materials strength, fracture toughness and crack growth rate test results for use in remaining life calculations for the vessels described above, Southwest Research Institute® (SwRI®) was contracted in two phases to obtain relevant material property data from a representative vessel. An initial characterization of the strength, fracture and fatigue crack growth properties was performed in Phase 1. Based on the results and recommendations of Phase 1, a more extensive material property characterization effort was developed in this Phase 2 effort. This Phase 2 characterization included additional strength, fracture and fatigue crack growth of the multilayer vessel and head materials. In addition, some more limited characterization of the welds and heat affected zones (HAZs) were performed. This report

  14. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  15. Reactor Materials Research

    International Nuclear Information System (INIS)

    Van Walle, E.

    2002-01-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel

  16. Materials surveillance program for C-E NSSS reactor vessels

    International Nuclear Information System (INIS)

    Koziol, J.J.

    1977-01-01

    Irradiation surveillance programs for light water NSSS reactor vessels provide the means by which the utility can assess the extent of neutron-induced changes in the reactor vessel materials. These programs are conducted to verify, by direct measurement, the conservatism in the predicted radiation-induced changes and hence the operational parameters (i.e., heat-up, cooldown, and pressurization rates). In addition, such programs provide assurance that the scheduled adjustments in the operational parameters are made with ample margin for safe operation of the plant. During the past 3 years, several documents have been promulgated establishing the criteria for determining both the initial properties of the reactor vessel materials as well as measurement of changes in these initial properties as a result of irradiation. These documents, ASTM E-185-73, ''Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels,'' and Appendix H to 10 CFR 50, ''Reactor Vessel Material Surveillance Program Requirements,'' are complementary to each other. They are the result of a change in the basic philosophy regarding the design and analysis of reactor vessels. In effect, the empirical ''transition temperature approach,'' which was used for design, was replaced by the ''analytical fracture mechanics approach.'' The implementation of this technique was described in Welding Research Council Bulletin 1975 and Appendix G to ASME Code Section III. Further definition of requirements appears in Appendix G to 10 CFR 50 published in July 1973. It is the intent of this paper to describe (1) a typical materials surveillance program for the reactor vessel of a Combustion Engineering NSSS, and (2) how the results of such programs, as well as experimental programs provide feed-back for improvement of materials to enhance their radiation resistance and thereby further improve the safety and reliability of future plants. (author)

  17. Materials for high temperature reactor vessels

    International Nuclear Information System (INIS)

    Buenaventura Pouyfaucon, A.

    2004-01-01

    Within the 5th Euraton Framework Programme, a big effort is being made to promote and consolidate the development of the High Temperature Reactor (HTR). Empresarios Agrupados is participating in this project and among others, also forms part of the HTR-M project Materials for HTRs. This paper summarises the work carried out by Empresarios Agrupados regarding the material selection of the HTR Reactor Pressure Vessel (RPV). The possible candidate materials and the most promising ones are discussed. Design aspects such as the RPV sensitive zones and material damage mechanisms are considered. Finally, the applicability of the existing design Codes and Standards for the design of the HTR RPV is also discussed. (Author)

  18. 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

  19. Storage vessel for containing radiation contaminated material

    International Nuclear Information System (INIS)

    Ogawa, Kazuya.

    1995-01-01

    A container pipe and an outer pipe are coaxially assembled integrally in a state where securing spacers are disposed between the container pipe and the outer pipe, and an annular flow channel is formed around the container pipe. Radiation contaminated material-containing body (glass solidified package) is contained in the container pipe. The container pipe and the outer pipe in an integrated state are suspended from a ceiling plug of a cell chamber of a storage vessel, and supporting devices are assembled between the pipes and a support structure. A shear/lug mechanism is used for the supporting devices. The combination of the shear/lug allows radial and vertical movement but restrict horizontal movement of the outer tube. The supporting devices are assembled while visually recognizing the state of the shear/lug mechanism between the outer pipe and the support mechanism. Accordingly, operationability upon assembling the container pipe and the outer pipe is improved. (I.N.)

  20. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

    Kamino, Yoshikazu; Nishioka, Eiji; Toyota, Michinori.

    1997-01-01

    A containing vessel for radioactive materials (for example, spent fuels) comprises an inner cylinder made of stainless steel having a space for containing radioactive materials at the inside and an outer cylinder made of stainless steel disposed at the outer side of the inner cylinder. Lead homogenization is applied to a space between the inner and the outer cylinders to deposit a lead layer. Then, molten lead heated to a predetermined temperature is cast into the space between the inner and the outer cylinders. A valve is opened to discharge the molten lead in the space from a molten lead discharge pipe, and heated molten lead is injected from a molten lead supply pipe. Then, the discharge of the molten lead and the injection of the molten lead are stopped, and the lead in the space is coagulated. With such procedures, gaps are not formed between the lead of the homogenized portion and the lead of cast portion even when the thickness of the inner and the outer cylinders is great. (I.N.)

  1. Vessel calibration for accurate material accountancy at RRP

    International Nuclear Information System (INIS)

    Yanagisawa, Yuu; Ono, Sawako; Iwamoto, Tomonori

    2004-01-01

    RRP has a 800t·Upr capacity a year to re-process, where would be handled a large amount of nuclear materials as solution. A large scale plant like RRP will require accurate materials accountancy system, so that the vessel calibration with high-precision is very important as initial vessel calibration before operation. In order to obtain the calibration curve, it is needed well-known each the increment volume related with liquid height. Then we performed at least 2 or 3 times run with water for vessel calibration and careful evaluation for the calibration data should be needed. We performed vessel calibration overall 210 vessels, and the calibration of 81 vessels including IAT and OAT were held under presence of JSGO and IAEA inspectors taking into account importance on the material accountancy. This paper describes outline of the initial vessel calibration and calibration results based on back pressure measurement with dip tubes. (author)

  2. Research materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Development of techniques required for the preparation and characterization of ultrahigh-purity and controlled-impurity research specimens of interest to ORNL and other ERDA installations is described

  3. Manufacturing method for radioactive material containing vessel

    International Nuclear Information System (INIS)

    Nishioka, Hideharu; Matsushita, Kazuo; Toyota, Michinori.

    1997-01-01

    Lead homogenization is applied on the inner surface of a space formed between an inner cylinder and an outer cylinder, and a molten lead heated to about 400 to 500degC is cast into a space formed between the inner cylinder and the outer cylinder in a state where the inner and the outer cylinders are heated to from 200 to 300degC. The space formed between the inner cylinder and the outer cylinder is heated to and kept at 330degC or higher for at least 2minutes after the casting of the molten lead, and then it is cooled. Thus, lowering of density of the molten lead due to excess elevation of temperature or dropping of the lead at the homogenization portion by heating the inner and the outer cylinders to an excessively high temperature are not caused. In addition, formation of gaps in the boundary between the inner cylinder and the outer cylinder or between the lead of the homogenized portion and that of the cast portion due to the melting of the lead of the homogenized portion in the space is prevented reliably thereby capable of forming a satisfactory shielding member. Then, even when the thickness of the inner cylinder and the outer cylinder is large, radioactive material containing vessel excellent in heat releasing property and radiation shielding property can be manufactured. (N.H.)

  4. 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

  5. Research program plan: reactor vessels. Volume 1

    International Nuclear Information System (INIS)

    Vagins, M.; Taboada, A.

    1985-07-01

    The ability of the licensing staff of the NRC to make decisions concerning the present and continuing safety of nuclear reactor pressure vessels under both normal and abnormal operating conditions is dependent upon the existence of verified analysis methods and a solid background of applicable experimental data. It is the role of this program to provide both the analytical methods and the experimental data needed. Specifically, this program develops fracture mechanics analysis methods and design criteria for predicting the stress levels and flaw sizes required for crack initiation, propagation, and arrest in LWR pressure vessels under all known and postulated operations conditions. To do this, not only must the methods be developed but they must be experimentally validated. Further, the materials data necessary for input to these analytical methods must be developed. Thus, in addition to methods development and large scale experimental verification this program also develops data to show that slow-load fracture toughness, rapid-load fracture toughness, and crack arrest toughness obtained from small laboratory specimens are truly representative of the toughness characteristics of the material behavior in pressure vessels in both the unirradiated and the irradiated conditions

  6. Processing and analysis techniques involving in-vessel material generation

    Science.gov (United States)

    Schabron, John F [Laramie, WY; Rovani, Jr., Joseph F.

    2012-09-25

    In at least one embodiment, the inventive technology relates to in-vessel generation of a material from a solution of interest as part of a processing and/or analysis operation. Preferred embodiments of the in-vessel material generation (e.g., in-vessel solid material generation) include precipitation; in certain embodiments, analysis and/or processing of the solution of interest may include dissolution of the material, perhaps as part of a successive dissolution protocol using solvents of increasing ability to dissolve. Applications include, but are by no means limited to estimation of a coking onset and solution (e.g., oil) fractionating.

  7. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  8. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  9. 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)

  10. Study on Material Selection of Reactor Pressure Vessel of SCWR

    Science.gov (United States)

    Ma, Shuli; Luo, Ying; Yin, Qinwei; Li, Changxiang; Xie, Guofu

    This paper first analyzes the feasibility of SA-508 Grade 3 Class 1 Steel as an alternative material for Supercritical Water-Cooled Reactor (SCWR) Reactor Pressure Vessel (RPV). This kind of steel is limited to be applied in SCWR RPV due to its quenching property, though large forging could be accomplished by domestic manufacturers in forging aspect. Therefore, steels with higher strength and better quenching property are needed for SWCR RPV. The chemical component of SA-508 Gr.3 Cl.2 steel is similar to that of SA-508 Gr.3 Cl.1 steel, and more appropriate matching of strength and toughness could be achieved by the adjusting the elements contents, as well as proper control of tempering temperature and time. In light of the fact that Cl.2 steel has been successfully applied to steam generator, it could be an alternative material for SWCR RPV. SA-508 Gr.4N steel with high strength and good toughness is another alternative material for SCWR RPV. But large amount of research work before application is still needed for the lack of data on welding and irradiation etc.

  11. Evaluation of WWER-1000 vessel materials fracture toughness

    International Nuclear Information System (INIS)

    Grinik, Eh.U.; Revka, V.N.; Chirko, L.I.; Chajkovskij, Yu.V.

    2007-01-01

    The lifetime of WWER-1000-type reactor vessels is finally conditioned by the fracture toughness (crack growth resistance) of RPV materials. Up to now in line with the regulations the fracture toughness is characterized by the critical temperature of brittleness determined by the results of the Charpy specimen impact testing. Such approach is typical for all countries operating the water pressure reactors. However, regulatory approach is known from the western specialists not always to characterize adequately the crack growth resistance of the vessel materials and in some cases to underestimate their characteristics in the reference state that leads to unreasonably high conservatism. Excessive conservatism may lead to the invalid restrictions in the operating modes and the service life of the reactor vessel. Therefore there appeared the necessity to apply another approaches based on the state-of-the-art experimental methods of the fracture mechanics and allowing evaluating the fracture toughness parameters sufficiently. The paper presents the results of the comparison of the regulatory approach and the Master curve approach from the point of view of the adequate determination of the vessel material crack growth resistance parameters. Analysis of the experimental data of the surveillance specimens illustrated the potential possibility of applying the new statistical method for the WWER-1000- type reactor vessel lifetime extension

  12. Advancing materials research

    International Nuclear Information System (INIS)

    Langford, H.D.; Psaras, P.A.

    1987-01-01

    The topics discussed in this volume include historical perspectives in the fields of materials research and development, the status of selected scientific and technical areas, and current topics in materials research. Papers are presentd on progress and prospects in metallurgical research, microstructure and mechanical properties of metals, condensed-matter physics and materials research, quasi-periodic crystals, and new and artifically structured electronic and magnetic materials. Consideration is also given to materials research in catalysis, advanced ceramics, organic polymers, new ways of looking at surfaces, and materials synthesis and processing

  13. Isotope research materials

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Preparation of research isotope materials is described. Topics covered include: separation of tritium from aqueous effluents by bipolar electrolysis; stable isotope targets and research materials; radioisotope targets and research materials; preparation of an 241 Am metallurgical specimen; reactor dosimeters; ceramic and cermet development; fission-fragment-generating targets of 235 UO 2 ; and wire dosimeters for Westinghouse--Bettis

  14. Cast iron as structural material for hot-working reactor vessels (PCIV)

    International Nuclear Information System (INIS)

    Ostendorf, H.; Schmidt, G.; Pittack, W.

    1977-01-01

    Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary: mechanical properties under compressive stresses; material properties at elevated temperatures; influence of irradiation on mechanical and physical properties; production standards and quality control. The state of the research and the available data of the material testing program are reported

  15. Cast iron as structural material for hot-working reactor vessels (PCIV)

    International Nuclear Information System (INIS)

    Ostendorf, H.; Schmidt, G.; Pittack, W.

    1977-01-01

    Cast iron with lamellar graphite is best suited for prestressed structures, because its compressive strength is nearly 4 times its tensile strength. In comparison to room temperature, cast iron with lamellar graphite shows essentially no loss of strength up to temperatures of 400 0 C. Under the particular aspect to use cast iron for hot-working prestressed reactor pressure vessels (PCIV) (Prestressed cast iron vessel=PCIV) a materials testing program is carried out, which meets the strict certification requirements for materials in the construction of reactor pressure vessels and which completes the presently available knowledge of cast iron. Especially in the following fields an extension and supplement of the present level of knowledge is necessary. - Mechanical properties under compressive stresses. - Material properties at elevated temperatures. - Influence of irradiation on mechanical and physical properties. - Production standards and quality control. The state of the research and the available data of the material testing program are reported. (Auth.)

  16. 49 CFR 176.194 - Stowage of Class 1 (explosive) materials on magazine vessels.

    Science.gov (United States)

    2010-10-01

    ... magazine vessels. 176.194 Section 176.194 Transportation Other Regulations Relating to Transportation... REGULATIONS CARRIAGE BY VESSEL Detailed Requirements for Class 1 (Explosive) Materials Magazine Vessels § 176.194 Stowage of Class 1 (explosive) materials on magazine vessels. (a) General. The requirements of...

  17. Accelerated irradiation test of gundremmingen reactor vessel trepan material

    International Nuclear Information System (INIS)

    Hawthorne, J.R.

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

  18. Accelerated irradiation test of Gundremmingen reactor vessel trepan material

    Energy Technology Data Exchange (ETDEWEB)

    Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

    1992-08-01

    Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

  19. Progress in understanding the mechanical behavior of pressure-vessel materials at elevated temperatures

    International Nuclear Information System (INIS)

    Swindeman, R.W.; Brinkman, C.R.

    1981-01-01

    Progress during the 1970's on the production of high-temperature mechanical properties data for pressure vessel materials was reviewed. The direction of the research was toward satisfying new data requirements to implement advances in high-temperature inelastic design methods. To meet these needs, servo-controlled testing machines and high-resolution extensometry were developed to gain more information on the essential behavioral features of high-temperature alloys. The similarities and differences in the mechanical response of various pressure vessel materials were identified. High-temperature pressure vessel materials that have received the most attention included Type 304 stainless steel, Type 316 stainless steel, 2 1/4 Cr-1 Mo steel, alloy 800H, and Hastelloy X

  20. Development of an integrated data acquision system for research vessels

    Digital Repository Service at National Institute of Oceanography (India)

    Mehra, P.; Desai, R.G.P.

    This article describes an integrated data acquisition system (IDAS) designed and developed for multi-oceanographic research vessels. The prime motivation was to provide a flexible system, which could be used in the context of ocean related...

  1. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  2. Application of material databases for improved reliability of reactor pressure vessels

    International Nuclear Information System (INIS)

    Griesbach, T.J.; Server, W.L.; Beaudoin, B.F.; Burgos, B.N.

    1994-01-01

    A vital part of reactor vessel Life Cycle Management program must begin with an accurate characterization of the vessel material properties. Uncertainties in vessel material properties or use of bounding values may result in unnecessary conservatisms in vessel integrity calculations. These conservatisms may be eliminated through a better understanding of the material properties in reactor vessels, both in the unirradiated and irradiated conditions. Reactor vessel material databases are available for quantifying the chemistry and Charpy shift behavior of individual heats of reactor vessel materials. Application of the databases for vessels with embrittlement concerns has proven to be an effective embrittlement management tool. This paper presents details of database development and applications which demonstrate the value of using material databases for improving material chemistry and for maximizing the data from integrated material surveillance programs

  3. Clearance potential of ITER vacuum vessel activated materials

    International Nuclear Information System (INIS)

    Cepraga, D.G.; Cambi, G.; Frisoni, M.

    2002-01-01

    To demonstrate fusion's environmental attractiveness over the entire life cycle, a waste analysis is mandatory. The clearance is recommended by IAEA for releasing activated solid materials from regulatory control and for waste management policy. The paper focuses on the approach used to support waste analyses for ITER Generic Site Safety Report. The Material Unconditional Clearance Index of all the materials/zones on the equatorial mid-plane of ITER machine have been evaluated, based on IAEA-TECDOC-855. The Bonami-Nitawl-XSDNRPM sequence of the Scale-4.4a code system (using Vitenea-J library) has been firstly used for radiation transport analyses. Then the Anita-2000 code package is used for the activation calculation. The paper presents also, as an example, an application of the clearance indexes estimation for the ITER vacuum vessel materials. The results of the Anita-2000 have been compared with those obtained using the Fispact-99 activation code. (author)

  4. 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

  5. Materials research at CMAM

    International Nuclear Information System (INIS)

    Zucchiatti, Alessandro

    2013-01-01

    The Centro de Micro Analisis de Materiales (CMAM) is a research centre of the Universidad Autónoma de Madrid dedicated to the modification and analysis of materials using ion beam techniques. The infrastructure, based on a HVEE 5MV tandem accelerator, provided with a coaxial Cockcroft Walton charging system, is fully open to research groups of the UAM, to other public research institutions and to private enterprises. The CMAM research covers a few important lines such as advanced materials, surface science, biomedical materials, cultural heritage, materials for energy production. The Centre gives as well support to university teaching and technical training. A detail description of the research infrastructures and their use statistics will be given. Some of the main research results will be presented to show the progress of research in the Centre in the past few years and to motivate the strategic plans for the forthcoming

  6. Materials research at CMAM

    Science.gov (United States)

    Zucchiatti, Alessandro

    2013-07-01

    The Centro de Micro Analisis de Materiales (CMAM) is a research centre of the Universidad Autónoma de Madrid dedicated to the modification and analysis of materials using ion beam techniques. The infrastructure, based on a HVEE 5MV tandem accelerator, provided with a coaxial Cockcroft Walton charging system, is fully open to research groups of the UAM, to other public research institutions and to private enterprises. The CMAM research covers a few important lines such as advanced materials, surface science, biomedical materials, cultural heritage, materials for energy production. The Centre gives as well support to university teaching and technical training. A detail description of the research infrastructures and their use statistics will be given. Some of the main research results will be presented to show the progress of research in the Centre in the past few years and to motivate the strategic plans for the forthcoming.

  7. The technology development for surveillance test of reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Sun Phil; Park, Day Young; Choi, Kwen Jai

    1997-12-01

    Benchmark test was performed in accordance with the requirement of US NRC Reg. Guide DG-1053 for Kori unit-1 in order to determine best-estimated fast neutron fluence irradiated into reactor vessel. Since the uncertainty of radiation analysis comes from the calculation error due to neutron cross-section data, reactor core geometrical dimension, core source, mesh density, angular expansion and convergence criteria, evaluation of calculational uncertainty due to analytical method was performed in accordance with the regulatory guide and the proof was performed for entire analysis by comparing the measurement value obtained by neutron dosimetry located in surveillance capsule. Best-estimated neutron fluence in reactor vessel was calculated by bias factor, neutron flux measurement value/calculational value, from reanalysis result from previous 1st through 4th surveillance testing and finally fluence prediction was performed for the end of reactor life and the entire period of plant life extension. Pressurized thermal shock analysis was performed in accordance with 10 CFR 50.61 using the result of neutron fluence analysis in order to predict the life of reactor vessel material and the criteria of safe operation for Kori unit 1 was reestablished. (author). 55 refs., 55 figs.

  8. New sacrificial material for ex-vessel core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Komlev, Andrei A., E-mail: komlev@kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Nuclear Power Safety Division, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Almjashev, Vyacheslav I., E-mail: vac@mail.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Bechta, Sevostian V., E-mail: bechta@safety.sci.kth.se [Kungliga Tekniska Högskolan (KTH), AlbaNova University Centre, Roslagstullsbacken 21, SE-106 91, Stockholm (Sweden); Khabensky, Vladimir B., E-mail: vladimirkhabensky@gmail.com [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Granovsky, Vladimir S., E-mail: gran@niti.ru [A.P. Aleksandrov Research Institute of Technology, NITI, DSAR, Sosnovy Bor, 188540 (Russian Federation); Gusarov, Victor V., E-mail: victor.v.gusarov@gmail.com [Ioffe Institute, 26 Polytekhnicheskaya Str., St. Petersburg, 194021 (Russian Federation)

    2015-12-15

    A new functional (sacrificial) material has been developed in the Fe{sub 2}O{sub 3}–SrO–Al{sub 2}O{sub 3}–CaO system based on strontium hexaferrite ceramic in concrete matrix. The method of producing SM has been advanced technologically; this technological effectiveness allows the SM to be used in ex-vessel core catchers with corium spreading as well as in crucible-type core catchers. Critical properties regarding the efficiency of SM in ex-vessel core catchers, such as porosity, pycnometric density, apparent density, solidus and liquidus temperatures, and water content have been measured. Suitable fractions of SrFe{sub 12}O{sub 19} and high alumina cement (HAC) were found in the SM based on thermodynamic analysis of the SM/corium interaction. The use of sacrificial steel as an additional heat adsorption component in the core catcher allowed us to increase the mass fraction range of SrFe{sub 12}O{sub 19} in the SM from 0.3−0.5 to 0.3–0.85. The activation temperature of the SM/corium interaction has been shown to correspond to the liquidus temperature of the local composition at the SM/corium interface. The calculated value of this temperature was 1716 °C. Analysis of phase transformations in the SrO–Fe{sub 2}O{sub 3} system revealed advantages of the SrFe{sub 12}O{sub 19}–based sacrificial material compared with the Fe{sub 2}O{sub 3}-contained material owing to the time proximity of SrFe{sub 12}O{sub 19} decomposition and corium interaction activation. - Highlights: • A sacrificial material (SM) was developed for ex-vessel core catcher. • Suitable proportions in the SrFe{sub 12}O{sub 19}–Al{sub 2}O{sub 3}·CaO–Fe system were determined. • Hydrogen release limitation was shown for ex-vessel corium retention with the SM. • Calculated temperature of the active initiation of corium/SM interaction is 1716 °C. • Functional properties of the SM were measured.

  9. Possible research program on a large scale nuclear pressure vessel

    International Nuclear Information System (INIS)

    1983-01-01

    The nuclear pressure vessel structural integrity is actually one of the main items in the nuclear plants safety field. An international study group aimed at investigating the feasibility of a ''possible research program'' on a scale 1:1 LWR pressure vessel. This report presents the study group's work. The different research programs carried out or being carried out in various countries of the European Community are presented (phase I of the study). The main characteristics of the vessel considered for the program and an evaluation of activities required for making them available are listed. Research topic priorities from the different interested countries are summarized in tables (phase 2); a critical review by the study group of the topic is presented. Then, proposals for possible experimental programs and combination of these programs are presented, only as examples of possible useful research activities. The documents pertaining to the results of phase I inquiry performed by the study group are reported in the appendix

  10. The Materiality of Research

    DEFF Research Database (Denmark)

    Meier, Ninna

    2016-01-01

    In this feature essay, Ninna Meier reflects on the materiality of the writing – and re-writing – process in academic research. She explores the ways in which our ever-accummulating thoughts come to form layers on the material objects in which we write our notes and discusses the pleasures of co-authorship....

  11. Reactor vessel dismantling at the high flux materials testing reactor Petten

    International Nuclear Information System (INIS)

    Tas, A.; Teunissen, G.

    1986-01-01

    The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

  12. 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.

  13. 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.).

  14. Arctic research vessel design would expand science prospects

    Science.gov (United States)

    Elsner, Robert; Kristensen, Dirk

    The U.S. polar marine science community has long declared the need for an arctic research vessel dedicated to advancing the study of northern ice-dominated seas. Planning for such a vessel began 2 decades ago, but competition for funding has prevented construction. A new design program is underway, and it shows promise of opening up exciting possibilities for new research initiatives in arctic marine science.With its latest design, the Arctic Research Vessel (ARV) has grown to a size and capability that will make it the first U.S. academic research vessel able to provide access to the Arctic Ocean. This ship would open a vast arena for new studies in the least known of the world's seas. These studies promise to rank high in national priority because of the importance of the Arctic Ocean as a source of data relating to global climate change. Other issues that demand attention in the Arctic include its contributions to the world's heat budget, the climate history buried in its sediments, pollution monitoring, and the influence of arctic conditions on marine renewable resources.

  15. Different approaches to estimation of reactor pressure vessel material embrittlement

    Directory of Open Access Journals (Sweden)

    V. M. Revka

    2013-03-01

    Full Text Available The surveillance test data for the nuclear power plant which is under operation in Ukraine have been used to estimate WWER-1000 reactor pressure vessel (RPV material embrittlement. The beltline materials (base and weld metal were characterized using Charpy impact and fracture toughness test methods. The fracture toughness test data were analyzed according to the standard ASTM 1921-05. The pre-cracked Charpy specimens were tested to estimate a shift of reference temperature T0 due to neutron irradiation. The maximum shift of reference temperature T0 is 84 °C. A radiation embrittlement rate AF for the RPV material was estimated using fracture toughness test data. In addition the AF factor based on the Charpy curve shift (ΔTF has been evaluated. A comparison of the AF values estimated according to different approaches has shown there is a good agreement between the radiation shift of Charpy impact and fracture toughness curves for weld metal with high nickel content (1,88 % wt. Therefore Charpy impact test data can be successfully applied to estimate the fracture toughness curve shift and therefore embrittlement rate. Furthermore it was revealed that radiation embrittlement rate for weld metal is higher than predicted by a design relationship. The enhanced embrittlement is most probably related to simultaneously high nickel and high manganese content in weld metal.

  16. 49 CFR 176.166 - Transport of Class 1 (explosive) materials on passenger vessels.

    Science.gov (United States)

    2010-10-01

    ... 49 Transportation 2 2010-10-01 2010-10-01 false Transport of Class 1 (explosive) materials on....166 Transport of Class 1 (explosive) materials on passenger vessels. (a) Only the following Class 1 (explosive) materials may be transported as cargo on passenger vessels: (1) Division 1.4 (explosive...

  17. R and D Developments. Research Programs on Irradiation Embrittlement of Reactor Vessel Steels

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Serrano, M.; Perosanz, F.

    2000-01-01

    Irradiation embrittlement of pressure vessel steels is a degradation mechanism time dependent that can lead to operational restrictions with adverse effects in the efficiency and life of a plant. For the last year, several research programs have been devoted to study thye evaluation of neutronic radiation effect on mechanical properties of pressure vessel steels. However, at the present, there is a growing interest on the development of new methodologies to optimize the surveillance program information, and the understanding of the irradiation damage mechanism. This paper give an overview of international research programs, and on the R+D activities carried out by the Structural Materials Project on irradiation embrittlement on pressure vessel steels. (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. 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)

  20. Research and development of the prestressed concrete reactor vessel

    International Nuclear Information System (INIS)

    Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

    1975-01-01

    Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

  1. Materials Engineering Research Facility (MERF)

    Data.gov (United States)

    Federal Laboratory Consortium — Argonne?s Materials Engineering Research Facility (MERF) enables engineers to develop manufacturing processes for producing advanced battery materials in sufficient...

  2. Characteristics analysis on a superconductor resonance coil WPT system according to cooling vessel materials in different distances

    International Nuclear Information System (INIS)

    Jeong, In-Sung; Lee, Yu-Kyeong; Choi, Hyo-Sang

    2016-01-01

    Highlights: • WPT using the superconductor coil was needed research for cooling vessel. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance efficiency. • When the distance between the transmitter and receiver coils was 2000 mm, FRP being used for the cooling vessel made the transmission efficiency higher than any other materials. The efficiency and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP. - Abstract: The interest in wireless power transfer (WPT) that can send power without using wires has been increasing recently. Especially, there is a great interest in the wireless power devices for portable IT devices. The WPT devices that have been developed so far use the magnetic induction method, and they are not active due to their distance problem. A magnetic resonance WPT method was developed and has been actively researched to resolve this problem. A superconductor coil was applied in this study to increase the efficiency of the magnetic resonance WPT. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance. The distance between the transmitter and receiver coils started from 800 mm and was increased by 200 mm. The reflection coefficient was measured at each distance. As a result, FRP, bakelite, plastic PVC, polystyrene of the reflection coefficient was similar. From among these FRP being used for the cooling vessel made the transmission characteristics higher than any other materials when the distance between the transmitter and receiver coils was 2,000 mm. On the other hand, the reflection coefficient dropped when iron was used. It is estimated based on the experimental results that the wireless power transmission characteristics and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP.

  3. Characteristics analysis on a superconductor resonance coil WPT system according to cooling vessel materials in different distances

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, In-Sung, E-mail: no21park@hanmail.net; Lee, Yu-Kyeong; Choi, Hyo-Sang, E-mail: hyosang@chosun.ac.kr

    2016-11-15

    Highlights: • WPT using the superconductor coil was needed research for cooling vessel. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance efficiency. • When the distance between the transmitter and receiver coils was 2000 mm, FRP being used for the cooling vessel made the transmission efficiency higher than any other materials. The efficiency and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP. - Abstract: The interest in wireless power transfer (WPT) that can send power without using wires has been increasing recently. Especially, there is a great interest in the wireless power devices for portable IT devices. The WPT devices that have been developed so far use the magnetic induction method, and they are not active due to their distance problem. A magnetic resonance WPT method was developed and has been actively researched to resolve this problem. A superconductor coil was applied in this study to increase the efficiency of the magnetic resonance WPT. FRP, bakelite, polystyrene, aluminum, and iron were applied as the cooling vessel material to analyze the WPT distance. The distance between the transmitter and receiver coils started from 800 mm and was increased by 200 mm. The reflection coefficient was measured at each distance. As a result, FRP, bakelite, plastic PVC, polystyrene of the reflection coefficient was similar. From among these FRP being used for the cooling vessel made the transmission characteristics higher than any other materials when the distance between the transmitter and receiver coils was 2,000 mm. On the other hand, the reflection coefficient dropped when iron was used. It is estimated based on the experimental results that the wireless power transmission characteristics and distance of sending power can be improved in the superconductor coil if the cooling vessel is made with FRP.

  4. Manufacturing and material properties of forgings for reactor pressure vessel of high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Sato, I.; Suzuki, K.

    1994-01-01

    For the reactor pressure vessel (RPV) of high temperature engineering test reactor (HTTR) which has been developed by Japan Atomic Energy Research Institute (JAERI), 2 1/4Cr-1Mo steel is used first in the world. Material confirmation test has been carried out to demonstrate good applicability of forged low Si 2 1/4Cr-1Mo steel to the RPV of HTTR. Recently, JSW has succeeded in the manufacturing of large size ring forgings and large size forged cover dome integrated with nozzles for stand pipe for the RPV. This paper describes the results of the material confirmation test as well as the manufacturing and material properties of the large forged cover dome integrated with nozzles for stand pipe. (orig.)

  5. Fusion program research materials inventory

    International Nuclear Information System (INIS)

    Roche, T.K.; Wiffen, F.W.; Davis, J.W.; Lechtenberg, T.A.

    1984-01-01

    Oak Ridge National Laboratory maintains a central inventory of research materials to provide a common supply of materials for the Fusion Reactor Materials Program. This will minimize unintended material variations and provide for economy in procurement and for centralized record keeping. Initially this inventory is to focus on materials related to first-wall and structural applications and related research, but various special purpose materials may be added in the future. The use of materials from this inventory for research that is coordinated with or otherwise related technically to the Fusion Reactor Materials Program of DOE is encouraged

  6. The 5th surveillance testing for Kori unit 1 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed primarily by Korea Atomic Energy Research Institute and Westinhouse corporation partially involved in testing and calculation data evaluation in order to obtain reliable test result. Fast neutron fluences for capsule V, T, S, R and P were 5.087E+18, 1.115E+19, 1.228E+19, 2.988E+19, and 3.938E+19n/cm2, respectively. The bias factor, the ratio of calculation/measurement, was 0.940 for the 1st through 5th testing and the calculational uncertainty, 7% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.9846E+19n/cm{sup 2} based on the end of 17th fuel cycle and it was predicted that the fluences of vessel inside surface at 24, 32, 40 and 48EFPY would reach 3.0593E+19, 4.0695E+19, 5.0797E+19 and 6.0900E+19n/cm{sup 2} based on the current calculation. PTS analysis for Kori unit 1 showed that 27.93EFPY was the threshold value for 300 deg F requirement. 71 refs., 33 figs., 52 tabs. (Author)

  7. Fabrication techniques of metal liner used for pressure vessels made by composite material

    International Nuclear Information System (INIS)

    Takahashi, W.K.; Al-Qureshi, H.A.

    1982-01-01

    Different viable techniques for the manufacturing of metal liner used for pressure vessels are presented. The aim of these metal liner is to avoid the fluid leakage from the pressurized vessel and to serve as a mandreal to be wound by composite material. The studied techniques are described and the practical results are illustrated. Finally a comparative study of the manufacturing techniques is made in order to define the process that furnishes the metal liner with the best characteristics. The advantages offered by these type of pressure vessels when compared with the conventional metallic vessels, are also presented. (Author) [pt

  8. Materials Behavior Research Laboratory

    Data.gov (United States)

    Federal Laboratory Consortium — The purpose is to evaluate mechanical properties of materials including metals, intermetallics, metal-matrix composites, and ceramic-matrix composites under typical...

  9. The 4th surveillance testing for Kori unit 3 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-10-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

  10. The 5th surveillance testing for Kori unit 2 reactor vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules V, R, P, T and N are 2.837E+18, 1.105E+19, 2.110E+19, 3.705E+19 and 4.831E+19n/cm{sup 2}, respectively. The bias factor, the ratio of measurement/calculation, was 0.918 for the 1st through 5th testing and the calculational uncertainty, 11.6% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.898E+19n/cm{sup 2} based on the end of 15th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 4.203E+19, 5.232E+19, 6.262E+19 and 7.291E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 49 refs., 35 figs., 48 tabs. (Author)

  11. A review of formulas for predicting irradiation embrittlement of reactors vessel materials

    International Nuclear Information System (INIS)

    Petrequin, P.

    1995-01-01

    Formulas developed in different countries for predicting irradiation embrittlement of reactors vessel materials are presented. Results of predictions were compared with different data sets, from surveillance programmes or studies in test reactors, with different residual elements contents. Figs

  12. Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

    International Nuclear Information System (INIS)

    Raske, D.T.; Wang, Z.

    1992-01-01

    The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

  13. Role of Outgassing of ITER Vacuum Vessel In-Wall Shielding Materials in Leak Detection of ITER Vacuum Vessel

    Science.gov (United States)

    Maheshwari, A.; Pathak, H. A.; Mehta, B. K.; Phull, G. S.; Laad, R.; Shaikh, M. S.; George, S.; Joshi, K.; Khan, Z.

    2017-04-01

    ITER Vacuum Vessel is a torus-shaped, double wall structure. The space between the double walls of the VV is filled with In-Wall Shielding Blocks (IWS) and Water. The main purpose of IWS is to provide neutron shielding during ITER plasma operation and to reduce ripple of Toroidal Magnetic Field (TF). Although In-Wall Shield Blocks (IWS) will be submerged in water in between the walls of the ITER Vacuum Vessel (VV), Outgassing Rate (OGR) of IWS materials plays a significant role in leak detection of Vacuum Vessel of ITER. Thermal Outgassing Rate of a material critically depends on the Surface Roughness of material. During leak detection process using RGA equipped Leak detector and tracer gas Helium, there will be a spill over of mass 3 and mass 2 to mass 4 which creates a background reading. Helium background will have contribution of Hydrogen too. So it is necessary to ensure the low OGR of Hydrogen. To achieve an effective leak test it is required to obtain a background below 1 × 10-8 mbar 1 s-1 and hence the maximum Outgassing rate of IWS Materials should comply with the maximum Outgassing rate required for hydrogen i.e. 1 x 10-10 mbar 1 s-1 cm-2 at room temperature. As IWS Materials are special materials developed for ITER project, it is necessary to ensure the compliance of Outgassing rate with the requirement. There is a possibility of diffusing the gasses in material at the time of production. So, to validate the production process of materials as well as manufacturing of final product from this material, three coupons of each IWS material have been manufactured with the same technique which is being used in manufacturing of IWS blocks. Manufacturing records of these coupons have been approved by ITER-IO (International Organization). Outgassing rates of these coupons have been measured at room temperature and found in acceptable limit to obtain the required Helium Background. On the basis of these measurements, test reports have been generated and got

  14. Material problems in accident analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

    Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

  15. Environmental TEM for Materials Research

    DEFF Research Database (Denmark)

    Hansen, Thomas Willum

    Over the last decades, electron microscopy has played a large role in materials research. The increasing use of particularly environmental transmission electron microscopy (ETEM) in materials science provides new possibilities for investigating nanoscale components at work. Careful experimentation...

  16. Environmental TEM in Materials Research

    DEFF Research Database (Denmark)

    Hansen, Thomas Willum; Wagner, Jakob Birkedal

    Over the last decades, electron microscopy has played a large role in materials research. The increasing use of particularly environmental transmission electron microscopy (ETEM) in materials science provides new possibilities for investigating nanoscale components at work. Careful experimentation...

  17. Energy Materials Research Laboratory (EMRL)

    Data.gov (United States)

    Federal Laboratory Consortium — The Energy Materials Research Laboratory at the Savannah River National Laboratory (SRNL) creates a cross-disciplinary laboratory facility that lends itself to the...

  18. The Materiality of Research

    DEFF Research Database (Denmark)

    Meier, Ninna

    2016-01-01

    In this feature essay, Ninna Meier explores the relationship between time, space and academic writing. She ponders the ‘portable magic’ of research: namely, the capacity for our thoughts to be both grounded in a particular point in time and space and yet simultaneously ‘free from these dimensions...

  19. Material properties of Bohunice 1 and 2 reactor pressure vessel materials before and after annealing

    International Nuclear Information System (INIS)

    Brumovsky, M.; Novosad, P.; Vacek, M.

    1994-01-01

    Six types of experimental RPV materials were studied before and after irradiation in host nuclear power and research reactors. Recovery of RPV materials from radiation hardening and embrittlement after annealing was evaluated including a rate of radiation damage after reirradiation used. (author). 3 refs, 4 figs, 2 tabs

  20. Strategic research on materials

    International Nuclear Information System (INIS)

    Williams, J.

    1987-01-01

    Strategic research is defined as that which is necessary to support not only an understanding of the phenomenon on which a new technology is based, but also the raft of other technologies needed to exploit the new phenomenon. The theme is illustrated by reference to the development of ceramics of importance to the nuclear industry and in particularly with relation to the AGR. Starting from natural uranium, the underlying and wide ranging research effort devoted to the technology of isotopic enrichment, the investigation of the uranium-oxygen binary system, fabrication of uranium dioxide fuel, interactions between the fuel and stainless steel cans, between the cans and CO 2 coolant and between the coolant and graphite moderator, is outlined. The role of ceramics in stable radioactive waste containment is also briefly mentioned. (author)

  1. Materials Sciences Research.

    Science.gov (United States)

    1975-07-01

    the vicinity of the LaCoO composition. Several derivative compounds with structures related to the Perovskite structure have been identified. The...physical, chemical, and electrical properties results. Glass-Ceramics are used as substrates and as insulation in hybrid electronic circuits, as... Photoluminescence Characterization of Laser-Quality (100) In1 Ga P • Journal of Crystal Growth 27, 154-165 (1974) , Supported by the Advanced Research Projects

  2. German boiler and pressure vessel codes and standards: materials, manufacture, testing, equipment, erection and operation

    International Nuclear Information System (INIS)

    Steffen, H.P.

    1987-01-01

    The methods by which the safety objectives on the operation of steam boilers and pressure vessels in Germany can be reached are set out in Technical Rules which are compiled and established in technical committees. Typical applications are described in the Technical Rules. A chart shows how the laws, provisions and Technical Rules for the sections 'steam boiler plant' and 'pressure vessels' are interlinked. This chapter concentrates on legal aspects, materials, manufacture, testing, erection and operation of boilers and pressure vessels in Germany. (U.K.)

  3. Weld repair of helium degraded reactor vessel material

    International Nuclear Information System (INIS)

    Kanne, W.R. Jr.; Lohmeier, D.A.; Louthan, M.R. Jr.; Rankin, D.T.; Franco-Ferreira, E.A.; Bruck, G.J.; Madeyski, A.; Shogan, R.P.; Lessmann, G.G.

    1990-01-01

    Welding methods for modification or repair of irradiated nuclear reactor vessels are being evaluated at the Savannah River Site. A low-penetration weld overlay technique has been developed to minimize the adverse effects of irradiation induced helium on the weldability of metals and alloys. This technique was successfully applied to Type 304 stainless steel test plates that contained 3 to 220 appm helium from tritium decay. Conventional welding practices caused significant cracking and degradation in the test plates. Optical microscopy of weld surfaces and cross sections showed that large surface toe cracks formed around conventional welds in the test plates but did not form around overlay welds. Scattered incipient underbead cracks (grain boundary separations) were associated with both conventional and overlay test welds. Tensile and bend tests were used to assess the effect of base metal helium content on the mechanical integrity of the low-penetration overlay welds. The axis of tensile specimens was perpendicular to the weld-base metal interface. Tensile specimens were machined after studs were resistance welded to overlay surfaces

  4. Basic requirements of mechanical properties for nuclear pressure vessel materials in ASME-BPV code

    International Nuclear Information System (INIS)

    Ning Dong; Yao Weida

    2011-01-01

    The four basic aspects of strengths, ductility, toughness and fatigue strengths can be summarized for overall mechanical properties requirements of materials for nuclear pressure-retaining vessels in ASME-BPV code. These mechanical property indexes involve in the factors of melting, manufacture, delivery conditions, check or recheck for mechanical properties and chemical compositions, etc. and relate to degradation and damage accumulation during the use of materials. This paper specifically accounts for the basic requirements and theoretic basis of mechanical properties for nuclear pressure vessel materials in ASME-BPV code and states the internal mutual relationships among the four aspects of mechanical properties. This paper focuses on putting forward at several problems on mechanical properties of materials that shall be concerned about during design and manufacture for nuclear pressure vessels according to ASME-BPV code. (author)

  5. 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

  6. The National Shipbuilding Research Program: Producibility Cost Reductions through Alternative Materials and Processes

    National Research Council Canada - National Science Library

    Horsmon, Jr., Albert W; Johnson, Karl; Gans-Devney, Barbara

    1999-01-01

    This report describes research into the use of alternative materials and processes to reduce material and labor costs while also looking at the influence of these choices on the life cycle costs of the vessel...

  7. Closed vessel miniaturized microwave assisted chelating extraction for determination of trace metals in plant materials

    Science.gov (United States)

    Czarnecki, Sezin; Duering, Rolf-Alexander

    2013-04-01

    In recent years, the use of closed vessel microwave assisted extraction (MAE) for plant samples has shown increasing research interest which will probably substitute conventional procedures in the future due to their general disadvantages including consumption of time and solvents. The objective of this study was to demonstrate an innovative miniaturized closed vessel microwave assisted extraction (µMAE) method under the use of EDTA (µMAE-EDTA) to determine metal contents (Cd, Co, Cu, Mn, Ni, Pb, Zn) in plant samples (Lolio-Cynosuretum) by inductively coupled plasma-optical emission spectrometry (ICP-OES). Validation of the method was done by comparison of the results with another miniaturized closed vessel microwave HNO3 method (µMAE-H) and with two other macro scale MAE procedures (MAE-H and MAE-EDTA) which were applied by using a mixture of nitric acid (HNO3) and hydrogen peroxide (H2O2) (MAE-H) and EDTA (MAE-EDTA), respectively. The already established MAE-H method is taken into consideration as a reference validation MAE method for plant material. A conventional plant extraction (CE) method, based on dry ashing and dissolving of the plant material in HNO3, was used as a confidence comparative method. Certified plant reference materials (CRMs) were used for comparison of recovery rates from different extraction protocols. This allowed the validation of the applicability of the µMAE-EDTA procedure. For 36 real plant samples with triplicates each, µMAE-EDTA showed the same extraction yields as the MAE-H in the determination of Cd, Co, Cu, Mn, Ni, Pb, and Zn contents in plant samples. Analytical parameters in µMAE-EDTA should be further investigated and adapted for other metals of interest. By the reduction and elimination of the use of hazardous chemicals in environmental analysis and thus allowing a better understanding of metal distribution and accumulation process in plants and also the metal transfer from soil to plants and into the food chain, µ

  8. Thermophysical methods in materials research

    International Nuclear Information System (INIS)

    Rohde, M.

    2003-01-01

    Thermophysical properties, namely the thermal conductivity, diffusivity and the heat capacity determine the behavior of every material under heat load. Therefore these properties are important not only for design purposes but also for the development of advanced materials. Within this contribution an overview will be given about measurement techniques for thermophysical properties. Some aspects of materials characterization and process development will be highlighted using selected research results. (orig.)

  9. General and crevice corrosion study of the in-wall shielding materials for ITER vacuum vessel

    Science.gov (United States)

    Joshi, K. S.; Pathak, H. A.; Dayal, R. K.; Bafna, V. K.; Kimihiro, Ioki; Barabash, V.

    2012-11-01

    Vacuum vessel In-Wall Shield (IWS) will be inserted between the inner and outer shells of the ITER vacuum vessel. The behaviour of IWS in the vacuum vessel especially concerning the susceptibility to crevice of shielding block assemblies could cause rapid and extensive corrosion attacks. Even galvanic corrosion may be due to different metals in same electrolyte. IWS blocks are not accessible until life of the machine after closing of vacuum vessel. Hence, it is necessary to study the susceptibility of IWS materials to general corrosion and crevice corrosion under operations of ITER vacuum vessel. Corrosion properties of IWS materials were studied by using (i) Immersion technique and (ii) Electro-chemical Polarization techniques. All the sample materials were subjected to a series of examinations before and after immersion test, like Loss/Gain weight measurement, SEM analysis, and Optical stereo microscopy, measurement of surface profile and hardness of materials. After immersion test, SS 304B4 and SS 304B7 showed slight weight gain which indicate oxide layer formation on the surface of coupons. The SS 430 material showed negligible weight loss which indicates mild general corrosion effect. On visual observation with SEM and Metallography, all material showed pitting corrosion attack. All sample materials were subjected to series of measurements like Open Circuit potential, Cyclic polarization, Pitting potential, protection potential, Critical anodic current and SEM examination. All materials show pitting loop in OC2 operating condition. However, its absence in OC1 operating condition clearly indicates the activity of chloride ion to penetrate oxide layer on the sample surface, at higher temperature. The critical pitting temperature of all samples remains between 100° and 200°C.

  10. Research reactors and materials testing

    International Nuclear Information System (INIS)

    Vidal, H.

    1986-01-01

    Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

  11. The influence of fire exposure on austenitic stainless steel for pressure vessel fitness-for-service assessment: Experimental research

    Science.gov (United States)

    Li, Bo; Shu, Wenhua; Zuo, Yantian

    2017-04-01

    The austenitic stainless steels are widely applied to pressure vessel manufacturing. The fire accident risk exists in almost all the industrial chemical plants. It is necessary to make safety evaluation on the chemical equipment including pressure vessels after fire. Therefore, the present research was conducted on the influences of fire exposure testing under different thermal conditions on the mechanical performance evolution of S30408 austenitic stainless steel for pressure vessel equipment. The metallurgical analysis described typical appearances in micro-structure observed in the material suffered by fire exposure. Moreover, the quantitative degradation of mechanical properties was investigated. The material thermal degradation mechanism and fitness-for-service assessment process of fire damage were further discussed.

  12. Structural mechanisms of the flux effect for VVER-1000 reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Gurovich, B.; Kuleshova, E.; Fedotova, S.; Maltsev, D.; Zabusov, O.; Frolov, A.; Erak, D.; Zhurko, D.

    2015-01-01

    To justify the lifetime extension of VVER-1000 reactor pressure vessels (RPV) up to 60 years and more it is necessary to expand the existing surveillance samples database to beyond design fluence by means of accelerated irradiation in a research reactor. Herewith since the changes in mechanical properties of materials under irradiation are due to occurring structural changes, correct analysis of the data obtained at accelerated irradiation of VVER-1000 RPV materials requires a clear understanding of the structural mechanisms that are responsible for the flux effect in VVER-1000 RPV steels. Two mechanisms are responsible for radiation embrittlement of VVER-1000 RPV steels: the hardening one (radiation hardening due to formation of radiation-induced Ni-based precipitates and radiation defects) and non-hardening one (due to formation of impurities segregations at grain boundaries - reversible temper brittleness). In this context for an adequate interpretation of the mechanical tests results when justifying the lifetime extension of existing units a complex of comparative structural studies (TEM, SEM and AES) of VVER-1000 RPV materials irradiated in different conditions (in research reactor IR-8 and within surveillance samples) was performed. It is shown that the flux effect is observed for materials with high nickel content (weld metals with Ni content > 1.35%) and it is mostly due to the contribution of non-hardening mechanism of radiation embrittlement (the difference in the accumulation kinetics of grain boundary phosphorus segregation) and somewhat contribution of the hardening mechanism (the difference in density of radiation-induced precipitates). Therefore when analyzing the results obtained from the accelerated irradiation of VVER-1000 WM the correction for the flux effect should be made. (authors)

  13. Crafting glass vessels: current research on the ancient glass collections in the Freer Gallery of Art, Washington, D.C.

    Science.gov (United States)

    Nagel, Alexander; McCarthy, Blythe; Bowe, Stacy

    Our knowledge of glass production in ancient Egypt has been well augmented by the publication of recently excavated materials and glass workshops, but also by more recent materials analysis, and experiments of modern glass-makers attempting to reconstruct the production process of thin-walled coreformed glass vessels. From the mounting of a prefabricated core to the final glass product our understanding of this profession has much improved. The small but well preserved glass collection of the Freer Gallery of Art in Washington, D.C. is a valid tool for examining and studying the technology and production of ancient Egyptian core formed glass vessels. Charles Lang Freer (1854-1919) acquired most of the material from Giovanni Dattari in Cairo in 1909. Previously the glass had received only limited discussion, suggesting that most of these vessels were produced in the 18th Dynasty in the 15th and 14th centuries BCE, while others date from the Hellenistic period and later. In an ongoing project we conducted computed radiography in conjunction with qualitative x-ray fluorescence analysis on a selected group of vessels to understand further aspects of the ancient production process. This paper will provide an overview of our recent research and present our data-gathering process and preliminary results. How can the examinations of core formed glass vessels in the Freer Gallery contribute to our understanding of ancient glass production and technology? By focusing on new ways of looking at old assumptions using the Freer Gallery glass collections, we hope to increase understanding of the challenges of the production process of core-vessel technology as represented by these vessels.

  14. Embrittlement of the nuclear icebreaker Lenin reactor pressure vessel materials reconstruction

    International Nuclear Information System (INIS)

    Krasikov, E.A.; Nikolaenko, V.A.

    2008-01-01

    Paper deals with the results of the efforts to examine the radiation damage of the Lenin nuclear-powered ice-breaker decommissioned reactor pressure vessel on the basis of which one has determined the peculiar features of the metal radiation embrittlement. Under 10 10 -10 11 s -1 cm -2 low density neutron flux irradiation one notes the most intensive embrittlement of the metal. Then, as the noxious element content in the metal matrix grows smaller the embrittlement reduces up to the change of sign as to the normal curve plotted at the neutron flux density exceeding 10 13 s -1 cm -2 . One assumes that as a result of the low density neutron flux irradiation the reactor pressure vessel edge spaces at some operation stages may be damaged more severely in contrast to these near the reactor core. The neutron irradiation density is the factor affecting the reactor vessel material embrittlement, that is why, it is important to study the damage mechanism of the materials of the power reactor vessels under design characterized by the low radiation load. The mentioned is important, as well, to evaluate the efficiency of the efforts undertaken to mitigate the effect of the neutron radiation on the reactor vessel [ru

  15. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Energy Technology Data Exchange (ETDEWEB)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R. [ITER JCT, Garching (Germany)

    1998-10-01

    Design and R and D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R and D results. The resulting design changes are discussed for each system. (orig.) 11 refs.

  16. Design and material selection for ITER first wall/blanket, divertor and vacuum vessel

    Science.gov (United States)

    Ioki, K.; Barabash, V.; Cardella, A.; Elio, F.; Gohar, Y.; Janeschitz, G.; Johnson, G.; Kalinin, G.; Lousteau, D.; Onozuka, M.; Parker, R.; Sannazzaro, G.; Tivey, R.

    1998-10-01

    Design and R&D have progressed on the ITER vacuum vessel, shielding and breeding blankets, and the divertor. The principal materials have been selected and the fabrication methods selected for most of the components based on design and R&D results. The resulting design changes are discussed for each system.

  17. Use of Master Curve technology for assessing shallow flaws in a reactor pressure vessel material

    International Nuclear Information System (INIS)

    Bass, Bennett Richard; Taylor, Nigel

    2006-01-01

    In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.

  18. RESEARCH OF REFRIGERATION SYSTEMS FAILURES IN POLISH FISHING VESSELS

    Directory of Open Access Journals (Sweden)

    Waldemar KOSTRZEWA

    2013-07-01

    Full Text Available Temperature is a basic climatic parameter deciding about the quality change of fishing products. Time, after which qualitative changes of caught fish don’t exceed established, acceptable range, is above all the temperature function. Temperature reduction by refrigeration system of the cargo hold is a basic technical method, which allows extend transport time. Failures of refrigeration systems in fishing vessels have a negative impact on the environment in relation to harmful refrigerants emission. The paper presents the statistical analysis of failures occurred in the refrigeration systems of Polish fishing vessels in 2007‐2011 years. Analysis results described in the paper can be a base to draw up guidelines, both for designers as well as operators of the marine refrigeration systems.

  19. 77 FR 60042 - Safety Zone; Research Vessel SIKULIAQ Launch, Marinette, WI

    Science.gov (United States)

    2012-10-02

    ...: Temporary final rule. SUMMARY: The Coast Guard is establishing a temporary safety zone on the Menominee River in Marinette Wisconsin. This zone is intended to restrict vessels from a portion of Menominee River during the launching of the Research vessel SIKULIAQ, on October 13th, 2012. This temporary safety...

  20. Materials interaction tests to identify base and coating materials for an enhanced in-vessel core catcher design

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J.L.; Knudson, D.L.; Condie, K.G.; Swank, W.D. [Idaho National Engineering and Environmental Laboratory, Idaho Falls ID (United States); Cheung, F.B. [Pennsylvania State University, Department of Mechanical and Nuclear Engineering, University Park PA (United States); Suh, K.Y. [Seoul National University, Department of Nuclear Engineering, Seoul (Korea, Republic of); Kim, S.B. [Korea Atomic Energy Research Institute, Severe Accident Research Project, Taejon (Korea, Republic of)

    2004-07-01

    An enhanced in-vessel core catcher is being designed and evaluated, it must ensure In-Vessel Retention of core materials that may relocate under severe accident conditions in advanced reactors. To reduce cost and simplify manufacture and installation, this new core catcher design consists of several interlocking sections that are machined to fit together when inserted into the lower head. If needed, the core catcher can be manufactured with holes to accommodate lower head penetrations. Each section of the core catcher consists of two material layers with an option to add a third layer (if deemed necessary): a base material, which has the capability to support and contain the mass of core materials that may relocate during a severe accident; an insulating oxide coating material on top of the base material, which resists interactions with high-temperature core materials; and an optional coating on the bottom side of the base material to prevent any potential oxidation of the base material during the lifetime of the reactor. Initial evaluations suggest that a thermally-sprayed oxide material is the most promising candidate insulator coating for a core catcher. Tests suggest that 2 coatings can provide adequate protection to a stainless steel core catcher: -) a 500 {mu}m thick zirconium dioxide coating over a 100-200 {mu}m Inconel 718 bond coating, and -) a 500 {mu}m thick magnesium zirconate coating.

  1. 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.

  2. 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.

  3. Elasto-Plastic Stress Analysis in Rotating Disks and Pressure Vessels Made of Functionally Graded Materials

    Directory of Open Access Journals (Sweden)

    Amir T. Kalali

    Full Text Available Abstract A new elastio-plastic stress solution in axisymmetric problems (rotating disk, cylindrical and spherical vessel is presented. The rotating disk (cylindrical and spherical vessel was made of a ceramic/metal functionally graded material, i.e. a particle-reinforced composite. It was assumed that the material's plastic deformation follows an isotropic strain-hardening rule based on the von-Mises yield criterion. The mechanical properties of the graded material were modeled by the modified rule of mixtures. By assuming small strains, Hencky's stress-strain relation was used to obtain the governing differential equations for the plastic region. A numerical method for solving those differential equations was then proposed that enabled the prediction of stress state within the structure. Selected finite element results were also presented to establish supporting evidence for the validation of the proposed approach.

  4. 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.).

  5. Research and materials irradiation reactors

    International Nuclear Information System (INIS)

    Ballagny, A.; Guigon, B.

    2004-01-01

    Devoted to the fundamental and applied research on materials irradiation, research reactors are nuclear installations where high neutrons flux are maintained. After a general presentation of the research reactors in the world and more specifically in France, this document presents the heavy water cooled reactors and the water cooled reactors. The third part explains the technical characteristics, thermal power, neutron flux, operating and details the Osiris, the RHF (high flux reactor), the Orphee and the Jules Horowitz reactors. The last part deals with the possible utilizations. (A.L.B.)

  6. Thermal and mechanical cyclic loading of thick spherical vessels made of transversely isotropic materials

    International Nuclear Information System (INIS)

    Komijani, M.; Mahbadi, H.; Eslami, M.R.

    2013-01-01

    The aim of this paper is to obtain the dependency of the ratcheting, reversed plasticity, or shakedown behavior of spherical vessels made of some anisotropic materials to the stress category of imposed cyclic loading. The Hill anisotropic yield criterion with the kinematic hardening theories of plasticity based on the Prager and Armstrong–Frederick models are used to predict the yield of the vessel and obtain the plastic strains. An iterative numerical method is used to simulate the cyclic loading behavior of the structure. The effect of mean and amplitude of the mechanical and thermal loads on cyclic behavior and ratcheting rate of the vessel is investigated respectively. The ratcheting rate for the vessels made of transversely isotropic material is evaluated for the various ratios of anisotropy. -- Highlights: ► Cyclic loading analysis of anisotropic spheres is assessed. ► Using the Prager model results in ratcheting. ► Armstrong-Frederick model predicts ratcheting for load controlled cyclic loadings. ► The A-F model predicts ratcheting to a stabilized cycle for thermal loadings

  7. Evaluation and prediction of neutron embrittlement in reactor pressure vessel materials. Final report

    International Nuclear Information System (INIS)

    Hawthorne, J.R.; Menke, B.H.; Loss, F.J.; Watson, H.E.; Hiser, A.L.; Gray, R.A.

    1982-12-01

    This study evaluates the effects of fast neutron irradiation on the mechanical properties of eight nuclear reactor vessel materials. The materials include submerged arc weldments, three plates, and one forging. The materials are in the unirradiated and irradiated conditions with regard to tensile, Charpy impact, and static and dynamic fracture toughness properties. Correlations between impact and fracture toughness parameters are developed from the experimental results. The observed shifts in transition temperature and the drop in upper-shelf energy are compared with predictions developed from the Regulatory Guide 1.99.1 trend curves

  8. Reactor pressure vessel embrittlement management through EPRI-Developed material property databases

    International Nuclear Information System (INIS)

    Rosinski, S.T.; Server, W.L.; Griesbach, T.J.

    1997-01-01

    Uncertainties and variability in U.S. reactor pressure vessel (RPV) material properties have caused the U.S. Nuclear Regulatory Commission (NRC) to request information from all nuclear utilities in order to assess the impact of these data scatter and uncertainties on compliance with existing regulatory criteria. Resolving the vessel material uncertainty issues requires compiling all available data into a single integrated database to develop a better understanding of irradiated material property behavior. EPRI has developed two comprehensive databases for utility implementation to compile and evaluate available material property and surveillance data. RPVDATA is a comprehensive reactor vessel materials database and data management program that combines data from many different sources into one common database. Searches of the data can be easily performed to identify plants with similar materials, sort through measured test results, compare the ''best-estimates'' for reported chemistries with licensing basis values, quantify variability in measured weld qualification and test data, identify relevant surveillance results for characterizing embrittlement trends, and resolve uncertainties in vessel material properties. PREP4 has been developed to assist utilities in evaluating existing unirradiated and irradiated data for plant surveillance materials; PREP4 evaluations can be used to assess the accuracy of new trend curve predictions. In addition, searches of the data can be easily performed to identify available Charpy shift and upper shelf data, review surveillance material chemistry and fabrication information, review general capsule irradiation information, and identify applicable source reference information. In support of utility evaluations to consider thermal annealing as a viable embrittlement management option, EPRI is also developing a database to evaluate material response to thermal annealing. Efforts are underway to develop an irradiation

  9. Plutonium contaminated materials research programme

    International Nuclear Information System (INIS)

    Higson, S.G.

    1986-01-01

    The paper is a progress report for 1985 from the Plutonium Contaminated Materials Working Party (PCMWP). The PCMWP co-ordinates research and development on a national basis in the areas of management, treatment and immobilisation of plutonium contaminated materials, for the purpose of waste management. The progress report contains a review of the development work carried out in eight areas, including: reduction of arisings, plutonium measurement, sorting and packaging, washing of shredded combustible PCM, decommissioning and non-combustible PCM treatment, PCM immobilisation, treatment of alpha bearing liquid wastes, and engineering objectives. (UK)

  10. Investigation and analysis on ITER in-vessel coils’ raw-materials

    International Nuclear Information System (INIS)

    Jin, Huan; Wu, Yu; Long, Feng; Yu, Min; Han, Qiyang; Liu, Huajun

    2013-01-01

    Highlights: • The R and D works for the ITER in-vessel coils (IVC) are now being conducted in Institute of Plasma Physics, and the analysis work are being done by Princeton Plasma Physics Laboratory. • There is little published paper about the raw materials for ITER IVC coils. • This manuscript points out the progress of the selected materials for ITER IVC coils. -- Abstract: The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor

  11. Overview of research trends and problems on Cr-Mo low alloy steels for pressure vessel

    International Nuclear Information System (INIS)

    Chi, Byung Ha; Kim, Jeong Tae

    2000-01-01

    Cr-Mo low alloy steels have been used for a long time for pressure vessel due to its excellent corrosion resistance, high temperature strength and toughness. The paper reviewed the latest trends on material development and some problems on Cr-Mo low alloy steel for pressure vessel, such as elevated temperature strength, hardenability, synergetic effect between temper and hydrogen embrittlement, hydrogen attack and hydrogen induced disbonding of overlay weld-cladding

  12. Nuclear reactor pressure vessel surveillance capsule examinations. Application of American Society for Testing and Materials Standards

    International Nuclear Information System (INIS)

    Perrin, J.S.

    1978-01-01

    A series of pressure vessel surveillance capsules is installed in each commercial nuclear power plant in the United States. A capsule typically contains neutron dose meters, thermal monitors, tensile specimens, and Charpy V-notch impact specimens. In order to determine property changes of the pressure vessel resulting from irradiation, surveillance capsules are periodically removed during the life of a reactor and examined. There are numerous standards, regulations, and codes governing US pressure vessel surveillance capsule programmes. These are put out by the US Nuclear Regulatory Commission, the Boiler and Pressure Vessel Committee of the American Society of Mechanical Engineers, and the American Society for Testing and Materials (ASTM). A majority of the pertinent ASTM standards are under the jurisdiction of ASTM Committee E-10 on Nuclear Applications and Measurements of Radiation Effects. The standards, regulations, and codes pertaining to pressure vessel surveillance play an important role in ensuring reliability of the nuclear pressure vessels. ASTM E 185-73 is the Standard Recommended Practice for Surveillance Tests for Nuclear Reactors. This standard recommends procedures for both the irradiation and subsequent testing of surveillance capsules. ASTM E 185-73 references many additional specialized ASTM standards to be followed in specific areas of a surveillance capsule examination. A key element of surveillance capsule programmes is the Charpy V-notch impact test, used to define curves of fracture behaviour over a range of temperatures. The data from these tests are used to define the adjusted reference temperature used in determining pressure-temperature operating curves for a nuclear power plant. (author)

  13. Reactor pressure vessel integrity research at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Corwin, W.R.; Pennell, W.E.; Pace, J.V.

    1995-01-01

    Maintaining the integrity of the reactor pressure vessel (RPV) in a light-water-cooled nuclear power plant is crucial in preventing and controlling severe accidents that have the potential for major contamination release. The RPV is the only key safety-related component of the plant for which a duplicate or redundant backup system does not exist. It is therefore imperative to understand and be able to predict the integrity inherent in the RPV. For this reason, the U.S. Nuclear Regulatory Commission has established the related research programs at ORNL described herein to provide for the development and confirmation of the methods used for: (1) establishing the irradiation exposure conditions within the RPV in the Embrittlement Data Base and Dosimetry Evaluation Program, (2) assessing the effects of irradiation on the RPV materials in the Heavy-Section Steel Irradiation Program, and (3) developing overall structural and fracture analyses of RPVs in the Heavy-Section Steel Technology Program

  14. Selected advances in materials research

    International Nuclear Information System (INIS)

    Cunningham, J.E.

    1979-01-01

    Several findings emanating from materials research that should have a beneficial impact on technological advancement in the future are described. The report deals with the GRAPHNOL, a new class of high-temperature brazing alloy for joining refractory components, gel-sphere-pac process for manufacture of nuclear fuel, and noble-metal fuel cladding for service in radioisotope thermoelectric generators designed to provide auxiliary power aboard spacecraft for planetary exploration

  15. Optimization of Composite Material System and Lay-up to Achieve Minimum Weight Pressure Vessel

    Science.gov (United States)

    Mian, Haris Hameed; Wang, Gang; Dar, Uzair Ahmed; Zhang, Weihong

    2013-10-01

    The use of composite pressure vessels particularly in the aerospace industry is escalating rapidly because of their superiority in directional strength and colossal weight advantage. The present work elucidates the procedure to optimize the lay-up for composite pressure vessel using finite element analysis and calculate the relative weight saving compared with the reference metallic pressure vessel. The determination of proper fiber orientation and laminate thickness is very important to decrease manufacturing difficulties and increase structural efficiency. In the present work different lay-up sequences for laminates including, cross-ply [ 0 m /90 n ] s , angle-ply [ ±θ] ns , [ 90/±θ] ns and [ 0/±θ] ns , are analyzed. The lay-up sequence, orientation and laminate thickness (number of layers) are optimized for three candidate composite materials S-glass/epoxy, Kevlar/epoxy and Carbon/epoxy. Finite element analysis of composite pressure vessel is performed by using commercial finite element code ANSYS and utilizing the capabilities of ANSYS Parametric Design Language and Design Optimization module to automate the process of optimization. For verification, a code is developed in MATLAB based on classical lamination theory; incorporating Tsai-Wu failure criterion for first-ply failure (FPF). The results of the MATLAB code shows its effectiveness in theoretical prediction of first-ply failure strengths of laminated composite pressure vessels and close agreement with the FEA results. The optimization results shows that for all the composite material systems considered, the angle-ply [ ±θ] ns is the optimum lay-up. For given fixed ply thickness the total thickness of laminate is obtained resulting in factor of safety slightly higher than two. Both Carbon/epoxy and Kevlar/Epoxy resulted in approximately same laminate thickness and considerable percentage of weight saving, but S-glass/epoxy resulted in weight increment.

  16. Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

    Directory of Open Access Journals (Sweden)

    S.A. Santa

    2017-06-01

    Full Text Available Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of  neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g 60Co, 237Np (n,f 95Zr-103Ru,  45Sc (n,g 46Sc, 58Fe (n,g 59Fe, 181Ta (n,g 182Ta, and 58Ni (n,p58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1

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

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    1994-01-01

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

  18. Mini neutron monitor measurements at the Neumayer III station and on the German research vessel Polarstern

    Science.gov (United States)

    Heber, B.; Galsdorf, D.; Herbst, K.; Gieseler, J.; Labrenz, J.; Schwerdt, C.; Walter, M.; Benadé, G.; Fuchs, R.; Krüger, H.; Moraal, H.

    2015-08-01

    Neutron monitors (NMs) are ground-based devices to measure the variation of cosmic ray intensities, and although being reliable they have two disadvantages: their size as well as their weight. As consequence, [1] suggested the development of a portable, and thus much smaller and lighter, calibration neutron monitor that can be carried to any existing station around the world [see 2; 3]. But this mini neutron monitor, moreover, can also be installed as an autonomous station at any location that provides ’’office” conditions such as a) temperatures within the range of around 0 to less than 40 degree C as well as b) internet and c) power supply. However, the best location is when the material above the NM is minimized. In 2011 a mini Neutron Monitor was installed at the Neumayer III station in Antarctica as well as the German research vessel Polarstern, providing scientific data since January 2014 and October 2012, respectively. The Polarstern, which is in the possession of the Federal Republic of Germany represented by the Ministry of Education and Research and operated by the Alfred Wegener Institute, Helmholtz Centre for Polar and Marine Research and managed by the shipping company Laeisz, was specially designed for working in the polar seas and is currently one of the most sophisticated polar research vessels worldwide. It spends almost 310 days a year at sea usually being located in the waters of Antarctica between November and March while spending the northern summer months in Arctic waters. Therefore, the vessel scans the rigidity range below the atmospheric threshold and above 10 GV twice a year. In contrast to spacecraft measurements NM data are influenced by variations of the geomagnetic field as well as the atmospheric conditions. Thus, in order to interpret the data a detailed knowledge of the instrument sensitivity with geomagnetic latitude (rigidity) and atmospheric pressure is essential. In order to determine the atmospheric response data from the

  19. Mini neutron monitor measurements at the Neumayer III station and on the German research vessel Polarstern

    International Nuclear Information System (INIS)

    Heber, B; Galsdorf, D; Herbst, K; Gieseler, J; Labrenz, J; Schwerdt, C; Walter, M; Benadé, G; Fuchs, R; Krüger, H; Moraal, H

    2015-01-01

    Neutron monitors (NMs) are ground-based devices to measure the variation of cosmic ray intensities, and although being reliable they have two disadvantages: their size as well as their weight. As consequence, [1] suggested the development of a portable, and thus much smaller and lighter, calibration neutron monitor that can be carried to any existing station around the world [see 2; 3]. But this mini neutron monitor, moreover, can also be installed as an autonomous station at any location that provides ’’office” conditions such as a) temperatures within the range of around 0 to less than 40 degree C as well as b) internet and c) power supply. However, the best location is when the material above the NM is minimized. In 2011 a mini Neutron Monitor was installed at the Neumayer III station in Antarctica as well as the German research vessel Polarstern, providing scientific data since January 2014 and October 2012, respectively. The Polarstern, which is in the possession of the Federal Republic of Germany represented by the Ministry of Education and Research and operated by the Alfred Wegener Institute, Helmholtz Centre for Polar and Marine Research and managed by the shipping company Laeisz, was specially designed for working in the polar seas and is currently one of the most sophisticated polar research vessels worldwide. It spends almost 310 days a year at sea usually being located in the waters of Antarctica between November and March while spending the northern summer months in Arctic waters. Therefore, the vessel scans the rigidity range below the atmospheric threshold and above 10 GV twice a year. In contrast to spacecraft measurements NM data are influenced by variations of the geomagnetic field as well as the atmospheric conditions. Thus, in order to interpret the data a detailed knowledge of the instrument sensitivity with geomagnetic latitude (rigidity) and atmospheric pressure is essential. In order to determine the atmospheric response data from the

  20. Ultimate load analysis of prestressed concrete reactor pressure vessels considering a general material law

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1975-01-01

    A method of analysis is presented, by which progressive fracture processes in axisymmetric prestressed concrete pressure vessels during increasing internal pressure can be evaulated by means of a continuum calculation considering a general material law. Formulations used in the analysis concerning material behaviour are derived on one hand from appropriate results of testing small concrete specimens, and are on the other hand gained by parametric studies in order to solve questions still existing by recalulating fracture tests on concrete bodies with more complex states of stress. Due attention is focussed on investigating the behaviour of construction members subjected to high shear forces (end slabs.). (Auth.)

  1. Radioactive material-containing vessel and method of manufacturing the same

    International Nuclear Information System (INIS)

    Kanazawa, Hiroshi; Wada, Katsuyoshi; Ota, Shigeo; Nishioka, Eiji; Okuno, Michinori.

    1995-01-01

    In a vessel for containing radioactive materials having an outer wall with a structure of interposing a lead layer, as a shielding material between inner and outer cylinders made of steel plates, the inner cylinder and the lead layer are in close contact by way of a thin layer of a lead/tin type soldering material and to such an extent that the boundary layer is not detected by supersonic inspection. In addition, flux is coated to the steel plate, which forms the inner cylinder, on the surface being in contact with the lead layer, then a thin layer of the soldering material such as lead or tin is formed, to cast the lead between the inner and the outer cylinders. Then, since the inner cylinder and the lead layer are thermally joined tightly, heat generated at the inside can effectively be released to the outside, so that it is effective as a high-performance cask for transporting a large amount of radioactive materials such as spent nuclear fuels having high temperature afterheat. In addition, a containing vessel with good contact between the inner cylinder and the lead can be manufactured at a low cost only applying a simple primer treatment on the surface of the inner cylinder in addition to an existent lead casting method. (N.H.)

  2. Research to sustain cases for Magnox-reactor steel pressure vessels

    International Nuclear Information System (INIS)

    Graham, W.J.

    1997-01-01

    Britain's Magnox Electric plc owns and operates six power stations, each of which has twin gas-cooled reactors of the Magnox-fuel type. The older group of four power stations has steel pressure-circuits. The reactor cores are housed within spherical, steel vessels. This article describes some of the research which is undertaken to sustain the safety cases for these steel vessels which have now been in operation for just over 30 years. (author) 2 figs., 4 refs

  3. Heavy section steel technology program technical report No. 38. Fracture toughness characterization of HSST intermediate pressure vessel material

    International Nuclear Information System (INIS)

    Mager, T.R.; Yanichko, S.E.; Singer, L.R.

    1974-12-01

    The primary objective of the Heavy Section Steel Technology (HSST) Program is to develop pertinent fracture technology to demonstrate the structural reliability of present and contemplated water-cooled nuclear reactor pressure vessels. In order to demonstrate the ability to predict failure of large, heavy-walled pressure vessels under service type loading conditions, the fracture toughness properties of the vessel's materials must be characterized. The sampling procedure and test results are presented for vessel material supplied by the Oak Ridge National Laboratory that were used to characterize the fracture toughness of the HSST Intermediate Test Vessels. The metallurgical condition and heat treatment of the test material was representative of the vessel simulated service test condition. Test specimen locations and orientations were selected by the Oak Ridge National Laboratory and are representative of flaw orientations incorporated in the test vessels. The fracture toughness is documented for the materials from each of the eight HSST Intermediate Pressure Vessels tested to date. 7 references. (U.S.)

  4. Reactor pressure vessel structural integrity research in the US Nuclear Regulatory Commission HSST and HSSI Programs

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    This report discusses development on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels containing flaws. Fracture mechanics tests on reactor pressure vessel steel have shown that local brittle zones do not significantly degrade the material fracture toughness, constraint relaxation at the crack tip of shallow surface flaws results in increased fracture toughness, and biaxial loading reduces but does not eliminate the shallow-flaw fracture toughness elevation. Experimental irradiation investigations have shown that the irradiation-induced shift in Charpy V-notch versus temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement and the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  5. 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)

  6. From Deck Hand to Program Manager - 30 years with Research Vessels

    Science.gov (United States)

    Prince, J. M.

    2012-12-01

    Starting in 1980 as a Mate and Deck Hand and working my way up to Captain, Marine Superintendent, UNOLS Executive Secretary and now as an ONR Research Facilities Program Manager focused on the acquisition of two new Ocean Class Research Vessels, I have witnessed first hand the evolution of the U.S. Academic Research Fleet. The author will focus on a few key events in the evolution of the modern research fleet. As a deck hand, mate and Captain, I was involved in an early multi-disciplinary effort often using two ships working together to conduct sampling and analysis in Physical, Chemical and Biological oceanography. The VERTEX cruises led by John Martin and others used the R/V CAYUSE and R/V WECOMA extensively through out the NE Pacific Ocean conducting research that led to Dr. Martin's Iron Hypothesis. This work and that of others involving trace metal clean sampling and clean laboratories on board our ships pushed many new and demanding requirements for future vessels. As a ship scheduler and as chair of the Research Vessel Operators Committee (RVOC) I saw the increasing use of Remotely Operated Vehicles to complement the work being done with the ALVIN and other occupied submersibles. This led to scheduling challenges and changes to our safety standards, but also to many new opportunities for discoveries on the many mid-ocean ridges and hydro-thermal vent fields. More recently, Autonomous Underwater Vehicles (AUV) and Unmanned Aerial Vehicles (UAV) and aircraft have been used simultaneously with research vessels such as during a multi-PI, multi-ship program in the Monterey Bay. Communications at sea have changed dramatically in the past thirty years. No longer are we limited to reading the data from a spreadsheet over a Single Side Band radio so that the PI ashore can track the progress of a cruise and provide guidance for the next day's sampling. Full bandwidth communications are becoming the norm with the capability of streaming video from an ROV to shore or to

  7. Application of micromechanical models of ductile fracture initiation to reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Chaouadi, R.; Walle, E. van; Fabry, A.; Velde, J. van de; Meester, P. de

    1996-01-01

    The aim of the current study is the application of local micromechanical models to predict crack initiation in ductile materials. Two reactor pressure vessel materials have been selected for this study: JRQ IAEA monitor base metal (A533B Cl.1) and Doel-IV weld material. Charpy impact tests have been performed in both un-irradiated and irradiated conditions. In addition to standard tensile tests, notched tensile specimens have been tested. The upper shelf energy of the weld material remains almost un-affected by irradiation, whereas a decrease of 20% is detected for the base metal. Accordingly, the tensile properties of the weld material do not reveal a clear irradiation effect on the yield and ultimate stresses, this in contrast to the base material flow properties. The tensile tests have been analyzed in terms of micromechanical models. A good correlation is found between the standard tests and the micromechanical models, that are able to predict the ductile damage evolution in these materials. Additional information on the ductility behavior of these materials is revealed by this micromechanical analysis

  8. Overview of materials research in South Africa

    CSIR Research Space (South Africa)

    Du Preez, W

    2011-09-01

    Full Text Available : Metals and Metals Processes Materials Science and Manufacturing 7 September 2011 ? CSIR 2010 Slide 5 Outline of presentation ? Introduction ? Drivers of Materials Research Since 1996 ? Research Themes and Focus ? CSIR 2010 Slide 6 Introduction...-metal matrix composites ? Piezoelectric materials ? Light metals ? Laser processing of materials ? CSIR 2010 Slide 7 Drivers of Materials Research Since 1996 ? 1996 White Paper on Science and Technology (S&T) ? 1999 Manufacturing/Materials & Mining...

  9. Mechanical properties and examination of cracking in TMI-2 pressure vessel lower head material

    International Nuclear Information System (INIS)

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

    1993-09-01

    Mechanical tests have been conducted on material from 15 samples removed from the lower head of the Three Mile Island unit 2 nuclear reactor pressure vessel. Measured properties include tensile properties and hardness profiles at room temperature, tensile and creep properties at temperatures of 600 to 1200 degrees C, and Charpy V-notch impact properties at -20 to +300 degrees C. These data, which were used in the subsequent analyses of the margin-to-failure of the lower head during the accident, are presented here. In addition, the results of metallographic and scanning electron microscope examinations of cladding cracking in three of the lower head samples are discussed

  10. Users manual data base MATSURV. Reactor pressure vessel material surveillance data management system

    International Nuclear Information System (INIS)

    Kenworthy, L.D.; Tether, C.D.

    1980-02-01

    This Users Guide to the data management system MATSURV has been prepared to assist the user in all facets of the task of processing data related to reactor pressure vessel materials surveillance; preparation of raw data for input, input of data, modification of existing data, retrieval and display of data, and the creation of data reports. MATSURV is structured upon the System 2000 data base management system which is maintained on the IBM 370/168 computer at National Institutes of Health. An overview of System 2000 is provided

  11. Research and development of spent fuel shipping casks and the criteria for seagoing vessel carrying casks

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Considering that the transportation of spent fuel will increase rapidly and extensively in the near future, Japanese Atomic Energy Committee enacted ''Technical Standard for Transportation of Radioactive Materials'' based on ''IAEA Regulation for the Safe Transport of Radioactive Materials 1973 Revised Edition''. Coping with the recommendation of AEC, Atomic Energy Bureau in Science and Technology Agency and other authorities concerned started to review the former ordinances for transportation of radioactive materials and to consolidate a unified system of relevant laws and standards. On the other hand, Atomic Energy Bureau has invested in research and development since ten years ago in order to obtain the data for design and licensing work of spent fuel shipping casks. In those studies some different scale models of a prototype of 80 t in weight have been used to make clear the scale effect at the drop, pucture and fire tests, which are one of the features of Japanese research and development. And also the immersion test in high pressure water up to about 500 bars is now carried out to investigate the integrity of cask body and sealing structure to prevent leakage of radioactive contents to the ambient when the cask falls into deep sea. In Japan, depending on the site conditions of nuclear plants, almost all transportations of unirradiated and spent fuels are done on the sea. Therefore, in order to secure the safety of transportation, the design criteria of the seagoing vessels for exclusive transportation of spent fuel shipping casks, namely full load shipping, has been enacted, which aims to make minimum the probability of sinking at collison, stranding and other unforeseen accidents at sea and also to restrain radiation exposure of the crew as low as possible

  12. 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

  13. The significance of cladding material on the integrity of nuclear pressure vessels with cracks

    International Nuclear Information System (INIS)

    Sattari-Far, Iradj.

    1989-05-01

    The significance of the austenitic cladding layer is reviewed in this literature study. The cladding induced stresses are generally not considered when evaluating the severity of flaws in reactor pressure vessels. It has been shown that this emission may be misleading. The necessity to consider the cladding induced stresses is also emphasized in the latest edition of ASME XI. Contrary to what is commonly assumed, the austenitic cladding displays a charpy V transition region with a low ductility. The interface material (HAZ) is the most influenced region by irradiation, and a transition shift of over 100 degree C may be expected. Because of the significant difference in the thermal expansion coefficients of the cladding and the base metal, cladding induced stresses can be set up. Even after PWHT, residual stresses of yield magnitude remain in the cladding and the HAZ at ambient temperature. The cladding induced stresses are temperature dependent and decrease as the temperature increases. The cladding induced stresses have a significant influence on small defects near the inside surface of a pressure vessel. For semielliptical surface cracks, the maximum CTOD-value along the crack front is not found at the deepest point, but in the cladding/base metal interface, having a magnitude three times higher than the value in the deepest point. It implies that this type of crack would propagate along the clad/base material interface. At some point in time, the crack will reach a geometry which may cause such a severe condition at the deepest point that it will start to grow in the depth direction as well. The initiation and growth behaviour of such cracks need to be investigated to be able to assess the significance of cladding on the integrity of nuclear pressure vessels. (author) (50 figs., 33 refs.)

  14. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    Results of reactor-material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address ex-vessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debris characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity

  15. Materials Research Department Annual report 1998

    Energy Technology Data Exchange (ETDEWEB)

    Winther, Grethe; Hansen, N [eds.

    1999-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1998 are described. The scientific work is presented in five chapters: Materials Science, Materials Engineering, Materials Technology, Materials Chemistry and Fusion Materials. A survey is given of the Departments collaboration with national and international industries and research institutions. Furthermore, the main figures outlining the funding and expenditure of the Department are given. Lists of staff members, visiting scientists and educational activities are included. (au) 165 refs.

  16. Materials Research Department Annual report 1998

    International Nuclear Information System (INIS)

    Winther, Grethe; Hansen, N.

    1999-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1998 are described. The scientific work is presented in five chapters: Materials Science, Materials Engineering, Materials Technology, Materials Chemistry and Fusion Materials. A survey is given of the Departments collaboration with national and international industries and research institutions. Furthermore, the main figures outlining the funding and expenditure of the Department are given. Lists of staff members, visiting scientists and educational activities are included. (au)

  17. Materials research with ion beams

    International Nuclear Information System (INIS)

    Meyer, J.D.

    1988-01-01

    This report gives a series of helpful programs which are used in materials research with ion beams. In this context algorithms which can substitute table books are dealt with. This is true for the programs DEDX and PRAL; they are used in order to determine the energy loss of ions in solid bodies, their working range and straggling. Furthermore, simulator routines and analyzers are described. The program TRIM simulates the physical phenomena which occur with the penetration of high-energy ions into solid bodies. In this context electronic excitations, phonons and lattice distortions which are caused by the ions are dealt with. For the experimental ion implantation it is interesting to know the final distribution of the simulated ions in the solid body. The program RBS simulates the Rutherford spectrum of ions which are scattered from a solid body which may consist of up to nine elements and up to one hundred layers. The unknown composition of a solid body can be determined in direct comparison with the experimental spectrum. The program NRA determines concentration and penetrative distribution of an impurity by means of the experimental nuclear reaction spectrum of this impurity. All programs are written in FORTRAN 77. (orig./MM) [de

  18. Fracture toughness requirements of reactor vessel material in evaluation of the safety analysis report of nuclear power plants

    International Nuclear Information System (INIS)

    Widia Lastana Istanto

    2011-01-01

    Fracture toughness requirements of reactor vessel material that must be met by applicants for nuclear power plants construction permit has been investigated in this paper. The fracture toughness should be described in the Safety Analysis Reports (SARs) document that will be evaluated by the Nuclear Energy Regulatory Agency (BAPETEN). Because BAPETEN does not have a regulations or standards/codes regarding the material used for the reactor vessel, especially in the fracture toughness requirements, then the acceptance criteria that applied to evaluate the fracture toughness of reactor vessel material refers to the regulations/provisions from the countries that have been experienced in the operation of nuclear power plants, such as from the United States, Japan and Korea. Regulations and standards used are 10 CFR Part 50, ASME and ASTM. Fracture toughness of reactor vessel materials are evaluated to ensure compliance of the requirements and provisions of the Regulatory Body and the applicable standards, such as ASME or ASTM, in order to assure a reliability and integrity of the reactor vessels as well as providing an adequate safety margin during the operation, testing, maintenance, and postulated accident conditions over the reactor vessel lifetime. (author)

  19. Development of in-vessel type control rod drive mechanism for a innovative small reactor (Contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Yoritsune, Tsutomu; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Although the control rod drive mechanism of an existing large scale light water reactor is generally installed outside the reactor vessel, an in-vessel type control rod drive mechanism (INV-CRDM) is installed inside the reactor vessel. The INV-CRDM contributes to compactness and simplicity of the reactor system, and it can eliminate the possibility of a rod ejection accident. Therefore, INV-CRDM is an important technology adopted in an innovative small reactor. Japan Atomic Energy Research Institute (JAERI) has developed this type of CRDM driven by an electric motor, which can work under high temperature and high pressure water for the advanced marine reactor. On the basis of this research result, a driving motor coil and a bearing were developed to be used under the high temperature steam, severe condition for an innovative small reactor. About the driving motor, we manufactured the driving motor available for high temperature steam and carried out performance test under room temperature atmosphere to confirm the electric characteristic and coolability of the driving coil. With these test results and the past test results under high temperature water, we analyzed and evaluated the electric performance and coolability of the driving coil under high temperature steam. Concerning bearing, we manufactured the test pieces using some candidate material for material characteristic test and carried out the rolling wear test under high temperature steam to select the material. Consequently, we confirmed that performance of the driving coil for the advanced type driving motor, is enough to be used under high temperature steam. And, we evaluated the performance of the bearing and selected the material of the bearing, which can be used under high temperature steam. From these results, we have obtained the prospect that the INV-CRDM can be used for an innovative small reactor under steam atmosphere could be developed. (author)

  20. Compilation of contract research for the Materials Engineering Branch, Division of Engineering Technology. Annual report for FY 1982

    International Nuclear Information System (INIS)

    1983-03-01

    This report presents summaries of the research work performed during Fiscal Year 1982 by laboratories and organizations under contracts administered by the NRC's Materials Engineering Branch, Office of Nuclear Regulatory Research. The contractor reports are organized into the major areas of concern to Primary System Integrity: Vessel and Piping Fracture Mechanics; Pressure Vessel Surveillance Dosimetry; Steam Generators and Environmental Cracking; and Nondestructive Examination

  1. Concept of a nuclear powered submersible research vessel and a compact reactor

    International Nuclear Information System (INIS)

    Kusunoki, Tsuyoshi; Odano, Naoteru; Yoritsune, Tsutomu; Ishida, Toshihisa; Nishimura, Hajime; Tokunaga, Sango

    2001-07-01

    A conceptual design study of a submersible research vessel navigating in 600 m depth and a compact nuclear reactor were carried out for the expansion of the nuclear power utilization. The mission of the vessel is the research of mechanism of the climate change to predict the global environment. Through conditions of the Arctic Ocean and the sea at high latitude have significant impacts on the global environmental change, it is difficult to investigate those areas by ordinary ships because of thick ice or storm. Therefore the research vessel is mainly utilized in the Arctic Ocean and the sea at high latitude. By taking account of the research mission, the basic specifications of the vessel are decided; the total weight is 500 t, the submersible depth is 600 m, the maximum speed is 12 knots (22.2 km/h), and the number of crews is 16. Nuclear power has an advantage in supplying large power of electricity in the sea for long period. Based on the requirements, it has been decided that two sets of submersible compact reactor, SCR, which is light-weighted and of enhanced safety characteristics of supply the total electricity of 500 kW. (author)

  2. Design and preliminary analysis of in-vessel core catcher made of high-temperature ceramics material in PWR

    International Nuclear Information System (INIS)

    Xu Hong; Ma Li; Wang Junrong; Zhou Zhiwei

    2011-01-01

    In order to protect the interior wall of pressure vessel from melting, as an additional way to external reactor vessel cooling (ERVC), a kind of in-vessel core catcher (IVCC) made of high-temperature ceramics material was designed. Through the high-temperature and thermal-resistance characteristic of IVCC, the distributing of heat flux was optimized. The results show that the downward average heat flux from melt in ceramic layer reduces obviously and the interior wall of pressure vessel doesn't melt, keeping its integrity perfectly. Increasing of upward heat flux from metallic layer makes the upper plenum structure's temperature ascend, but the temperature doesn't exceed its melting point. In conclusion, the results indicate the potential feasibility of IVCC made of high-temperature ceramics material. (authors)

  3. Advance of investigation of irradiation embrittlement mechanism of nuclear reactor pressure vessel steels. History and future of irradiation embrittlement researches

    International Nuclear Information System (INIS)

    Ishino, Shiori

    2007-01-01

    The nuclear reactor pressure vessel is the most important component of LWR plants required to be safe. This paper describes contents of the title consisting of four chapters. The first chapter states the general theory of irradiation effects, irradiation embrittlement and decreasing of toughness, and some kinds of pressure vessel steels. The second chapter explains history of irradiation embrittlement investigations and the advance of research methods for experiments and calculation. The third chapter contains information of inner structure of irradiated materials and development of prediction equations, recent information of embrittlement mechanism and mechanism guided prediction method, USA model and Central Research Institute of Electric Power Industry (CRIEPI) model. The fourth chapter states recent problems from viewpoints of experimental and analytical approaches. Comparison of standards of LWR pressure vessel steels between Japan and USA, relation between the density of number of cluster and the copper content, effect of flux on clustering of copper atoms, and CRIEPI's way of approaching the prediction method are illustrated. (S.Y.)

  4. Compilation of contract research for the Materials Engineering Branch, Division of Engineering

    International Nuclear Information System (INIS)

    1991-03-01

    This compilation of annual reports for FY 1990 by contractors to the Materials Engineering Branch of the Nuclear Regulatory Commission Office of Research concentrates on achievements in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators, and nondestructive examination of primary system components. Separate abstracts have been prepared for each of the reports which are divided into the following categories: (1) vessel and piping fracture mechanics (including irradiation embrittlement); (2) pressure vessel surveillance dosimetry; (3) steam generators, aging, and environmental cracking; and (4) nondestructive examination techniques

  5. Materials research in the Nuclear Research Centre Karlsruhe

    International Nuclear Information System (INIS)

    Kleykamp, H.

    1990-03-01

    This report gives a survey of the research work done at the Institute for Material and Solids Research at Karlsruhe. The following subjects are dealt with: Instrumental analysis; producing thin films; corrosion; failure mechanism and damage analysis; fuel elements, ceramic nuclear fuels and can and structure materials for fast breeder reactors; material problems and ceramic breeding materials for nuclear fusion plants; glass materials for the treatment of radioactive waste; super-conducting materials; amorphous metals, new high alloyed steels; ceramic high performance materials; hard materials; compound materials and polymers. (MM) [de

  6. Assessment of Radiation Embrittlement in Nuclear Reactor Pressure Vessel Surrogate Materials

    Science.gov (United States)

    Balzar, Davor

    2010-10-01

    The radiation-enhanced formation of small (1-2 nm) copper-rich precipitates (CRPs) is critical for the occurrence of embrittlement in nuclear-reactor pressure vessels. Small CRPs are coherent with the bcc matrix, which causes local matrix strain and interaction with the dislocation strain fields, thus impeding dislocation mobility. As CRPs grow, there is a critical size at which a phase transformation occurs, whereby the CRPs are no longer coherent with the matrix, and the strain is relieved. Diffraction-line-broadening analysis (DLBA) and small-angle neutron scattering (SANS) were used to characterize the precipitate formation in surrogate ferritic reactor-pressure vessel steels. The materials were aged for different times at elevated temperature to produce a series of specimens with different degrees of copper precipitation. SANS measurements showed that the precipitate size distribution broadens and shifts toward larger sizes as a function of ageing time. Mechanical hardness showed an increase with ageing time, followed by a decrease, which can be associated with the reduction in the number density as well as the loss of coherency at larger sizes. Inhomogeneous strain correlated with mechanical hardness.

  7. Generic analyses for evaluation of low Charpy upper-shelf energy effects on safety margins against fracture of reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Dickson, T.L.

    1993-07-01

    Appendix G to 10 CFR Part 50 requires that reactor pressure vessel beltline material maintain Charpy upper-shelf energies of no less than 50 ft-lb during the plant operating life, unless it is demonstrated in a manner approved by the Nuclear Regulatory Commission (NRC), that lower values of Charpy upper-shelf energy provide margins of safety against fracture equivalent to those in Appendix G to Section XI of the ASME Code. Analyses based on acceptance criteria and analysis methods adopted in the ASME Code Case N-512 are described herein. Additional information on material properties was provided by the NRC, Office of Nuclear Regulatory Research, Materials Engineering Branch. These cases, specified by the NRC, represent generic applications to boiling water reactor and pressurized water reactor vessels. This report is designated as HSST Report No. 140

  8. Materials and Waste Management Research

    Science.gov (United States)

    EPA is developing data and tools to reduce waste, manage risks, reuse and conserve natural materials, and optimize energy recovery. Collaboration with states facilitates assessment and utilization of technologies developed by the private sector.

  9. Reactor pressure vessel steels[1997 Scientific Report of the Belgian Nuclear Research Centre

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

  10. Assessment of aluminum structural materials for service within the ANS reflector vessel

    International Nuclear Information System (INIS)

    Farrell, K.

    1995-08-01

    Most of the components in the Advanced Neutron Source (ANS) reactor, including the reflector vessel, will be built from the aluminum alloy 6061 (lMg,0.6Si) in its precipitation-hardened T6 and T651 conditions. The microstructural and mechanical characteristics of the alloy are described, and its operating boundaries of stress, temperature, and time in its unirradiated state are defined. The material's responses to neutron radiation exposure in aqueous environments are reviewed in detail. The particular service conditions of stress, temperature, and radiation exposure expected for individual components in the ANS are listed, and the suitability of each component to meet the demands is assessed. Areas of uncertainties are outlined, and various suggestions and recommendations are made to give improved confidence in the predictions

  11. Contribution of the different erosion processes to material release from the vessel walls of fusion devices during plasma operation

    International Nuclear Information System (INIS)

    Behrisch, R.

    2002-01-01

    In high temperature plasma experiments several processes contribute to erosion and loss of material from the vessel walls. This material may enter the plasma edge and the central plasma where it acts as impurities. It will finally be re-deposited at other wall areas. These erosion processes are: evaporation due to heating of wall areas. At very high power deposition evaporation may become very large, which has been named ''blooming''. Large evaporation and melting at some areas of the vessel wall surface may occur during heat pulses, as observed in plasma devices during plasma disruptions. At tips on the vessel walls and/or hot spots on the plasma exposed solid surfaces electrical arcs between the plasma and the vessel wall may ignite. They cause the release of ions, atoms and small metal droplets, or of carbon dust particles. Finally, atoms from the vessel walls are removed by physical and chemical sputtering caused by the bombardment of the vessel walls with ions as well as energetic neutral hydrogen atoms from the boundary plasma. All these processes have been, and are, observed in today's plasma experiments. Evaporation can in principle be controlled by very effective cooling of the wall tiles, arcing is reduced by very stable plasma operation, and sputtering by ions can be reduced by operating with a cold plasma in front of the vessel walls. However, sputtering by energetic neutrals, which impinge on all areas of the vessel walls, is likely to be the most critical process because ions lost from the plasma recycle as neutrals or have to be refuelled by neutrals leading to the charge exchange processes in the plasma. In order to quantify the wall erosion, ''materials factors'' (MF) have been introduced in the following for the different erosion processes. (orig.)

  12. Historical summary of the heavy-section steel technology program and some related activities in light-water reactor pressure vessel safety research

    International Nuclear Information System (INIS)

    Whitman, G.D.

    1986-03-01

    The accomplishments of the Heavy-Section Steel Technology Program and other programs having a close relationship to the development of information used in the assessment of light-water reactor pressure vessel integrity are reviewed. The early Pressure Vessel Research Committee planning, the principals contributing to program formulation, the role of the US Atomic Energy Commission, and the developments under the US Nuclear Regulatory Commission sponsorship are identified. The need for major research and development accomplishments in fracture mechanics, heavy-section steel procurement, materials properties, irradiation effects, fatigue crack growth, and structural testing are summarized. The impact of program results on regulatory issues and the development of data used in the preparation of codes, standards, and guides are discussed. Continuing activities and recommendations for future research and development in support of pressure vessel integrity assessments are presented

  13. Research Vessel R/V Sikuliaq: Joining the UNOLS Fleet in 2014

    Science.gov (United States)

    Whitledge, T. E.

    2013-12-01

    The global class research vessel R/V Sikuliaq is being constructed on behalf of the NSF to support future scientific studies in high latitude waters. The 261 foot vessel will be capable of breaking 2.5 foot thick ice at 2 knots with an endurance of 45 days at sea and cruising at 11 knots. The R/V Sikuliaq has a beam of 52 feet and a draft of 18.9 feet that will carry 26 scientists and a crew of 20. Berthing accommodations are a combination of single/double rooms with one stateroom and the common areas of the vessel are designed for ADA access and accommodations. The total laboratory space (main, analytical, electronics, wet, upper, and Baltic room are 2100 square feet. The 4360 square foot working deck that is approximately 70 feet in length will accommodate 2-4 vans and multiple science operations. The vessel design strives to have the lowest possible environmental impact, including a low underwater-radiated noise signature. The science systems are prescribed to be state-of-the-art for bottom mapping, over-the-side 'hands free' gear handling, broad band communications and scientific walk-in freezer and environmental chamber. More details and photos of the construction progress are available on the website at www.sfos.uaf.edu/arrv. The vessel was launched in October 2012 and delivery to the University of Alaska Fairbanks is scheduled for November 2013. Scientific operations following testing and science sea trials are planned to start in summer of 2014. Questions about the science systems or vessel capabilities should be directed to Terry Whitledge (terry@ims.uaf.edu).

  14. Materials Research Department annual report 2000

    International Nuclear Information System (INIS)

    Winther, G.; Hansen, N.

    2001-03-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 2000 are described. The scientific work is presented in three chapters: Materials Science, Materials Engineering and Materials Technology. A survey is given of the Department's industrial collaboration, educational activities and academic activities, such as collaboration with other research institutions, committee work and a list of publications. Furthermore, the main figures outlining the funding and expenditures of the Department are given. Lists of staff members and visiting scientists are included. (au)

  15. Materials irradiation research in neutron science

    Energy Technology Data Exchange (ETDEWEB)

    Noda, Kenji; Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    Materials irradiation researches are planned in Neutron Science Research Program. A materials irradiation facility has been conceived as one of facilities in the concept of Neutron Science Research Center at JAERI. The neutron irradiation field of the facility is characterized by high flux of spallation neutrons with very wide energy range up to several hundred MeV, good accessibility to the irradiation field, good controllability of irradiation conditions, etc. Extensive use of such a materials irradiation facility is expected for fundamental materials irradiation researches and R and D of nuclear energy systems such as accelerator-driven incineration plant for long-lifetime nuclear waste. In this paper, outline concept of the materials irradiation facility, characteristics of the irradiation field, preliminary technical evaluation of target to generate spallation neutrons, and materials researches expected for Neutron Science Research program are described. (author)

  16. Materials Research Department annual report 1997

    Energy Technology Data Exchange (ETDEWEB)

    Soerensen, B F; Hansen, N [eds.

    1998-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1997 are described. The scientific work is presented in four chapters: Materials Science, Materials Chemistry, Materials Engineering and Materials Technology. A survey is given of the Department`s participation in international collaboration and of its activities within education and training. Furthermore, the main figures outlining the funding and expenditure of the Department are given. Lists of staff members, visiting scientists, publications and other Department activities are included. (au) 278 refs.

  17. Materials Research Department annual report 1996

    International Nuclear Information System (INIS)

    Soerensen, B.F.; Hansen, N.

    1997-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1996 are described. The scientific work is presented in four chapters: Materials Science, Materials Chemistry, Materials Engineering and Materials Technology. A survey is given of the Department's participation in international collaboration and of its activities within education and training. Furthermore, the main figures outlining the funding and expenditure of the Department are given. Lists of staff members, visiting scientists, publications and other Department activities are included. (au)

  18. Materials Research Department annual report 1997

    International Nuclear Information System (INIS)

    Soerensen, B.F.; Hansen, N.

    1998-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1997 are described. The scientific work is presented in four chapters: Materials Science, Materials Chemistry, Materials Engineering and Materials Technology. A survey is given of the Department's participation in international collaboration and of its activities within education and training. Furthermore, the main figures outlining the funding and expenditure of the Department are given. Lists of staff members, visiting scientists, publications and other Department activities are included. (au)

  19. A group of painted vessels from Singidunum: A contribution to the researches on painted ceramics

    Directory of Open Access Journals (Sweden)

    Nikolić Snežana

    2005-01-01

    Full Text Available About 20 vessels, made of fine clay fired in whitish tones (10YR 8/2, 10YR 8/2-3, 5Y 8/1, with the polished surface ornamented with painting in fading brown, originate from Singidunum. In comparison with analogous material from Donja (Lower Panonia and Dalmatia, the importance of these vessels is to be found in the fact that they were excavated from settlement horizons dated to the second half of the 3rd and early 4th century. Based on the shapes and technological features of ceramics from Lower Panonia and Dalmatia, which have been published, as well as on the observations of the finds from Singidunum, it is to be assumed that they were the output of the same workshop which not only had a small scale of production but also a meager scope of shapes, meaning goblets i.e. cups as favorable form.

  20. Applying the TOC Project Management to Operation and Maintenance Scheduling of a Research Vessel

    Science.gov (United States)

    Manti, M. Firdausi; Fujimoto, Hideo; Chen, Lian-Yi

    Marine research vessels and their systems are major assets in the marine resources development. Since the running costs for the ship are very high, it is necessary to reduce the total cost by an efficient scheduling for operation and maintenance. To reduce project period and make it efficient, we applied TOC project management method that is a project management approach developed by Dr. Eli Goldratt. It challenges traditional approaches to project management. It will become the most important improvement in the project management since the development of PERT and critical path methodologies. As a case study, we presented the marine geology research project for the purpose of operations in addition to repair on the repairing dock projects for maintenance of vessels.

  1. Report on material and fabrication tests of the KUHFR core vessel

    International Nuclear Information System (INIS)

    Yoshida, H.; Kozuka, T.; Achiwa, N.; Mitani, S.; Kawano, S.; Araki, Y.; Shibata, T.

    1983-01-01

    For the material of the cylindrical reactor core vessel of the Kyoto University High Flux Reactor (KUHFR), A6061 alloy is selected because the aged state of the alloy is known to show the highest resistance against void swelling due to high-dose irradiation. The fabrication possibility of the large-scale tubes is also tested because the sizes (40 cmdiameter and 43 cmdiameter x 960 cm with a thickness of 10 mm for the inner- and outer-tubes, respectively) are just over the largest limit of the conventional factory fabrication. The results are summarized as follows. (1) From an ingot of A6061 alloy a raw inner-tube is hot-extruded by the 3,000 ton press machine. The shape of the extruded tubes is effectively corrected by stretch forming and other special methods. (2) The real scale tubes are heat-treated under the various conditions (T1, T4 and T6) and their size changes are measured just after the every heat-treatment. (3) The hydropressure for a pipe prepared by welding from an aged-tube shows a fairly uniform strain distribution and the breaking initiation at the reasonable pressure in the welded part. (4) Each of the welded specimens prepared using three kinds of welding rods shows sufficient strength in both of bending and tensile test for the JIS standard. Their microstructures correspond to the result of the mechanical tests for each welded specimen. The confidence for the fabrication possibility of the real core vessel has been given through the present tests. (author)

  2. Endurance test report of rubber sealing materials for the containment vessel

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  3. Results and analysis of reactor-material experiments on ex-vessel corium quench and dispersal

    International Nuclear Information System (INIS)

    Spencer, B.W.; McUmber, L.M.; Sienicki, J.J.; Squarer, D.

    1984-01-01

    The results of reactor material experiments and related analysis are described in which molten corium is injected into a mock-up of the reactor cavity region of a PWR. The experiments address exvessel interactions such as steam generation (for those cases in which water is present), water and corium dispersal from the cavity, hydrogen generation, direct atmosphere heating by dispersed corium, and debrids characterization. Test results indicate efficiencies of steam generation by corium quench ranging up to 65%. Corium sweepout of up to 62% of the injected material was found for those conditions in which steam generation flowrate was augmented by vessel blowdown. The dispersed corium caused very little direct heating of the atmosphere for the configuration employing a trap at the exit of the cavity-to-containment pathway. Corium sweepout phenomena were modeled for high-pressure blowdown conditions, and the results applied to the full-size reactor system predict essentially complete sweepout of corium from the reactor cavity. (orig.)

  4. Materials Processing Research and Development

    Science.gov (United States)

    2010-08-01

    2 2.1.4 The Origins of Microstexture in Duplex Ti Alloys...Controlled Growth and Coarsening ....... 14 2.11 PUBLISHED RESEARCH ON FRICTION STIR WELDING OF SC-MODIFIED AL-ZN-MG-CU EXTRUDED PLATES...14 2.11.1 Friction Stir Welding of Sc

  5. An innovative methodology for the transmission of information, using Sensor Web Enablement, from ongoing research vessels.

    Science.gov (United States)

    Sorribas, Jordi; Sinquin, Jean Marc; Diviacco, Paolo; De Cauwer, Karien; Danobeitia, Juanjo; Olive, Joan; Bermudez, Luis

    2013-04-01

    Research vessels are sophisticated laboratories with complex data acquisition systems for a variety of instruments and sensors that acquire real-time information of many different parameters and disciplines. The overall data and metadata acquired commonly spread using well-established standards for data centers; however, the instruments and systems on board are not always well described and it may miss significant information. Thus, important information such as instrument calibration or operational data often does not reach to the data center. The OGC Sensor Web Enablement standards provide solutions to serve complex data along with the detailed description of the process used to obtain them. We show an innovative methodology on how to use Sensor Web Enablement standards to describe and serve information from the research vessels, the data acquisition systems used onboard, and data sets resulting from the onboard work. This methodology is designed to be used in research vessels, but also applies to data centers to avoid loss of information in between The proposed solution considers (I) the difficulty to describe a multidisciplinary and complex mobile sensor system, (II) it can be easily integrated with data acquisition systems onboard, (III) it uses the complex and incomplete typical vocabulary in marine disciplines, (IV) it provides contacts with the data and metadata services at the Data Centers, and (V) it manages the configuration changes with time of the instrument.

  6. Research Vessel R/V Sikuliaq: A New Asset For The UNOLS Fleet

    Science.gov (United States)

    Whitledge, T. E.

    2012-12-01

    The research vessel R/V Sikuliaq is currently being constructed on behalf of the NSF to support future scientific studies in high latitude waters. The 261 foot global class vessel will be capable of breaking 2.5 foot thick ice at 2 knots with an endurance of 45 days at sea and cruising at 11 knots. The R/V Sikuliaq will have a beam of 52 feet and a draft of 18.9 feet that will carry 26 scientists and a crew of 20. Berthing accommodations are a combination of single/double rooms with one stateroom and the common areas of the vessel are designed for ADA access and accommodations. The total laboratory space (main, analytical, electronics, wet, upper, and Baltic room will be 2100 square feet. The 4360 square foot working deck that is approximately 70 feet in length will accommodate 2-4 vans and multiple science operations. The vessel design strives to have the lowest possible environmental impact, including a low underwater-radiated noise signature. The science systems are prescribed to be state-of-the-art for bottom mapping, over-the-side "hands free" gear handling, broad band communications and scientific walk-in freezer and environmental chamber. More details and photos of the construction progress are available on the website at www.sfos.uaf.edu/arrv. The shipyard schedule has a launch date of October 2012 and delivery to the University of Alaska Fairbanks in July 2013. Scientific operations following trials and testing is planned to start in January 2014. Questions about the science systems or vessel capabilities should be directed to Terry Whitledge (terry@ims.uaf.edu).;

  7. In-vessel tritium

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  8. Fusion reactor materials research in China

    International Nuclear Information System (INIS)

    Qian Jiapu

    1994-10-01

    The fusion materials research in China is introduced. Many kinds of structural materials (such as Ti-modified stainless steel, ferritic steel, HT-9, HT-7, oxide dispersion strengthening ferritic steel), tritium breeders (lithium, Li 2 O, γ-LiAlO 2 ) and plasma facing materials (PFMs) (graphite with TiC and SiC coatings) have been developed or being developed. A systematic research activities on irradiation effects, compatibility, plasma materials interaction, thermal shock during disruption, tritium production, release and permeation, neutron multiplication in Be and Pb, etc. have been performed. The research activities are summarized and some experimental results are also given

  9. Materials research in AECL, Spring 1970

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1970-05-15

    This report gives a summary of materials research at Atomic Energy of Canada Limited. The topics covered in this report include engineering design with brittle materials, texture and mechanical properties of zirconium alloy tubing, structural damage by ion bombardment, research on silicon carbide, shallow phosphorus diffusion in p-type silicon and scanning electron microscopy. CRNL facilities for the examination of irradiated materials is also discussed.

  10. Materials research in AECL, Spring 1970

    International Nuclear Information System (INIS)

    1970-05-01

    This report gives a summary of materials research at Atomic Energy of Canada Limited. The topics covered in this report include engineering design with brittle materials, texture and mechanical properties of zirconium alloy tubing, structural damage by ion bombardment, research on silicon carbide, shallow phosphorus diffusion in p-type silicon and scanning electron microscopy. CRNL facilities for the examination of irradiated materials is also discussed

  11. Safety Research Experiment Facility Project. Conceptual design report. Volume V. Reactor vessel and closure

    International Nuclear Information System (INIS)

    1975-12-01

    The Prestressed Concrete Reactor Vessel (PCRV) will serve as the primary pressure retaining structure for the Safety Research Experiment Facility (SAREF) reactor. The reactor core, control rod drive room, primary heat exchangers, and gas circulators will be located in cavities within the PCRV. The orientation of these cavities, except for the control rod drive room, will be similar to the high-temperature gas-cooled reactor (HTGR) designs that are currently proposed or under design. Due to the nature of this type of structure, all biological and radiological shielding requirements are incorporated into the basic vessel design. At the midcore plane there are three radially oriented slots that will extend from the outside surface of the PCRV to the reactor core liner. These slots will accommodate each of the fuel motion monitoring systems which will be part of the observation apparatus used with the loop experiments

  12. Research study of conjugate materials; Conjugate material no chosa kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

    The paper reported an introductory research on possibilities of new glass `conjugate materials.` The report took up the structure and synthetic process of conjugate materials to be researched/developed, classified them according to structural elements on molecular, nanometer and cluster levels, and introduced the structures and functions. Further, as glasses with new functions to be proposed, the paper introduced transparent and high-strength glass used for houses and vehicles, light modulation glass which realizes energy saving and optical data processing, and environmentally functional glass which realizes environmental cleaning or high performance biosensor. An initial survey was also conducted on rights of intellectual property to be taken notice of in Japan and abroad in the present situation. Reports were summed up and introduced of Osaka National Research Institute, Electrotechnical Laboratory, and National Industrial Research Institute of Nagoya which are all carrying out leading studies of conjugate materials. 235 refs., 135 figs., 6 tabs.

  13. European Fusion Materials Research Program - Recent Results and Future Strategy

    International Nuclear Information System (INIS)

    Diegele, E.; Andreani, R.; Laesser, R.; Schaaf, B. van der

    2005-01-01

    The paper reviews the objectives and the status of the current EU long-term materials program. It highlights recent results, discusses some of the key issues and major existing problems to be resolved and presents an outlook on the R and D planned for the next few years. The main objectives of the Materials Development program are the development and qualification of reduced activation structural materials for the Test Blanket Modules (TBMs) in ITER and of low activation structural materials resistant to high fluence neutron irradiation for in-vessel components such as breeding blanket, divertor and first wall in DEMO. The EU strategy assumes: (i) ITER operation starting in 2015 with DEMO relevant Test Blanket Modules to be installed from day one of operation, (ii) IFMIF operation in 2017 and (iii) DEMO final design activities in 2022 to 2025. The EU candidate structural material EUROFER for TBMs has to be fully code qualified for licensing well before 2015. In parallel, research on materials for operation at higher temperatures is conducted following a logical sequence, by supplementing EUROFER with the oxide dispersion strengthened ferritic steels and, thereafter, with fibre-reinforced Silicon Carbide (SiC f /SiC). Complementary, tungsten alloys are developed as structural material for high temperature applications such as gas-cooled divertors

  14. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (Capsule 2) of Ulchin Nuclear Power Plant Unit 4

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2007-04-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin Unit 4 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.306E+18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.918 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.615E+18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 8.478E+18 and 1.673E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin Unit 4 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  15. Final Report for the 1st Surveillance Test of the Reactor Pressure Vessel Material (CAPSULE 2) of Ulchin Nuclear Power Plant Unit 3

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai (and others)

    2006-12-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 1st surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejon after the capsule was transported from Ulchin site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Ulchin unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsule 2 is 4.674E 18n/cm{sup 2}. The bias factor, the ratio of calculation/measurement, was 0.920 for the 1st testing and the calculational uncertainty,7.0% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 3.913E 18n/cm{sup 2} based on the end of 6th fuel cycle and it was predicted that the fluences of vessel inside surface at 16 and 32EFPY would reach 9.249E 18 and 1.834E 19n/cm{sup 2} based on the current calculation. The result through this analysis for Ulchin unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life.

  16. Materials Research Department annual report 1999

    DEFF Research Database (Denmark)

    Sørensen, Bent F.; Hansen, Niels

    2000-01-01

    with national and international industries and research institutions and of its activities within education and training. Furthermore, the main figures outlining the funding and expenditures of theDepartment are given. Lists of staff members, visiting scientists, publications and other Department activities......Selected activities of the Materials Research Department at Risø National Laboratory during 1999 are described. The scientific work is presented in three chapters: Materials Science, Materials Engineering and Materials Technology. A survey is given ofthe Department's participation in collaboration...

  17. Materials Research Department annual report 2000

    Energy Technology Data Exchange (ETDEWEB)

    Winther, G.; Hansen, N. [eds.

    2001-03-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 2000 are described. The scientific work is presented in three chapters: Materials Science, Materials Engineering and Materials Technology. A survey is given of the Department's industrial collaboration, educational activities and academic activities, such as collaboration with other research institutions, committee work and a list of publications. Furthermore, the main figures outlining the funding and expenditures of the Department are given. Lists of staff members and visiting scientists are included. (au)

  18. Materials Research Department annual report 1999

    Energy Technology Data Exchange (ETDEWEB)

    Soerensen, B.F.; Hansen, N. [eds.

    2000-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1999 are described. The scientific work is presented in three chapters: Materials Science, Materials Engineering and Materials Technology. A survey is given of the Department's participation in collaboration with national and international industries and research institutions and of its actitivities within education and training. Furthermore, the main figures outlining the funding and expenditures of the Department are given. Lists of staff members, visiting scientists, publications and other Department activities are included. (au)

  19. Materials Research Department annual report 1999

    International Nuclear Information System (INIS)

    Soerensen, B.F.; Hansen, N.

    2000-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 1999 are described. The scientific work is presented in three chapters: Materials Science, Materials Engineering and Materials Technology. A survey is given of the Department's participation in collaboration with national and international industries and research institutions and of its actitivities within education and training. Furthermore, the main figures outlining the funding and expenditures of the Department are given. Lists of staff members, visiting scientists, publications and other Department activities are included. (au)

  20. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  1. Construction Progress and Science Planning for the New Research Vessel R/V Sikuliaq

    Science.gov (United States)

    Whitledge, T. E.

    2011-12-01

    The research vessel R/V Sikuliaq (pronounced [see-KOO-lee-auk]) is currently being constructed on behalf of the NSF to support future scientific studies in high latitude waters. The 261 foot global class vessel will be capable of breaking 2.5 foot thick ice at 2 knots with an endurance of 45 days at sea and cruising at 11 knots. The R/V Sikuliaq will have a beam of 52 feet and a draft of 18.9 feet that will carry 26 scientists and a crew of 20. Berthing accommodations are a combination of single/double rooms with one stateroom and the common areas of the vessel are designed for ADA access and accommodations. The total laboratory space (main, analytical, electronics, wet, upper, and Baltic room will be 2100 square feet. The 4360 square foot working deck that is approximately 70 feet in length will accommodate 2-4 vans and multiple science operations. The vessel design strives to have the lowest possible environmental impact, including a low underwater-radiated noise signature. The science systems are prescribed to be state-of-the-art for bottom mapping, over-the-side "hands free" gear handling, broad band communications and scientific walk-in freezer and environmental chamber. More details and photos of the construction progress are available on the website at www.sfos.uaf.edu/arrv. The tentative shipyard schedule has a launch date of June 2012 and delivery to the University of Alaska Fairbanks in June 2013. Scientific operations following trials and testing is planned to start in January 2014. A Sikuliaq science planning workshop has been arranged for 18-19 February 2012 in Salt Lake City, UT just prior to the 2012 Ocean Sciences meeting. Interested participants should contact Terry Whitledge (terry@ims.uaf.edu).

  2. Research projects on life management: materials ageing

    International Nuclear Information System (INIS)

    Gomez Briceno, D.

    1997-01-01

    Materials ageing is a time-dependent process, that involves the loss of availability of nuclear plants. Radiation embrittlement, stress corrosion cracking, irradiation assisted stress corrosion cracking, and thermal ageing are the most relevant time-dependent material degradation mechanisms that can be identified in the materials ageing process. The Materials Programme of Nuclear Energy Institute at CIEMAT carries out research projects and metallurgical examinations of failed components to gain some insight into the mechanisms of materials degradation with a direct impact on the life management of nuclear plants. (Author)

  3. Measurement and flow visualization research of thermal hydraulic characteristics for the SFR reactor Vessel

    International Nuclear Information System (INIS)

    Cha, J. E.; Kim, S. O.; Choi, H. L.; Kim, H. B.; Kim, H. W.; Lee, S. H.

    2012-01-01

    In this report, the thermal hydraulic and flow visualization experiment was described for the KALIMER-600 water-scaled model. In order to investigate a thermal hydraulic characteristics for the SFR KALIMER-600, which has been conceptually designed in the KAERI, a water-scaled 1/10 reactor vessel model was designed and prepared through the scaling analysis during three-years research. In this research, SFR Photos system, which has inherently very complicated the internal structures, was fabricated with a transparent vessel. It was shown that a serious of thermal hydraulic test was conducted within a short period if modeled with water than sodium. Natural circulation test was successfully performed with the modeled heater assembly and heat exchanger system coupled with cooling system. The water-scaled RSV experimental facility made in this research could be used to study the USA development for the future SFR system and utilized to analyze the flow characteristics before changing a main internal part of Photos system. It could also be used to test a pool-inspection study and a sensor selection study before large scale sodium experiment. The PCV system prepared in this research could be utilized to test other TSH experiment and temperature field measurement

  4. Data Collected in 1959 by English Research Vessels at Serial and Surface Hydrographic Stations (NODC Accession 6900852)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The present volume contains data collected in 1959 by English research vessels at serial and surface hydrographic stations. The data list are preceded by a number of...

  5. CTD Data from Research Vessel New Horizon in the NE Pacific, 24 April - 01 May 2014 (NCEI Accession 0157699)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Two consecutive expeditions by research vessel New Horizon in April/May 2014 (NH1408 and NH1409) had the objective to recover and re-deploy a number of moored...

  6. CTD Data from Research Vessel New Horizon in the NE Pacific, 15-20 December 2009 (NCEI Accession 0156689)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The expedition by research vessel New Horizon from 15 to 20 December 2009 had the objective to recover and re-deploy a number of moored platforms off southern...

  7. Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

    2007-03-21

    In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

  8. The instrumentation of the prestressed concrete vessel with hot liner at Seibersdorf Research Centre

    International Nuclear Information System (INIS)

    Zemann, H.

    1975-11-01

    The joint project ''Prestressed Concrete Pressure Vessel with Hot Liner'' at Seibersdorf Research Centre now is in the process of testing the PCPV both in construction and operation from the safety point of view. The physical state of the PCPV (modulus of elasticity, humidity of concrete, creeping, etc.) is brought to stable conditions by ''pre-aging''. In order to control this process of stabilisation, an extensive knowledge of the concrete and an elaborated instrumentation is a necessity. This paper presents a survey about the philosophy and the realisation of the instrumentation of the PCPV and the investigations we performed to interpret the measurements. (author)

  9. MSRR Rack Materials Science Research Rack

    Science.gov (United States)

    Reagan, Shawn

    2017-01-01

    The Materials Science Research Rack (MSRR) is a research facility developed under a cooperative research agreement between NASA and the European Space Agency (ESA) for materials science investigations on the International Space Station (ISS). The MSRR is managed at the Marshall Space Flight Center (MSFC) in Huntsville, AL. The MSRR facility subsystems were manufactured by Teledyne Brown Engineering (TBE) and integrated with the ESA/EADS-Astrium developed Materials Science Laboratory (MSL) at the MSFC Space Station Integration and Test Facility (SSITF) as part of the Systems Development Operations Support (SDOS) contract. MSRR was launched on STS-128 in August 2009, and is currently installed in the U. S. Destiny Laboratory Module on the ISS. Materials science is an integral part of developing new, safer, stronger, more durable materials for use throughout everyday life. The goal of studying materials processing in space is to develop a better understanding of the chemical and physical mechanisms involved, and how they differ in the microgravity environment of space. To that end, the MSRR accommodates advanced investigations in the microgravity environment of the ISS for basic materials science research in areas such as solidification of metals and alloys. MSRR allows for the study of a variety of materials including metals, ceramics, semiconductor crystals, and glasses. Materials science research benefits from the microgravity environment of space, where the researcher can better isolate chemical and thermal properties of materials from the effects of gravity. With this knowledge, reliable predictions can be made about the conditions required on Earth to achieve improved materials. MSRR is a highly automated facility with a modular design capable of supporting multiple types of investigations. Currently the NASA-provided Rack Support Subsystem provides services (power, thermal control, vacuum access, and command and data handling) to the ESA developed Materials

  10. Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

    Energy Technology Data Exchange (ETDEWEB)

    Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A. [Royal Inst. of Tech., Stockholm (Sweden). Div. of Nuclear Power Safety; Helle, M.; Kymaelaeinen, O.; Tuomisto, H. [IVO Power Engineering Ltd., Vantaa (Finland); Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A. [CEA - Grenoble (France); Turland, B.D.; Dobson, G.P. [AEA Technology plc, Dorchester (United Kingdom); Siccama, A. [ECN Nuclear Research, Petten (Netherlands); Ikonen, K. [VTT Energy, Helsinki (Finland); Parozzi, F. [ENEL - SRI/PAM/GRA, Segrate, MI (Italy); Kolev, N. [Siemens AG, Erlangen (Germany); Caira, M. [Univ. of Roma (Italy)

    1999-04-01

    The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues.

  11. Final report for the 'Melt-Vessel Interactions' Project. European Union R and TD Program 4th Framework. MVI project final research report

    International Nuclear Information System (INIS)

    Sehgal, B.R.; Dinh, T.N.; Nourgaliev, R.R.; Bui, V.A.; Green, J.; Kolb, G.; Karbojian, A.; Theerthan, S.A.; Gubaidulline, A.; Bonnet, J.M.; Rouge, S.; Narcoux, M.; Liegeois, A.; Turland, B.D.; Dobson, G.P.; Siccama, A.; Ikonen, K.; Parozzi, F.; Kolev, N.; Caira, M.

    1999-04-01

    The Melt Vessel Interaction (MVI) project is concerned with the consequences of the interactions that a core melt, generated during a postulated severe accident in a light water reactor, may have with the pressure vessel. In particular, the issues concerned with the failure of the vessel bottom head are the focus of the research. The specific objectives of the project are to obtain data and develop validated models, which could be applied to prototypic plants, and accident conditions, for resolution of issues related to the melt vessel interactions. The project work has been performed by nine partners having varied responsibility. The work included a large number of experiments, with simulant materials, whose observations and results are employed, respectively, to understand the physical mechanisms and to develop validated models. Applications to the prototypic geometry and conditions have also been performed. This report is volume 1 of the Final Report for the Project, in which a summary of the progress achieved in the experimental program is provided. We have, however, included some aspects of the modeling activities. Volume 2 of the Final report describes the progress achieved in the modeling program. The progress achieved in the experimental and modeling parts of the Project has led to the resolution of some of the issues of melt vessel interaction. Considerable progress was also achieved towards resolution of the remaining issues

  12. The materials processing research base of the Materials Processing Center

    Science.gov (United States)

    Latanision, R. M.

    1986-01-01

    An annual report of the research activities of the Materials Processing Center of the Massachusetts Institute of Technology is given. Research on dielectrophoresis in the microgravity environment, phase separation kinetics in immiscible liquids, transport properties of droplet clusters in gravity-free fields, probes and monitors for the study of solidification of molten semiconductors, fluid mechanics and mass transfer in melt crystal growth, and heat flow control and segregation in directional solidification are discussed.

  13. Overview of materials research for fusion reactors

    International Nuclear Information System (INIS)

    Muroga, T.; Gasparotto, M.; Zinkle, S.J.

    2002-01-01

    Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale

  14. The future research of material science

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Hironobu [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan)

    1997-11-01

    High Energy Accelerator Research Organization (KEK), which was established on 1 April, consists of two institutes. One of these is Institute of Materials Structure Science. New research program in the new institute using synchrotron radiation, neutrons and muons are discussed. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-09-01

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

  16. Research on release rate of volatile organic compounds in typical vessel cabin

    Directory of Open Access Journals (Sweden)

    ZHANG Jinlan

    2018-02-01

    Full Text Available [Objectives] Volatile Organic Compounds (VOC should be efficiently controlled in vessel cabins to ensure the crew's health and navigation safety. As an important parameter, research on release rate of VOCs in cabins is required. [Methods] This paper develops a method to investigate this parameter of a ship's cabin based on methods used in other closed indoor environments. A typical vessel cabin is sampled with Tenax TA tubes and analyzed by Automated Thermal Desorption-Gas Chromatography-Mass Spectrometry (ATD-GC/MS. The lumped mode is used and the release rate of Benzene, Toluene, Ethylbenzene and Xylene (BTEX, the typical representatives of VOCs, is obtained both in closed and ventilated conditions. [Results] The results show that the content of xylene and Total Volatile Organic Compounds (TVOC exceed the indoor environment standards in ventilated conditions. The BTEX release rate is similar in both conditions except for the benzene. [Conclusions] This research builds a method to measure the release rate of VOCs, providing references for pollution character evaluation and ventilation and purification system design.

  17. Fracture-mechanics data deduced from thermal-shock and related experiments with LWR pressure-vessel material

    International Nuclear Information System (INIS)

    Cheverton, R.D.; Canonico, D.A.; Iskander, S.K.; Bolt, S.E.; Holz, P.P.; Nanstad, R.K.; Stelzman, W.J.

    1982-01-01

    Pressurized water reactors (PWRs) are susceptible to certain types of hypothetical accidents that can subject the reactor pressure vessel to severe thermal shock, that is, a rapid cooling of the inner surface of the vessel wall. The thermal-shock loading, coupled with the radiation-induced reduction in the material fracture toughness, introduces the possibility of propagation of preexistent flaws and what at one time were regarded as somewhat unique fracture-oriented conditions. Several postulated reactor accidents have been analyzed to discover flaw behavior trends; seven intermediate-scale thermal-shock experiments with steel cylinders have been conducted; and corresponding materials characterization studies have been performed. Flaw behavior trends and related fracture-mechanics data deduced from these studies are discussed

  18. Annual report of contract research for the Metallurgy and Materials Research Branch, Division of Reactor Safety Research, Fiscal Year 1977

    International Nuclear Information System (INIS)

    1978-05-01

    Research is reported in the areas of: fracture and structural mechanics; non-destructive testing; steam generator integrity and corrosion; pressure vessel surveillance dosimetry and piping system analysis

  19. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    Dierckx, Marc; Van-Dyck, Dries

    2013-06-01

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  20. Proceedings of the U.S. Nuclear Regulatory Commission on the fifteenth water reactor safety information meeting. V. 2. Materials engineering/pressure vessel research; materials engineering/radiation and degraded piping effects; non-destructive evaluation; environmental effects in primary systems

    International Nuclear Information System (INIS)

    Weiss, Allen J.

    1988-02-01

    This six-volume report contains 140 papers out of the 164 that were presented at the fifteenth water reactor safety information meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included twenty-two different papers presented by researchers from Belgium, Czechoslovakia, Germany, Italy, Japan, Russia, Spain, Sweden, The Netherlands and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. (author)

  1. Reactor core materials research and integrated material database establishment

    International Nuclear Information System (INIS)

    Ryu, Woo Seog; Jang, J. S.; Kim, D. W.

    2002-03-01

    Mainly two research areas were covered in this project. One is to establish the integrated database of nuclear materials, and the other is to study the behavior of reactor core materials, which are usually under the most severe condition in the operating plants. During the stage I of the project (for three years since 1999) in- and out of reactor properties of stainless steel, the major structural material for the core structures of PWR (Pressurized Water Reactor), were evaluated and specification of nuclear grade material was established. And the damaged core components from domestic power plants, e.g. orifice of CVCS, support pin of CRGT, etc. were investigated and the causes were revealed. To acquire more resistant materials to the nuclear environments, development of the alternative alloys was also conducted. For the integrated DB establishment, a task force team was set up including director of nuclear materials technology team, and projector leaders and relevant members from each project. The DB is now opened in public through the Internet

  2. New developments in photon and materials research

    CERN Document Server

    2013-01-01

    This book presents the most recent updates in the field of photon and optical materials research. It is devoted to various interdisciplinary subjects such as fundamental photon physics, bio and medical photon physics, ultrafast non-linear optics, quasiparticle excitation and spectroscopy, coherent mid-infrared (IR) light sources, functional optoelectronic materials and optical fibres, and quantum nano-structured devices for various important technological applications. It contains 19 authoritative peer-reviewed chapters regarding experimental and theoretical research in these fields, contributed by young scientists and engineers (assistant or associate professor level) along with well-established experts. The response of materials to electromagnetic fields, namely light-matter interaction, has been of special concern in fundamental optical sciences. The ability to fabricate and/or engineer new materials and structures is giving rise to revolutionary changes in the field, which also includes soft condensed mat...

  3. Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2002-01-01

    1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf (41-J) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material database that was available as of May 2000. The embrittlement correlation used in this guide was developed using the following variables: copper and nickel contents, irradiation temperature, and neutron fluence. The form of the model was based on current understanding for two mechanisms of embrittlement: stable matrix damage (SMD) and copper-rich precipitation (CRP); saturation of copper effects (for different weld materials) was included. This guide is applicable for the following specific materials, copper, nickel, and phosphorus contents, range of irradiation temperature, and neutron fluence based on the overall database: 1.1.1 MaterialsA 533 Type B Class 1 and 2, A302 Grade B, A302 G...

  4. Neutron monitor measurements on the German research vessel Polarstern. First results

    Energy Technology Data Exchange (ETDEWEB)

    Heber, B. [Insititut fuer Experimentelle und Angewandte Physik, Christian-Albrechts-Universitaet zu Kiel (Germany); Schwerdt, C.; Walter, M. [Deutsches Elektronen-Synchrotron DESY, D-15738 Zeuthen (Germany); Bernade, G.; Fuchs, R.; Krueger, H.; Moraal, H. [Center for Space Research, North-West University, Potchefstroom 2520 (South Africa)

    2014-07-01

    Cosmic-ray particles provide a unique opportunity to probe the dynamic conditions in the highly variable heliosphere. The longest continuous measurements of galactic cosmic rays come from cosmogenic isotopes and from neutron monitors located at different location on Earth. Understanding the effects of energetic particles in and on the atmosphere and the environment of Earth must address their transport to Earth and their interactions with the Earth's atmosphere, including their filtering by the terrestrial magnetosphere. Since neutron monitors are integral detectors of secondary cosmic rays produced in the atmosphere, a single neutron monitor can only derive the energy spectra of the particles impinging on the Earth during latitudinal surveys. A portable neutron monitor was built at the North-West University, South Africa, and was installed on the German research vessel Polarstern. Such latitude surveys have been done before, but this vessel is better suited for this purpose than previous platforms because it traverses all the locations with geomagnetic cutoff rigidities from <<1 GV to 15 GV at least twice per year. In this contribution we present first results from the measurement campaigns.

  5. Improvement of shipborne sky radiometer and its demonstration aboard the Antarctic research vessel Shirase

    Directory of Open Access Journals (Sweden)

    Noriaki Tanaka

    2014-11-01

    Full Text Available The sun-tracking performance of a shipborne sky radiometer was improved to attain accurate aerosol optical thickness (AOT from direct solar measurements on a pitching and rolling vessel. Improvements were made in the accuracy of sun-pointing measurements, field-of-view expansion, sun-tracking speed, and measurement method. Radiometric measurements of direct solar and sky brightness distribution were performed using the shipborne sky radiometer onboard the Antarctic research vessel (R/V Shirase during JARE-51 (2009-2010 and JARE-52 (2010-2011. The temporal variation of signal intensity measured by the radiometer under cloudless conditions was smooth, demonstrating that the radiometer could measure direct sunlight onboard the R/V. AOT at 500 nm ranged from 0.01 to 0.34, and values over Southeast Asia and over the western Pacific Ocean in spring were higher than those over other regions. The Angstrom exponent ranged from -0.06 to 2.00, and values over Southeast Asia and off the coast near Sydney were the highest. The improved shipborne sky radiometer will contribute to a good understanding of the nature of aerosols over the ocean.

  6. Advances in Functionalized Materials Research 2016

    International Nuclear Information System (INIS)

    Predoi, D.; Motelica-Heino, M.; Guegan, R.; Coustumer, L.Ph.

    2016-01-01

    In the last years, due to the rapid progress of technology, new materials at nano metric scale with special properties have become a flourishing field of research in materials science. The unique physicochemical properties of materials induced by various parameters such as mean size, shape, purity, crystallographic structure, and surface can generate effective solutions to challenging environmental and biomedical problems. As a result of this approach a large number of techniques were developed that enable obtaining novel materials at nano metric scale with specific and reproducible properties and parameters. Below will be highlighted studies on promising properties on the applicability of new materials that could lead to innovative applications in the medical field. Therefore, this special issue is focused on expected advances in the area of functionalized materials at nano metric scale. Due to multidisciplinarity of this topic, this special issue is comprised of a wide range of original research articles as well as review papers on the design and synthesis of functionalized nano materials, their structural, morphological, and biological characterization, and their potential uses in medical and environmental applications

  7. Research of footwear lining materials thermoconductive properties

    Science.gov (United States)

    Maksudova, U.; Ilkhamova, M.; Mirzayev, N.; Pazilova, D.

    2017-11-01

    Protective properties of footwear are influenced by a number of factors and the most important of them are: design features of the top and the bottom of the footwear, it’s shape, physical and mechanical properties of the components of which they are made. In course of work there were researched thermoconductive properties of different lining membrane materials used for production of high temperature protective footwear. Research results allow to select the appropriate materials by reference to thermoconductive properties during design of protective footwear for extreme conditions to prolong the wearer’s time of comfortable stay in conditions of exposure of elevated temperatures to a stack.

  8. Development of neutron irradiation embrittlement correlation of reactor pressure vessel materials of light water reactors

    International Nuclear Information System (INIS)

    Soneda, Naoki; Dohi, Kenji; Nomoto, Akiyoshi; Nishida, Kenji; Ishino, Shiori

    2007-01-01

    A large amount of surveillance data of the RPV embrittlement of the Japanese light water reactors have been compiled since the current Japanese embrittlement correlation has been issued in 1991. Understanding on the mechanisms of the embrittlement has also been greatly improved based on both experimental and theoretical studies. CRIEPI and the Japanese electric power utilities have started research project to develop a new embrittlement correlation method, where extensive study of the microstructural analyses of the surveillance specimens irradiated in the Japanese commercial reactors has been conducted. The new findings obtained from the experimental study are that the formation of solute-atom clusters with little or no copper is responsible for the embrittlement in low-copper materials, and that the flux effect exists especially in high-copper materials and this is supported by the difference in the microstructure of the high-copper materials irradiated at different fluxes. Based on these new findings, a new embrittlement correlation method is formulated using rate equations. The new methods has higher prediction capability than the current Japanese embrittlement correlation in terms of smaller standard deviation as well as smaller mean value of the prediction error. (author)

  9. Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

    Energy Technology Data Exchange (ETDEWEB)

    Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E. [Oak Ridge National Lab., Heavy-Section Steel Technology Program, Oak Ridge, TN (United States)

    2001-07-01

    A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

  10. Development of a Weibull model of cleavage fracture toughness for shallow flaws in reactor pressure vessel material

    International Nuclear Information System (INIS)

    Bass, B.R.; Williams, P.T.; McAfee, W.J.; Pugh, C.E.

    2001-01-01

    A primary objective of the United States Nuclear Regulatory Commission (USNRC) -sponsored Heavy-Section Steel Technology (HSST) Program is to develop and validate technology applicable to quantitative assessments of fracture prevention margins in nuclear reactor pressure vessels (RPVs) containing flaws and subjected to service-induced material toughness degradation. This paper describes an experimental/analytical program for the development of a Weibull statistical model of cleavage fracture toughness for applications to shallow surface-breaking and embedded flaws in RPV materials subjected to multi-axial loading conditions. The experimental part includes both material characterization testing and larger fracture toughness experiments conducted using a special-purpose cruciform beam specimen developed by Oak Ridge National Laboratory for applying biaxial loads to shallow cracks. Test materials (pressure vessel steels) included plate product forms (conforming to ASTM A533 Grade B Class 1 specifications) and shell segments procured from a pressurized-water reactor vessel intended for a nuclear power plant. Results from tests performed on cruciform specimens demonstrated that biaxial loading can have a pronounced effect on shallow-flaw fracture toughness in the lower-transition temperature region. A local approach methodology based on a three-parameter Weibull model was developed to correlate these experimentally-observed biaxial effects on fracture toughness. The Weibull model, combined with a new hydrostatic stress criterion in place of the more commonly used maximum principal stress in the kernel of the Weibull stress integral definition, is shown to provide a scaling mechanism between uniaxial and biaxial loading states for 2-dimensional flaws located in the A533-B plate material. The Weibull stress density was introduced as a matrice for identifying regions along a semi-elliptical flaw front that have a higher probability of cleavage initiation. Cumulative

  11. Advanced research workshop: nuclear materials safety

    International Nuclear Information System (INIS)

    Jardine, L J; Moshkov, M M.

    1999-01-01

    The Advanced Research Workshop (ARW) on Nuclear Materials Safety held June 8-10, 1998, in St. Petersburg, Russia, was attended by 27 Russian experts from 14 different Russian organizations, seven European experts from six different organizations, and 14 U.S. experts from seven different organizations. The ARW was conducted at the State Education Center (SEC), a former Minatom nuclear training center in St. Petersburg. Thirty-three technical presentations were made using simultaneous translations. These presentations are reprinted in this volume as a formal ARW Proceedings in the NATO Science Series. The representative technical papers contained here cover nuclear material safety topics on the storage and disposition of excess plutonium and high enriched uranium (HEU) fissile materials, including vitrification, mixed oxide (MOX) fuel fabrication, plutonium ceramics, reprocessing, geologic disposal, transportation, and Russian regulatory processes. This ARW completed discussions by experts of the nuclear materials safety topics that were not covered in the previous, companion ARW on Nuclear Materials Safety held in Amarillo, Texas, in March 1997. These two workshops, when viewed together as a set, have addressed most nuclear material aspects of the storage and disposition operations required for excess HEU and plutonium. As a result, specific experts in nuclear materials safety have been identified, know each other from their participation in t he two ARW interactions, and have developed a partial consensus and dialogue on the most urgent nuclear materials safety topics to be addressed in a formal bilateral program on t he subject. A strong basis now exists for maintaining and developing a continuing dialogue between Russian, European, and U.S. experts in nuclear materials safety that will improve the safety of future nuclear materials operations in all the countries involved because of t he positive synergistic effects of focusing these diverse backgrounds of

  12. Contributions to radiochemical and nuclear materials research

    International Nuclear Information System (INIS)

    Matzke, H.

    1982-01-01

    Series of talks given during a seminar of the European Institute for Transuranium Elements in april 1981 in honor of R. LINDNER on the occasion of his 60th birth day. The topics include general aspects of research practice and science prognosis, retrospective essays about the discovery of nuclear fission by O. HAHN as well as surveys of actual research activities concerning a radiochemistry and the use of radioactivity in material science

  13. Nuclear physics methods in materials research

    International Nuclear Information System (INIS)

    1980-01-01

    The brochure contains the abstracts of the papers presented at the 7th EPS meeting 1980 in Darmstadt. The main subjects were: a) Neutron scattering and Moessbauer effect in materials research, b) ion implantation in micrometallurgy, c) applications of nuclear reactions and radioisotopes in research on solids, d) recent developments in activation analysis and e) pions, positrons, and heavy ions applied in solid state physics. (RW) [de

  14. Research on removal technologies of fuel debris and in-vessel structures using laser light (1). Research plan and research activities on FY2012

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Yamada, Tomonori; Hanari, Toshihide; Takebe, Toshihiko; Matsunaga, Yukihiro

    2013-08-01

    In decommissioning works of the Fukushima Daiichi nuclear power plants, it is required that fuel debris solidifying mixed materials of fuels and in-vessel structures should be removed. The fuel debris is considered to have characteristics, such as indefinite shapes, porous bodies, multi-compositions, higher hardness, etc. from the knowledge in decommissioning process of the Three Mile Island nuclear power plant. Laser lights are characterized by higher power density, local processability, remote controllability, etc. and can be performed thermal cutting and crushing-up for various materials which does not depend on fracture toughness. This report describes a research program and research activities in FY2012 aiming at developing removal system of fuel debris by the use of laser lights. Main results obtained from research activities in FY2012 are as follows: (1) Improvements of experimental infrastructures. A beam switching unit for an existing fiber laser system, an x-y-z tri-axes robot system to investigate remote control performances, and a particle image velocimetry (PIV) system for quantitation of assist gas flow characteristics were introduced to the experimental laboratory of our Applied Laser Technology Institute in Tsuruga. (2) Laser cutting performances for thick metal plates. To quantify laser cutting performance for thick metal plates of in-vessel structures, after the evaluation of the relationship between the kerf depth and amount of laser irradiation energy to the metal test piece, we evaluated for heat transfer behavior due to temperature measurement of thick metal plate on the laser cutting process. It is suggested that the heat diffusion into the cutting object can affect the heat input efficiency of the laser irradiation energy to kerf front. On the viewpoint of suppressing this thermal diffusion, it was found that it is important in improving the laser cutting performance to increase the ejection of molten metal by the assist gas, and to optimize

  15. Evaluation of the Structural Safety of a Vessel with Different Material(Cr-13)-Supplemented Screw Thread

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Hoon; Bae, Jun Ho; Kim, Chul [Pusan National University, Busan (Korea, Republic of)

    2015-04-15

    The dome and neck part of a vessel is generally formed by a hot spinning process with a seamless tube. However, as studies on and design data from the hot spinning process are insufficient, this process has been performed based on trial and error and the experiences of field engineers. Changes in the inner diameter from the bottom to the top of the neck have occurred mainly because of the characteristics of the hot spinning process due to the high-speed rotation of the rollers. In this study, a theoretical and finite element analysis of the vessel is conducted with different material(Cr-13)-supplemented screw threads for tapping and to reduce shape errors. Based on the results, the structural safety under the operating conditions is evaluated.

  16. Resolution of the reactor vessel materials toughness safety issue; Task Action Plan A-11; Appendices C-K

    International Nuclear Information System (INIS)

    Johnson, R.E.

    1981-09-01

    The central problem in the Unresolved Safety Issue A-11, 'Reactor Vessel Materials Toughness,' was to provide guidance in performing analyses for reactor pressure vessels (RPVs) which fail to meet the toughness requirements during service life as a result of neutron radiation embrittlement. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which has been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the RPV fracture problem with assumed beltline region flaws. Volume I of this report is a brief presentation of the problem and the results; Volume II provides the detailed technical foundations

  17. A Dataset of Deep-Sea Fishes Surveyed by Research Vessels in the Waters around Taiwan

    Directory of Open Access Journals (Sweden)

    Kwang-Tsao Shao

    2014-12-01

    Full Text Available The study of deep-sea fish fauna is hampered by a lack of data due to the difficulty and high cost incurred in its surveys and collections. Taiwan is situated along the edge of the Eurasia fig, at the junction of three Large Marine Ecosystems or Ecoregions of the East China Sea, South China Sea and the Philippines. As nearly two-thirds of its surrounding marine ecosystems are deep-sea environments, Taiwan is expected to hold a rich diversity of deep-sea fish. However, in the past, no research vessels were employed to collect fish data on site. Only specimens, caught by bottom trawl fishing in the waters hundreds of meters deep and missing precise locality information, were collected from Dasi and Donggang fishing harbors. Began in 2001, with the support of National Science Council, research vessels were made available to take on the task of systematically collecting deep-sea fish specimens and occurrence records in the waters surrounding Taiwan. By the end of 2006, a total of 3,653 specimens, belonging to 26 orders, 88 families, 198 genera and 366 species, were collected in addition to data such as sampling site geographical coordinates and water depth, and fish body length and weight. The information, all accessible from the “Database of Taiwan’s Deep-Sea Fauna and Its Distribution (http://deepsea.biodiv.tw/” as part of the “Fish Database of Taiwan,” can benefit the study of temporal and spatial changes in distribution and abundance of fish fauna in the context of global deep-sea biodiversity.

  18. A miniature research vessel: A small-scale ocean-exploration demonstration of geophysical methods

    Science.gov (United States)

    Howell, S. M.; Boston, B.; Sleeper, J. D.; Cameron, M. E.; Togia, H.; Anderson, A.; Sigurdardottir, T. D.; Tree, J. P.

    2015-12-01

    Graduate student members of the University of Hawaii Geophysical Society have designed a small-scale model research vessel (R/V) that uses sonar to create 3D maps of a model seafloor in real-time. A pilot project was presented to the public at the School of Ocean and Earth Science and Technology's (SOEST) Biennial Open House weekend in 2013 and, with financial support from the Society of Exploration Geophysicists and National Science Foundation, was developed into a full exhibit for the same event in 2015. Nearly 8,000 people attended the two-day event, including children and teachers from Hawaii's schools, home school students, community groups, families, and science enthusiasts. Our exhibit demonstrates real-time sonar mapping of a cardboard volcano using a toy size research vessel on a programmable 2-dimensional model ship track suspended above a model seafloor. Ship waypoints were wirelessly sent from a Windows Surface tablet to a large-touchscreen PC that controlled the exhibit. Sound wave travel times were recorded using an ultrasonic emitter/receiver attached to an Arduino microcontroller platform and streamed through a USB connection to the control PC running MatLab, where a 3D model was updated as the ship collected data. Our exhibit demonstrates the practical use of complicated concepts, like wave physics, survey design, and data processing in a way that the youngest elementary students are able to understand. It provides an accessible avenue to learn about sonar mapping, and could easily be adapted to talk about bat and marine mammal echolocation by replacing the model ship and volcano. The exhibit received an overwhelmingly positive response from attendees and incited discussions that covered a broad range of earth science topics.

  19. Final report for the 5th surveillance test of the reactor pressure vessel material (capsule Y) of Yonggwang Nuclear Power Plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sam Lai; Kim, ByoungChul; Chang, Kee Ok (and others)

    2006-02-15

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 5th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Daejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X, W and Y are 5.777E+18, 1.5371E+19, 3.7634E+19, 4.3045E+19, and 4.8662E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.953 for the 1st through 5th testing and the calculational uncertainty,7.2% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.659E+19n/cm{sup 2} based on the end of 13th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 48, 56 and 64EFPY would reach 3.625E+19, 5.293E+19, 6.127E+19 and 6.960E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the Pressurized Thermal Shock(PTS) during the operation until design life.

  20. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Younggwang nuclear power plant unit1

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-08-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Yonggwang site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 1 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.555E+18, 1.662E+19, 3.358E+19, and 4.521E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.859 for the 1st through 4th testing and the calculational uncertainty, 11.80% satisfied the requirement of USNRC Reg.Guide 1.190, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.551E+19 n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.929E+19, 4.880E+19, 5.831E+19 and 6.782E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 1 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 4 refs., 41 figs., 35 tabs. (Author)

  1. The 4th surveillance test and evaluation of the reactor pressure vessel material (capsule W) of Yonggwang nuclear power plant unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwon Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-02-01

    Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Yonggwang unit 2 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 5.762E+18, 1.5391E+19, 3.5119E+19, and 4.2610E+19 n/cm{sup 2}, respectively. The bias factor, the ratio of measurement versus calculation, was 0.899 for the 1st through 4th testing and the calculational uncertainty, 12.3% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.357E+19 n/cm{sup 2} based on the end of 11th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.525E+19, 4.337E+19, 5.148E+19 and 5.960E+19 n/cm{sup 2} based on the current calculation. The result through this analysis for Yonggwang unit 2 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 42 tabs. (Author)

  2. NASA Materials Research for Extreme Conditions

    Science.gov (United States)

    Sharpe, R. J.; Wright, M. D.

    2009-01-01

    This Technical Memorandum briefly covers various innovations in materials science and development throughout the course of the American Space program. It details each innovation s discovery and development, explains its significance, and describes the applications of this material either in the time period discovered or today. Topics of research include silazane polymers, solvent-resistant elastomeric polymers (polyurethanes and polyisocyanurates), siloxanes, the Space Shuttle thermal protection system, phenolic-impregnated carbon ablator, and carbon nanotubes. Significance of these developments includes the Space Shuttle, Apollo programs, and the Constellation program.

  3. Researches of smart materials in Japan

    International Nuclear Information System (INIS)

    Furuya, Y.; Tani, J.

    2000-01-01

    The choice of sensor and actuator material as well as optimum design to combine the actuator element with the host structure become very essential to develop a smart materials and structures. In the present paper, first, the present state and issues of the main solid actuators are described from the viewpoint of material science and engineering. Next, the developments of smart materials and systems using shape memory materials in Japan are introduced. Shape memory TiNi fiber reinforced/Al or polymer matrix composites have been fabricated to confirm the enhancements of fracture toughness (K-value) by utilizing the compression stresses caused by shape memory shrinkage of embedded TiNi fibers. Sudden failure prevention system for structures are also proposed by combining non-destructive acoustic emission detecting system with suppression of crack-tip stress intensity by shape memory shrinkage effect. Lastly, the research project scheme and several targets on smart actuator development are introduced, which are imposed on the Tohoku University team in the Japanese National Project (1998∝2002 A.D.) on smart materials and structure system by NEDO/MITI. (orig.)

  4. Advanced Research Projects Agency on Materials Preparation and Characterization Research

    Science.gov (United States)

    Briefly summarized is research concerned with such topics as: Preparation of silica glass from amorphous silica; Glass structure by Raman ...ferroelectrics; Silver iodide crystals; Vapor phase growth; Refractory optical host materials; Hydroxyapatite ; Calcite; Characterization of single crystals with a double crystal spectrometer; Characterization of residual strain.

  5. Resolution of the Task A-11 reactor-vessel materials-toughness safety issue. Appendices C-K

    International Nuclear Information System (INIS)

    1982-10-01

    The central problem in the unresolved safety issue A-11, Reactor Vessel Materials Toughness, was to provide guidance in performing analyses required by 10 CFR Part 50, Appendix G, Section V.C. for reactor pressure vessels (RPVs) which fail to meet the toughness requirement during service life as a result of neutron radiation embrittlement. Although the methods of linear-elastic fracture mechanics (LEFM) were adequate for low-temperature RPV problems, they were inapplicable under operating conditions because vessel steels, even those which exhibit less than 50 ft-lb of C/sub v/ energy, were relatively tough at temperatures where the impact energy reached its upper shelf values. A technical team of recognized experts was organized to assist the NRC staff in addressing the problem. Using the foundation of the tearing modulus concept, which had been developed under earlier NRC sponsorship, relationships were obtained which provided approximate solutions to the problem of RPV fracture with assumed beltline region flaws. The first paper of this report is a summary of the problem, the solutions, and the results of verification analyses. The details are provided in a series of appendices in Volumes I and II

  6. New methods of analysis of materials strength data for the ASME Boiler and Pressure Vessel Code

    International Nuclear Information System (INIS)

    Booker, M.K.; Booker, B.L.P.

    1980-01-01

    Tensile and creep data of the type used to establish allowable stress levels for the ASME Boiler and Pressure Vessel Code have been examined for type 321H stainless steel. Both inhomogeneous, unbalanced data sets and well-planned homogeneous data sets have been examined. Data have been analyzed by implementing standard manual techniques on a modern digital computer. In addition, more sophisticated techniques, practical only through the use of the computer, have been applied. The result clearly demonstrates the efficacy of computerized techniques for these types of analyses

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

    International Nuclear Information System (INIS)

    1994-01-01

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

  8. Materials Research Department. Annual Report 2001

    Energy Technology Data Exchange (ETDEWEB)

    Cartensen, J.V.; Lindgaard, P.A.; Freidenhans' I, R. (eds.)

    2002-08-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 2001 are described. The scientific work is described in 10 chapters and a survey is given of the Department's educational activities along with a list of published work. Furthermore, the main figures outlining the funding and expenditures of the Department are given and a list of staff members is included. (au)

  9. Materials Research Department annual report 2003

    International Nuclear Information System (INIS)

    Bentzen, J.J.; Lindgaerd, P.A.; Feidenhans'l, R.

    2004-04-01

    Selected activities of the Materials Research Department at Risoe National Laboratory during 2003 are described. The Scientific work is described in five chapters and a survey is given of the Departments educational activities along with a list of published work, prizes, organized meetings, and membership of committees. Furthermore, the main figures outlining the funding and expenditures of the Department are given and a list of staff members is included. (au)

  10. Heavy-Section Steel Irradiation Program on irradiation effects in light-water reactor pressure vessel materials

    International Nuclear Information System (INIS)

    Nanstad, R.K.; Corwin, W.R.; Alexander, D.J.; Haggag, F.M.; Iskander, S.K.; McCabe, D.E.; Sokolov, M.A.; Stoller, R.E.

    1995-01-01

    The safety of commercial light-water nuclear plants is highly dependent on the structural integrity of the reactor pressure vessel (RPV). In the absence of radiation damage to the RPV, fracture of the vessel is difficult to postulate. Exposure to high energy neutrons can result in embrittlement of radiation-sensitive RPV materials. The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory, sponsored by the US Nuclear Regulatory Commission (USNRC), is assessing the effects of neutron irradiation on RPV material behavior, especially fracture toughness. The results of these and other studies are used by the USNRC in the evaluation of RPV integrity and regulation of overall nuclear plant safety. In assessing the effects of irradiation, prototypic RPV materials are characterized in the unirradiated condition and exposed to radiation under varying conditions. Mechanical property tests are conducted to provide data which can be used in the development of guidelines for structural integrity evaluations, while metallurgical examinations and mechanistic modeling are performed to improve understanding of the mechanisms responsible for embrittlement. The results of these investigations, in conjunction with results from commercial reactor surveillance programs, are used to develop a methodology for the prediction of radiation effects on RPV materials. This irradiation-induced degradation of the materials can be mitigated by thermal annealing, i.e., heating the RPV to a temperature above that of normal operation. Thus, thermal annealing and evaluation of reirradiation behavior are major tasks of the HSSI Program. This paper describes the HSSI Program activities by summarizing some past and recent results, as well as current and planned studies. 30 refs., 8 figs., 1 tab

  11. 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

  12. AECL research programmes in materials science

    International Nuclear Information System (INIS)

    Cox, B.; Eastwood, T.A.; Mitchell, I.V.; Dutton, R.

    1980-10-01

    The high capacity factors achieved by CANDU nuclear power reactors can be attributed in part to the careful attention which has been paid in the concept and design phases to the selection of materials. Improved tolerance of these materials to the hostile conditions of a reactor core depends upon our understanding of such phenomena as radiation damage, corrosion and cracking. This report is an introduction to some of the fundamental and underlying research programmes that have evolved at the AECL laboratories in response to this need. The interactions of energetic atomic particles with solids on a microscopic scale are considered, first under the general heading of radiation effects, followed by sections on energy loss processes, ion channeling, and crystal lattice defects. The latter section leads into the important programmes on deformation processes (creep and growth) in zirconium. The final section discusses the extensive work on the oxidation and environmental cracking of zirconium alloys. (auth)

  13. Nuclear radioactive techniques applied to materials research

    CERN Document Server

    Correia, João Guilherme; Wahl, Ulrich

    2011-01-01

    In this paper we review materials characterization techniques using radioactive isotopes at the ISOLDE/CERN facility. At ISOLDE intense beams of chemically clean radioactive isotopes are provided by selective ion-sources and high-resolution isotope separators, which are coupled on-line with particle accelerators. There, new experiments are performed by an increasing number of materials researchers, which use nuclear spectroscopic techniques such as Mössbauer, Perturbed Angular Correlations (PAC), beta-NMR and Emission Channeling with short-lived isotopes not available elsewhere. Additionally, diffusion studies and traditionally non-radioactive techniques as Deep Level Transient Spectroscopy, Hall effect and Photoluminescence measurements are performed on radioactive doped samples, providing in this way the element signature upon correlation of the time dependence of the signal with the isotope transmutation half-life. Current developments, applications and perspectives of using radioactive ion beams and tech...

  14. QFD-ANP Approach for the Conceptual Design of Research Vessels: A Case Study

    Science.gov (United States)

    Venkata Subbaiah, Kambagowni; Yeshwanth Sai, Koneru; Suresh, Challa

    2016-10-01

    Conceptual design is a subset of concept art wherein a new idea of product is created instead of a visual representation which would directly be used in a final product. The purpose is to understand the needs of conceptual design which are being used in engineering designs and to clarify the current conceptual design practice. Quality function deployment (QFD) is a customer oriented design approach for developing new or improved products and services to enhance customer satisfaction. House of quality (HOQ) has been traditionally used as planning tool of QFD which translates customer requirements (CRs) into design requirements (DRs). Factor analysis is carried out in order to reduce the CR portions of HOQ. The analytical hierarchical process is employed to obtain the priority ratings of CR's which are used in constructing HOQ. This paper mainly discusses about the conceptual design of an oceanographic research vessel using analytical network process (ANP) technique. Finally the QFD-ANP integrated methodology helps to establish the importance ratings of DRs.

  15. NASA Lewis Research Center's materials and structures division

    International Nuclear Information System (INIS)

    Weymueller, C.R.

    1976-01-01

    Research activities at the NASA Lewis Research Center on materials and structures are discussed. Programs are noted on powder metallurgy superalloys, eutectic alloys, dispersion strengthened alloys and composite materials. Discussions are included on materials applications, coatings, fracture mechanics, and fatigue

  16. Material Transfer Agreement (MTA) | Frederick National Laboratory for Cancer Research

    Science.gov (United States)

    Material Transfer Agreements are appropriate for exchange of materials into or out of the Frederick National Laboratory for research or testing purposes, with no collaborative research by parties involving the materials.

  17. Research with radioactive materials in man

    International Nuclear Information System (INIS)

    Roedler, H.D.

    1987-01-01

    In connection with the revision of the Radiation Protection Ordinance, for instance in section 41, the author - who can draw on his own experience as a referee for projects planned in the area of research with radioactive materials in man - deals with the following problems: 1. Quantifiable risk-benefit assessment as opposed to qualitative risk-benefit assessment based on medical experience. 2. Delimination of medicine and research by criteria such as application to healthy or sick persons, application of a new method or an already standardized one, application in the hope to achieve an individual benefit or without such hopes, and application with a view to obtaining results suitable to be generalized, in the course of which many borderline cases will crop up. 3. Legal requirements in section 41 of the Radiation Protection Ordinance with the demands for minimization of risks and quality assurance, and 4. application procedure and experience gathered so far. (TRV) [de

  18. After-operating properties of nuclear reactor vessel materials of Lenin atomic ice breaker and prospective of reactor vessels radiation life prolongation

    International Nuclear Information System (INIS)

    Platonov, P.A.; Shtrombakh, Ya.I.; Amaev, A.D.; Krasikov, E.A.; Korolev, Yu.N.; Zabusov, O.O.; Glushakov, G.M.

    2001-01-01

    A post-operational state of the icebreaker Lenin reactor vessel metal is investigated. It is shown that a base metal of the icebreaker Lenin reactor vessel is of high quality as by an initial value of critical temperature of embrittlement, so by its radiation resistance. The weld metal possesses a sufficient radiation resistance but has an insufficient initial ductile-brittle transition temperature (approximately 63 Deg C). It is necessary to note that the final stage of operation for nuclear steam-generating plant should be carried out at the coolant temperature as high as possible [ru

  19. Assessment of weld heat-affected zones in a reactor vessel material

    International Nuclear Information System (INIS)

    Marston, T.U.; Server, W.

    1978-01-01

    The mechanical properties of weld heat-affected zones (HAZ's) associated with the heavy section, nuclear quality weldments are evaluated and found to be superior to those of the parent base material. The nil ductility transition temperature (NDTT), Charpy impact and static and dynamic fracture toughness properties of a HAZ associated with a submerged arc weld and one associated with a manual metal arc weld are directly compared with those of the parent base material. It is concluded that the stigma normally associated with HAZ is not justified for this grade and quality of material and weld procedure

  20. Characterization of the weld HAZ properties of nuclear reactor pressure vessel materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Joo Hag; Shin, H. S.; Moon, J. G. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    This work contains an investigation on the microstructure and toughness in the weld heat-affected zone (HAZ) of a quenched and tempered SA 508 Cl. 3 reactor pressure vessel (RPV) steel. In order to evaluate systematically the notch toughness and microstructural alterations, a unit HAZ concept was applied to the multipass weld HAZ of RPV steel. Seven typical positions were selected to evaluate the spatial distribution of notch toughness and microstructure in the unit HAZ. As a result of notch toughness evaluation, three coarse-grained regions and two fine-grained regions of SA 508 Cl. 3 RPV steel HAZ showed relatively good toughness. On the contrary, an intercritically reheated and a subcritically reheated region showed lower toughness than the base metal. The region which first and second peak temperatures are 700 deg C showed the lowest toughness among the low toughness region because of carbide coarsening. Therefore, it was proposed that the notch position in the surveillance HAZ specimen should be placed to the boundary between the HAZ and the base metal. The method, which evaluates the fracture toughness in the transition region of ferritic steel, was effectively applicable to the various HAZ regions of RPV steel. The fracture toughness test results were nearly same as the notch toughness test results. The volume fraction of tempered martensite phase was revealed as the most dominant factor that determines fracture toughness. 59 refs., 29 figs., 10 tabs. (Author)

  1. A perspective on thermal annealing of reactor pressure vessel materials from the viewpoint of experimental results

    International Nuclear Information System (INIS)

    Iskander, S.K.; Sokolov, M.A.; Nanstad, R.K.

    1996-01-01

    It is believed that in the next decade or so, several nuclear reactor pressure vessels (RPVs) may exceed the reference temperature limits set by the pressurized thermal shock screening criteria. One of the options to mitigate the effects of irradiation on RPVs is to thermally anneal them to restore the toughness properties that have been degraded by neutron irradiation. This paper summarizes recent experimental results from work performed at the Oak Ridge National Laboratory to study the annealing response, or ''recovery'' of several irradiated RPV steels. The fracture toughness is one of the important properties used in the evaluation of the integrity of RPVs. Optimally, the fracture toughness is measured directly by fracture toughness specimens, such as compact tension or precracked Charpy specimens, but is often inferred from the results of Charpy V-notch impact specimens. The experimental results are compared to the predictions of models for embrittlement recovery which have been developed by Eason et al. Some of the issues in annealing that still need to be resolved are discussed

  2. 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

  3. Using the Steel Vessel Material-Cost Index to Mitigate Shipbuilder Risk

    National Research Council Canada - National Science Library

    Keating, Edward G; Murphy, Robert; Schank, John F; Birkler, John

    2008-01-01

    This paper describes how the US Navy structures fixed-price and fixed-price, incentive-fee shipbuilding contracts and how labor- and material-cost indexes can mitigate shipbuilder risk in either type of contract...

  4. An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

    Energy Technology Data Exchange (ETDEWEB)

    Calderoni, P., E-mail: Pattrick.Calderoni@inl.gov [Fusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113 (United States); Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G. [Fusion Safety Program, Idaho National Laboratory, PO Box 1625, Idaho Falls, ID 83415-7113 (United States); Hatano, Y.; Hara, M. [Hydrogen Isotope Research Center, University of Toyama, Gofuku 3190, Toyama 930-8555 (Japan); Oya, Y. [Radioscience Research Laboratory, Faculty of Science, Shizuoka University, 836 Ohya, Suruga-ku, Shizuoka 422-8529 (Japan); Otsuka, T.; Katayama, K. [Interdisciplinary Graduate School of Engineering Sciences, Kyushu University, 6-10-1 Hakozaki, Higashi-ku, Fukuoka 812-8581 (Japan); Konishi, S.; Noborio, K.; Yamamoto, Y. [Institute of Advanced Energy, Kyoto University, Gokasho, Uji, Kyoto 611-0011 (Japan)

    2011-10-01

    The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

  5. An overview of research activities on materials for nuclear applications at the INL Safety, Tritium and Applied Research facility

    International Nuclear Information System (INIS)

    Calderoni, P.; Sharpe, J.; Shimada, M.; Denny, B.; Pawelko, B.; Schuetz, S.; Longhurst, G.; Hatano, Y.; Hara, M.; Oya, Y.; Otsuka, T.; Katayama, K.; Konishi, S.; Noborio, K.; Yamamoto, Y.

    2011-01-01

    The Safety, Tritium and Applied Research facility at the Idaho National Laboratory is a US Department of Energy National User Facility engaged in various aspects of materials research for nuclear applications related to fusion and advanced fission systems. Research activities are mainly focused on the interaction of tritium with materials, in particular plasma facing components, liquid breeders, high temperature coolants, fuel cladding, cooling and blanket structures and heat exchangers. Other activities include validation and verification experiments in support of the Fusion Safety Program, such as beryllium dust reactivity and dust transport in vacuum vessels, and support of Advanced Test Reactor irradiation experiments. This paper presents an overview of the programs engaged in the activities, which include the US-Japan TITAN collaboration, the US ITER program, the Next Generation Power Plant program and the tritium production program, and a presentation of ongoing experiments as well as a summary of recent results with emphasis on fusion relevant materials.

  6. Materials and Molecular Research Division annual report, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    Research is presented concerning materials science including metallurgy and ceramics; solid state physics; and materials chemistry; chemical sciences covering radiation science, chemical physics, and chemical energy; nuclear science; coal research; solar energy; magnetic fusion, conservation; and environmental research. (FS)

  7. Scalable Atomistic Simulation Algorithms for Materials Research

    Directory of Open Access Journals (Sweden)

    Aiichiro Nakano

    2002-01-01

    Full Text Available A suite of scalable atomistic simulation programs has been developed for materials research based on space-time multiresolution algorithms. Design and analysis of parallel algorithms are presented for molecular dynamics (MD simulations and quantum-mechanical (QM calculations based on the density functional theory. Performance tests have been carried out on 1,088-processor Cray T3E and 1,280-processor IBM SP3 computers. The linear-scaling algorithms have enabled 6.44-billion-atom MD and 111,000-atom QM calculations on 1,024 SP3 processors with parallel efficiency well over 90%. production-quality programs also feature wavelet-based computational-space decomposition for adaptive load balancing, spacefilling-curve-based adaptive data compression with user-defined error bound for scalable I/O, and octree-based fast visibility culling for immersive and interactive visualization of massive simulation data.

  8. Structural materials requirements for in-vessel components of fusion power plants

    International Nuclear Information System (INIS)

    Schaaf, B. van der

    2000-01-01

    The economic production of fusion energy is determined by principal choices such as using magnetic plasma confinement or generating inertial fusion energy. The first generation power plants will use deuterium and tritium mixtures as fuel, producing large amounts of highly energetic neutrons resulting in radiation damage in materials. In the far future the advanced fuels, 3 He or 11 B, determine power plant designs with less radiation damage than in the first generation. The first generation power plants design must anticipate radiation damage. Solid sacrificing armour or liquid layers could limit component replacements costs to economic levels. There is more than radiation damage resistance to determine the successful application of structural materials. High endurance against cyclic loading is a prominent requirement, both for magnetic and inertial fusion energy power plants. For high efficiency and compactness of the plant, elevated temperature behaviour should be attractive. Safety and environmental requirements demand that materials have low activation potential and little toxic effects under both normal and accident conditions. The long-term contenders for fusion power plant components near the plasma are materials in the range from innovative steels, such as reduced activation ferritic martensitic steels, to highly advanced ceramic composites based on silicon carbide, and chromium alloys. The steels follow an evolutionary path to basic plant efficiencies. The competition on the energy market in the middle of the next century might necessitate the riskier but more rewarding development of SiCSiC composites or chromium alloys

  9. Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY14 Report

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, Steven J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2014-10-01

    Laboratory corrosion testing of candidate alloys—including Zr-4 and Zr-2.5Nb representing the target solution vessel, and 316L, 2304, 304L, and 17-4 PH stainless steels representing process piping and balance-of-plant components—was performed in support of the proposed SHINE process to produce 99Mo from low-enriched uranium. The test solutions used depleted uranyl sulfate in various concentrations and incorporated a range of temperatures, excess sulfuric acid concentrations, nitric acid additions (to simulate radiolysis product generation), and iodine additions. Testing involved static immersion of coupons in solution and in the vapor above the solution, and was extended to include planned-interval tests to examine details associated with stainless steel corrosion in environments containing iodine species. A large number of galvanic tests featuring couples between a stainless steel and a zirconium-based alloy were performed, and limited vibratory horn testing was incorporated to explore potential erosion/corrosion features of compatibility. In all cases, corrosion of the zirconium alloys was observed to be minimal, with corrosion rates based on weight loss calculated to be less than 0.1 mil/year with no change in surface roughness. The resulting passive film appeared to be ZrO2 with variations in thickness that influence apparent coloration (toward light brown for thicker films). Galvanic coupling with various stainless steels in selected exposures had no discernable effect on appearance, surface roughness, or corrosion rate. Erosion/corrosion behavior was the same for zirconium alloys in uranyl sulfate solutions and in sodium sulfate solutions adjusted to a similar pH, suggesting there was no negative effect of uranium resulting from fluid dynamic conditions aggressive to the passive film. Corrosion of the candidate stainless steels was similarly modest across the entire range of exposures. However, some sensitivity to corrosion of the stainless steels was

  10. Vessel Operating Units (Vessels)

    Data.gov (United States)

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

  11. Materials and Molecular Research Division annual report, 1977

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    Progress in research in structure of materials, mechanical, and physical properties, solid state physics, and materials chemistry, including chemical structure, high temperature and surface chemistry, is reported. (FS)

  12. HTGR Base Technology Program. Task 2: concrete properties in nuclear environment. A review of concrete material systems for application to prestressed concrete pressure vessels

    International Nuclear Information System (INIS)

    Naus, D.J.

    1981-05-01

    Prestressed concrete pressure vessels (PCPVs) are designed to serve as primary pressure containment structures. The safety of these structures depends on a correct assessment of the loadings and proper design of the vessels to accept these loadings. Proper vessel design requires a knowledge of the component (material) properties. Because concrete is one of the primary constituents of PCPVs, knowledge of its behavior is required to produce optimum PCPV designs. Concrete material systems are reviewed with respect to constituents, mix design, placing, curing, and strength evaluations, and typical concrete property data are presented. Effects of extreme loadings (elevated temperature, multiaxial, irradiation) on concrete behavior are described. Finally, specialty concrete material systems (high strength, fibrous, polymer, lightweight, refractory) are reviewed. 235 references

  13. Materials selection, qualification and manufacturing of the in-vessel divertor cryopump for JET

    International Nuclear Information System (INIS)

    Papastergiou, S.; Obert, W.; Thompson, E.

    1994-01-01

    The introduction of a cryopump into the interior of a large tokamak raises several technical problems related to the thermal stresses, eddy current forces and choice of materials. The JET divertor cryopump has been optimized in terms of stresses, flow stability and operation - the liquid nitrogen cooled chevron structure in particular having to fulfill conflicting requirements at cryogenic temperatures. These requirements include good thermal conductivity in order to minimize thermal gradients (to reduce the radiative heat load onto the liquid helium circuit), high electrical resistivity (to minimize eddy current stresses), high mechanical strength and good mechanical formability. This paper reports on the materials selection based on measurements of properties at cryogenic and elevated temperatures and the development of an optimized thermal treatment combining solution heat treatment, brazing and precipitation hardening. It also reports on the successful development of various manufacturing technologies which have been employed including (a) techniques for brazing of the chosen copper alloy onto inconel and stainless steel, (b) surface blackening of the copper alloy with plasma sprayed ceramic coatings that are vacuum compatible and able to withstand temperatures between 70 K and 1135 K and (c) plasma spray deposition of copper onto stainless steel in order to produce an anisotropic composite material with improved thermal conductivity, high strength and high electrical resistivity for use at temperatures between 70 K and 650 K

  14. Materials technology and the energy problem : application to the reliability and safety of nuclear pressure vessels

    International Nuclear Information System (INIS)

    Garrett, G.G.

    1975-01-01

    In the U.S.A. over the past few months, widespread plant shutdowns because of cracking problems has produced considerable public pressure for a reappraisal of the reliability and safety of nuclear reactors. The awareness of such problems, and their solution, is particularly relevant to South Africa at this time. Some materials problems related to nuclear plant failure are examined in this paper. Since catastrophic failure (without prior warning from slow leakage) is in principle possible for light water (pressurised) reactors under operating conditions, it is essential to maintain rigorous manufacturing and quality control procedures, in conjunction with thorough and frequent examination by non-destructive testing methods. Although tests currently in progress in the U.S.A. on large-scale model reactors suggest that mathematical stress and failure analyses, for simple geometries at least, are sound, current in situ surveillance programmes aimed at categorizing the effects of irradiation are inadequate. In addition, the effects on materials properties and subsequent fracture resistance of the combined effects of irradiation and thermal shock (arising from the injection of emergency cooling water during a loss-of coolant accident) are unknown. The problem of stress corrosion cracking in stainless steel pipelines is considerable, and at present virtually impossible to predict. Much of the available laboratory data is inapplicable in that it cannot account for the complex interactions of stress state, temperature, material variations and segregation effects, and water chemistry, especially in conjunction with irradiation effects, that are experienced in an operating environment

  15. Neuroimaging standards for research into small vessel disease and its contribution to ageing and neurodegeneration

    NARCIS (Netherlands)

    Wardlaw, J.M.; Smith, E.E.; Biessels, G.J.; Cordonnier, C.; Fazekas, F.; Frayne, R.; Lindley, R.I.; O'Brien, J. T.; Barkhof, F.; Benavente, O.R.; Black, S.E.; Brayne, C.; Breteler, M.; Chabriat, H.; deCarli, C.; de Leeuw, F.E.; Doubal, F.; Duering, M.; Fox, N.C.; Greenberg, S.; Hachinski, V.; Kilimann, I.; Mok, V.; van Oostenbrugge, R.; Pantoni, L.; Speck, O.; Stephan, B.C.M.; Teipel, S.; Viswanathan, A.; Werring, D.; Chen, C.; Smith, C.; van Buchem, M.; Norrving, B.; Gorelick, P.B.; Dichgans, M.

    2013-01-01

    Cerebral small vessel disease (SVD) is a common accompaniment of ageing. Features seen on neuroimaging include recent small subcortical infarcts, lacunes, white matter hyperintensities, perivascular spaces, microbleeds, and brain atrophy. SVD can present as a stroke or cognitive decline, or can have

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

    International Nuclear Information System (INIS)

    Albertini, C.

    1997-01-01

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

  17. Investigations on the behaviour of reactor pressure vessel material 20 MnMoNi 55 during heat and stress relieving treatments. Vol. 1 and 2

    International Nuclear Information System (INIS)

    Blind, D.; Schroeder-Obst, D.; Herz, K.; Maidorn, C.

    1984-01-01

    Variation of various heat treatment parameters with regard to forging, hardening, tempering and stress-relieving has been applied to several heats of pressure vessel steels with the aim of testing the possibility to obtain higher notch impact energy values. On one hand the variation of heat treatment parameters within the limits of the current VdTUeV material sheet 401/4 5.80 did not result in outstanding improvements of toughness. On the other hand, when employing procedures which did not correspond to the specifications, e.g. tempering up to 100 h, an evident decrease of the upper shelf and an increase of the transition temperature could be observed. Nevertheless, the specified values were generally reached. Essentially, the observations on the test materials confirm, apart from a few exceptions, the positive practical experience with the material 20 MnMoNi 5 5. Based on these relations between thoughness and forging as well as heat treatment the manufacturer obtained, in accordance with the current research program, an outstanding improvement of toughness by means of various optimization measures which had the effect of optimal, evidently increased upper shelfs and which excluded difficulties concerning acceptance criteria, e.g. too high notch impact energy transition temperatures. (orig./IHOE) [de

  18. New Directions of Research in Molecules and Materials

    Indian Academy of Sciences (India)

    Wintec

    New Directions of Research in Molecules and Materials. Foreword. 'Materials' has ... Solution phase chemistry is a central aspect of materials as demonstrated by. Panchakarla and ... Changes at the atomic scale affect bulk properties such as ...

  19. Influence of gamma irradiation on the deterioration of reactor pressure vessel materials and on reactor dosimetry measurements. Final report

    International Nuclear Information System (INIS)

    Boehmer, B.; Konheiser, J.; Kumpf, H.; Noack, K.; Vladimirov, P.

    2002-10-01

    Radiation embrittlement of pressure vessel steel in mixed neutron-gamma fields is mostly determined by neutrons, but in some cases also by gamma-radiation. Depending on the reactor type, gamma radiation can influence evaluations of lead factors of surveillance specimens, effect the interpretation of results of irradiation experiments and finally, it can result in changed pressure vessel lifetime evaluations. The project aimed at the evaluation of the importance of gamma radiation for RPV steel damage for several types of light-water reactors. Absolute neutron and gamma fluence rate spectra had been calculated for the Russian PWR types VVER-440 and two core loading variants of VVER-1000, for a German 1300 MW PWR and a German 900 MW BWR. Based on the calculated spectra several flux integrals and radiation damage parameters were derived for the region of the azimuthal flux maxima in the mid-planes for different radial positions between core and biological shield, especially, at the inner and outer surfaces of the PV walls, at the (1/4)-PV-thickness and at the surveillance positions. Fissionable materials are often used as activation detectors for neutron fluence measurements. To get the real value the analysis demands to take into account the gamma induced fissions. Therefore, the part of these fissions in the total number of fissions was estimated for the detector reactions 237 Np(n,f) and 238 U(n,f) in the calculated neutron/gamma fields. It has been found that considerable corrections of the neutron fluence measurements can be necessary, especially in case of 238 U(n,f). Most of the calculations were performed using a three-dimensional synthesis of 2D/1D-flux distributions obtained by the S N -code DORT with the BUGLE-96T group cross-section library. (orig.) [de

  20. Integral analysis of debris material and heat transport in reactor vessel lower plenum

    International Nuclear Information System (INIS)

    Suh, K.Y.; Henry, R.E.

    1994-01-01

    An integral, fast-running, two-region model has been developed to characterize the debris material and heat transport in the reactor lower plenum under severe accident conditions. The debris bed is segregated into the oxidic pool and an overlying metallic layer. Debris crusts can develop on three surfaces: the top of the molten pool, the RPV wall, and the internal structures. To account for the decay heat generation, the crust temperature profile is assumed to be parabolic. The oxidic debris pool is homogeneously mixed and has the same material composition, and hence the same thermophysical properties, as the crusts, while the metallic constituents are assumed to rise to the top of the debris pool. Steady-state relationships are used to describe the heat transfer rates, with the assessment of solid or liquid state, and the liquid superheat in the pool being based on the average debris temperature. Natural convection heat transfer from the molten debris pool to the upper, lower and embedded crusts is calculated based on the pool Rayleigh number with the conduction heat transfer from the crusts being determined by the crust temperature profile. The downward heat flux is transferred to the lowest part of the RPV lower head through a crust-to-RPV contact resistance. The sideward heat flux is transferred to the upper regions of the RPV lower head as well as to the internal structures. The upward heat flux goes to the metal layer, water, or available heat sink structures above. Quenching due to water ingression is modeled separately from the energy transfer through the crust. The RPV wall temperature distribution and the primary system pressure are utilized to estimate challenges to the RPV integrity. ((orig.))

  1. Biofunctionalization of scaffold material with nano-scaled diamond particles physisorbed with angiogenic factors enhances vessel growth after implantation.

    Science.gov (United States)

    Schimke, Magdalena M; Stigler, Robert; Wu, Xujun; Waag, Thilo; Buschmann, Peter; Kern, Johann; Untergasser, Gerold; Rasse, Michael; Steinmüller-Nethl, Doris; Krueger, Anke; Lepperdinger, Günter

    2016-04-01

    Biofunctionalized scaffold facilitates complete healing of large defects. Biological constraints are induction and ingrowth of vessels. Angiogenic growth factors such as vascular endothelial growth factor or angiopoietin-1 can be bound to nano-scaled diamond particles. Corresponding bioactivities need to be examined after biofunctionalization. We therefore determined the physisorptive capacity of distinctly manufactured, differently sized nDP and the corresponding activities of bound factors. The properties of biofunctionalized nDPs were investigated on cultivated human mesenchymal stem cells and on the developing chicken embryo chorio-allantoic membrane. Eventually porous bone substitution material was coated with nDP to generate an interface that allows biofactor physisorption. Angiopoietin-1 was applied shortly before scaffold implantation into an osseous defect in sheep calvaria. Biofunctionalized scaffolds exhibited significantly increased rates of angiogenesis already one month after implantation. Conclusively, nDP can be used to ease functionalization of synthetic biomaterials. With the advances in nanotechnology, many nano-sized materials have been used in the biomedical field. This is also true for nano-diamond particles (nDP). In this article, the authors investigated the physical properties of functionalized nano-diamond particles in both in-vitro and in-vivo settings. The positive findings would help improve understanding of these nanomaterials in regenerative medicine. Copyright © 2015 Elsevier Inc. All rights reserved.

  2. Melting and solidification characteristics of a mixture of two types of latent heat storage material in a vessel

    Science.gov (United States)

    Yu, JikSu; Horibe, Akihiko; Haruki, Naoto; Machida, Akito; Kato, Masashi

    2016-11-01

    In this study, we investigated the fundamental melting and solidification characteristics of mannitol, erythritol, and their mixture (70 % by mass mannitol: 30 % by mass erythritol) as potential phase-change materials (PCMs) for latent heat thermal energy storage systems, specifically those pertaining to industrial waste heat, having temperatures in the range of 100-250 °C. The melting point of erythritol and mannitol, the melting peak temperature of their mixture, and latent heat were measured using differential scanning calorimetry. The thermal performance of the mannitol mixture was determined during melting and solidification processes, using a heat storage vessel with a pipe heat exchanger. Our results indicated phase-change (fusion) temperatures of 160 °C for mannitol and 113 and 150 °C for the mannitol mixture. Nondimensional correlation equations of the average heat transfer during the solidification process, as well as the temperature and velocity efficiencies of flowing silicon oil in the pipe and the phase-change material (PCM), were derived using several nondimensional parameters.

  3. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    Energy Technology Data Exchange (ETDEWEB)

    Myodo, Masato; Miyajima, Kazutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Okane, Shogo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  4. Mock-up test of remote controlled dismantling apparatus for large-sized vessels (contract research)

    International Nuclear Information System (INIS)

    Myodo, Masato; Miyajima, Kazutoshi; Okane, Shogo

    2001-03-01

    The Remote dismantling apparatus, which is equipped with multi-units for functioning of washing, cutting, collection of cut pieces and so on, has been constructed to dismantle the large-sized vessels in the JAERI's Reprocessing Test Facility (JRTF). The apparatus has five-axis movement capability and its operation is performed remotely. The mock-up tests were performed to evaluate the applicability of the apparatus to actual dismantling activities by using the mock-ups of LV-3 and LV-5 in the facility. It was confirmed that each unit was satisfactory functioned by remote operation. Efficient procedures for dismantling the large-sized vessel was studied and various date was obtained in the mock-up tests. This apparatus was found to be applicable for the actual dismantling activity in JRTF. (author)

  5. Gammatography of thick lead vessels

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  6. Compilation of contract research for the Materials Engineering Branch, Division of Engineering: Annual report for FY 1987

    International Nuclear Information System (INIS)

    1988-06-01

    This compilation of annual reports by contractors to the Materials Engineering Branch of the NRC Office of Research concentrates on achievements in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators, nondestructive examination of primary components, and in safety research for decommissioning and decontamination, on-site storage, and engineered safety features. This report, covering research conducted during Fiscal Year 1987 is the sixth volume of the series of NUREG-0975, ''Compilation of Contractor Research for the Materials Engineering Branch, Division of Engineering.''

  7. Corrosion Assessment of Candidate Materials for the SHINE Subcritical Assembly Vessel and Components FY15 Report

    Energy Technology Data Exchange (ETDEWEB)

    Pawel, Steven J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-01-01

    In the previous report of this series, a literature review was performed to assess the potential for substantial corrosion issues associated with the proposed SHINE process conditions to produce 99Mo. Following the initial review, substantial laboratory corrosion testing was performed emphasizing immersion and vapor-phase exposure of candidate alloys in a wide variety of solution chemistries and temperatures representative of potential exposure conditions. Stress corrosion cracking was not identified in any of the exposures up to 10 days at 80°C and 10 additional days at 93°C. Mechanical properties and specimen fracture face features resulting from slow-strain rate tests further supported a lack of sensitivity of these alloys to stress corrosion cracking. Fluid velocity was found not to be an important variable (0 to ~3 m/s) in the corrosion of candidate alloys at room temperature and 50°C. Uranium in solution was not found to adversely influence potential erosion-corrosion. Potentially intense radiolysis conditions slightly accelerated the general corrosion of candidate alloys, but no materials were observed to exhibit an annualized rate above 10 μm/y.

  8. Status of and materials research at SSLS

    International Nuclear Information System (INIS)

    Moser, H.O.; Casse, B.D.F.; Chew, E.P.; Cholewa, M.; Diao, C.Z.; Ding, S.X.D.; Kong, J.R.; Li, Z.W.; Hua, Miao; Ng, M.L.; Saw, B.T.; Mahmood, Sharain bin; Vidyaraj, S.V.; Wilhelmi, O.; Wong, J.; Yang, P.; Yu, X.J.; Gao, X.Y.; Wee, A.T.S.; Sim, W.S.; Lu, D.; Faltermeier, R.B.

    2005-01-01

    A short overview is given on the status of SSLS, its four operational and one forthcoming experimental facilities and their use for material science exemplified by selected work on electromagnetic metamaterials, arrays of nanorods for near-IR photonics, thin films of low dielectric constant materials for semiconductor manufacturing, nanoparticles and art objects

  9. Compilation of contract research for the Materials Engineering Branch, Division of Engineering Technology. Annual report for FY 1985. Volume 4

    International Nuclear Information System (INIS)

    1986-03-01

    The compilation of annual reports by contractors to the Materials Engineering Branch of the NRC Office of Research, concentrates on achievements in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators and for non-destructive examination of primary system components. This report, covering research conducted during Fiscal Year 1985, is the fourth volume of the series of NUREG-0975, Compilation of Contractor Research for the Materials Engineering Branch, Division of Engineering Technology

  10. Compilation of contract research for the Materials Engineering Branch, Division of Engineering Technology. Annual report for FY 1984. Volume 3

    International Nuclear Information System (INIS)

    1985-04-01

    This compilation of annual reports by contractors to the Materials Engineering Branch of the NRC Office of Research, concentrates on achievments in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators and for non-destructive examination of primary system components. This report, covering research conducted during Fiscal Year 1984, is the third volume of the series of NUREG-0975, compilation of Contractor Research for the Materials Engineering Branch, Division of Engineering Technology

  11. Heat-Induced, Pressure-Induced and Centrifugal-Force-Induced Exact Axisymmetric Thermo-Mechanical Analyses in a Thick-Walled Spherical Vessel, an Infinite Cylindrical Vessel, and a Uniform Disk Made of an Isotropic and Homogeneous Material

    Directory of Open Access Journals (Sweden)

    Vebil Yıldırım

    2017-07-01

    Full Text Available Heat-induced, pressure-induced, and centrifugal force-induced axisymmetric exact deformation and stresses in a thick-walled spherical vessel, a cylindrical vessel, and a uniform disk are all determined analytically at a specified constant surface temperature and at a constant angular velocity. The inner and outer pressures are both included in the formulation of annular structures made of an isotropic and homogeneous linear elastic material. Governing equations in the form of Euler-Cauchy differential equation with constant coefficients are solved and results are presented in compact forms. For disks, three different boundary conditions are taken into account to consider mechanical engineering applications. The present study is also peppered with numerical results in graphical forms.

  12. Estimation on the Flow Phenomena and the Pressure Loss for the Inlet Part of a Research Reactor Vessel

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Oh, Jae Min; Seo, Jae Kwang; Yoon, Ju Hyeon; Lee, Doo Jeong

    2009-01-01

    For a research reactor, a conceptual primary cooling system (PCS) was designed for an adequate cooling to the reactor core. The developed primary cooling circuit consisted of decay tanks, pumps, heat exchangers, vacuum breakers, some isolation and check valves, connection piping, and instruments. The main function of the primary cooling pumps (PCPs) of the PCS was to circulate the reactor coolant through the fuel core and the heat exchangers during a normal operation. The head according to the design flow rate which was determined by the thermal hydraulic design analysis for the core should be estimated to design the PCPs in the fluid system. The pressure loss in the PCS can be calculated by the dimensional analysis of the pipe flow and the head loss coefficient of the components. However, it is insufficient to estimate the pressure loss for 3-dimensional flow phenomena such as the flow path in the reactor with the theoretical dimensional analysis based on experimental data. The purpose of this research is to evaluate the pressure loss of the part of a research reactor vessel. For evaluating the pressure loss, the commercially available CFD computer model, FLUENT, was employed. First, for validating the application of FLUENT to the pressure loss, a simple case was calculated and compared with the Idelchik empirical correlation. Secondly, several cases for the inlet part of a research reactor vessel were estimated by a FLUENT 3- dimensional calculation

  13. CTD and Water Sample Data from Research Vessel Robert Gordon Sproul in the NE Pacific, 24 October 2013 (NCEI Accession 0157082)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The expedition by research vessel Robert Gordon Sproul from 23 to 25 October 2013 had the objective to recover a broken mooring from the CORC project (Consortium on...

  14. CTD and Water Sample Data from Research Vessel New Horizon in the NE Pacific, 19-22 September 2008 (NCEI Accession 0156931)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The expedition by research vessel New Horizon from 19 to 22 September 2008 had the objective to deploy a number of moored platforms for the CORC project (Consortium...

  15. Processed CTD and Water Sample Data from Research Vessel Ocean Starr in the NE Pacific, Aug. 31 and Sept. 01, 2012 (NCEI Accession 0156932)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The expedition by research vessel Ocean Starr on Aug. 31 and Sept. 01, 2012 had the objective to recover and re-deploy a number of moored platforms from the CORC...

  16. Processed CTD and Water Sample Data from Research Vessel Roger Revelle, Expedition RR1214, in the NE Pacific in November 2012 (NCEI Accession 0156228)

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — Expedition RR1214 by research vessel Roger Revelle was primarily a transit from French Polynesia to the US mainland. However, a small scientific program was...

  17. Raw material variability as archaeological tools: Preliminary results from a geochemical study of the basalt vessel workshop at Iron Age Tel Hazor, Israel

    Directory of Open Access Journals (Sweden)

    Tatjana Gluhak

    2016-10-01

    Full Text Available The discovery of a basalt vessel workshop at Tel Hazor, one of the most important Iron Age sites in the Near East, marks a turning point in our understanding of stone artifact production and distribution during the1st millennium BCE. It offers a rare opportunity to characterize ancient raw material sources, production sites, and study production, trade and distribution systems. The basalt vessel workshop, the only one of its kind in the Levant, produced large quantities of bowl preforms and production waste. To better understand the production and distribution systems behind this specialized production center, in 2011 we initiated a focused geochemical project that concentrated on the products of this unique workshop.  We measured the major and trace element composition of 44 unfinished basalt vessels from the workshop and other contexts at Hazor, and can demonstrate that the majority of these objects were derived from one specific, geochemically well-constrained, basaltic rock source. Only a few bowls clearly deviate from this geochemical composition and were produced using raw material from other sources. Thus, we believe that one major quarry existed that supplied the Hazor workshop with the majority of the basaltic raw material. The products from this specific extraction site provide us with a “Hazor reference group” that can be used to test whether or not finished vessels from Hazor and contemporary sites were produced in the Hazor workshop.

  18. Magnetic materials research with polarized neutrons

    International Nuclear Information System (INIS)

    Hammer, J.; Rauch, H.; Badurek, G.

    1980-01-01

    In order to study the mechanisms of time dependent effects in magnetic materials with superparamagnetic or spinglass behaviour as well as in ferromagnetic materials a 'dynamic neutron depolarization' system has been developed as a beam hole experiment at the TRIGA Mark II Reactor in Vienna. In the course of this experiment an increasing or decreasing polarization can be observed as a consequence of the interaction between spins of the polarized neutron beam and the magnetic structure if the magnetic clusters in the sample are stimulated by a short magnetic pulse, lasting up to a few seconds. In accordance with numerical calculations and theoretical considerations we can draw conclusions from dynamics in the range of 10 ms to 1 h within magnetic materials which give us additional information that cannot be obtained from experiments used so far

  19. Nuclear physics methods in materials research

    International Nuclear Information System (INIS)

    Bethge, K.; Baumann, H.; Jex, H.; Rauch, F.

    1980-01-01

    Proceedings of the seventh divisional conference of the Nuclear Physics Division held at Darmstadt, Germany, from 23rd through 26th of September, 1980. The scope of this conference was defined as follows: i) to inform solid state physicists and materials scientists about the application of nuclear physics methods; ii) to show to nuclear physicists open questions and problems in solid state physics and materials science to which their methods can be applied. According to the intentions of the conference, the various nuclear physics methods utilized in solid state physics and materials science and especially new developments were reviewed by invited speakers. Detailed aspects of the methods and typical examples extending over a wide range of applications were presented as contributions in poster sessions. The Proceedings contain all the invited papers and about 90% of the contributed papers. (orig./RW)

  20. The importance of Forensic research in the Nuclear Power Industry. What the OECD Three Mile Island reactor vessel investigation. Means to the future of commercial nuclear power

    International Nuclear Information System (INIS)

    Rogers, K.C.

    1994-01-01

    TMI-2 altered the perception of the likelihood of severe accidents and their precursors and shortly after the accident, changes began at the NRC. NRC required its nuclear power licensees to make a rather large number of back fits to respond to the lessons learned; NRC broadened the study of severe accident phenomena (focus on studies involving molten core materials); other lessons learned concerned the severe accident source terms and the shift from a deterministic tradition point of view to a probabilistic risk assessment formalism. A Revised Severe Accident Research Program Plan was issued in 1989. A review of what was known before the TMI-Vessel Investigation Project and what was not known, is presented

  1. Battery Materials Synthesis | Transportation Research | NREL

    Science.gov (United States)

    thin-film. NREL's development of inexpensive, high-energy-density electrode materials is challenging introduction of metal oxide and hybrid inorganic-organic surface modification via atomic layer deposition has method for applying conformal thin film coatings to highly textured surfaces. These coatings have been

  2. Methods for estimation and enhancing of resistance of pressure vessel materials to fracture at different stages of service taking into account actual dimensions of the construction

    International Nuclear Information System (INIS)

    Pokrovsky, V.V.; Ivanchenko, A.G.

    1998-01-01

    In the present report a method is proposed for assessment of cracked materials fracture toughness over a wide range of temperatures taking into account the size-effect of structural elements. The procedure proposed was evaluated on specimens of different thicknesses (25... 150 mm) and geometries from the parent metal and welded joint metal of the WWER-Type nuclear reactor pressure vessels of different classes of strength. The method of enhancing of fracture resistance of pressure vessel materials has been develop which is based on warm prestressing of materials with cracks. The stability of the favourable effect of the warm prestressing has been, investigated and shown for the above steels after their long term (to 24000 hours) keeping under static loading and temperature of 350 deg C, under different conditions of cyclic loading, corrosive action. A model and calculation procedure are proposed for predicting the influence of thermomechanical loading conditions on the resistance of reactor steels to brittle fracture. (authors)

  3. Radiation embrittlement of WWER-1000 reactor vessel steels

    International Nuclear Information System (INIS)

    Nikolaeva, A.V.; Nikolaev, Yu.A.; Kevorkyan, Yu.R.

    2001-01-01

    Results obtained on the blank samples of materials of the WWER-1000 vessels irradiated by low density neutron flux are discussed. Chemical composition of the materials is characterized by the low content of the impurities (copper and phosphorus) and high content of nickel. Dependence of the radiation embrittlement of the WWER-1000 vessel materials on metallurgic variables and damage dose is treated. The research showed that nickel largely enhanced the radiation embrittlement. New dependences for determination of the radiation embrittlement real rate of the WWER-1000 vessel materials and its conservative estimation were developed [ru

  4. Research and development on in-service inspection system for reactor vessel of FBR's

    International Nuclear Information System (INIS)

    Rindo, Hiroshi; Mitabe, Noriaki; Ara, Kuniaki; Nagai, Keiichi; Otaka, Masahiko

    1993-01-01

    In-Service Inspection (ISI) is required for main components and piping of FBRs. Visual test and volumetric examination of the reactor vessel (RV) from the outer surface are to be performed under severe conditions such as limited space, high temperature and high gamma dose rate during the reactor shutdown. Therefore, ISI should be performed by using a remote operation system, and the ISI system should be very compact. PNC has been developing the ISI system to apply to the RV inspection. Verification and performance tests of ISI system were carried out by use of the RV test model. This paper describes the system structure, system verification tests including operation and controlling the inspection robot, the functions of the visual test and the volumetric examination under the high temperature

  5. Research Opportunities for Materials with Ultrafine Microstructures

    Science.gov (United States)

    1989-12-31

    network with uniformly large pores (see Figure 2). An acidic DCCA, such as oxalic acid , in contrast, results in a somewhat smaller-scale network...bacteriorhodopsin macromolecule 12 FIGURE 2 Control of sol-gel processing with organic acid DCCAs 16 FIGURE 3 Densification microstructures for SiO 2 gels...monodispersed particles and hydrothermal synthesis of composites. Of recent interest in polymeric materials has been the development of rigid-rod

  6. Computational research on lithium ion battery materials

    Science.gov (United States)

    Tang, Ping

    Crystals of LiFePO4 and related materials have recently received a lot of attention due to their very promising use as cathodes in rechargeable lithium ion batteries. This thesis studied the electronic structures of FePO 4 and LiMPO4, where M=Mn, Fe, Co and Ni within the framework of density-functional theory. The first study compared the electronic structures of the LiMPO 4 and FePO4 materials in their electrochemically active olivine form, using the LAPW (linear augmented plane wave) method [1]. A comparison of results for various spin configurations suggested that the ferromagnetic configuration can serve as a useful approximation for studying general features of these systems. The partial densities of states for the LiMPO4 materials are remarkably similar to each other, showing the transition metal 3d states forming narrow bands above the O 2p band. By contrast, in absence of Li, the majority spin transition metal 3d states are well-hybridized with the O 2p band in FePO4. The second study compared the electronic structures of FePO4 in several crystal structures including an olivine, monoclinic, quartz-like, and CrVO4-like form [2,3]. For this work, in addition to the LAPW method, PAW (Projector Augmented Wave) [4], and PWscf (plane-wave pseudopotential) [5] methods were used. By carefully adjusting the computational parameters, very similar results were achieved for the three independent computational methods. Results for the relative stability of the four crystal structures are reported. In addition, partial densities of state analyses show qualitative information about the crystal field splittings and bond hybridizations and help rationalize the understanding of the electrochemical and stability properties of these materials.

  7. Research Progress of Building Materials Used in Construction Land

    Science.gov (United States)

    Niu, Yan

    2018-01-01

    Construction land preparation is an important aspect of land remediation project. The research of materials in the process of land improvement is the foundation and the core. Therefore, it is necessary to study the materials that may be involved in the process of building land preparation. In this paper, the research on the construction materials such as recycled concrete, geosynthetics, soil stabilizers, soil improvers, building insulation materials and inorganic fibrous insulation materials, which are commonly used in construction sites, is reviewed and discussed in this paper. Land remediation project involved in the construction of land materials to provide reference.

  8. 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

  9. Engineer Research and Development Center's Materials Testing Center (MTC)

    Data.gov (United States)

    Federal Laboratory Consortium — The Engineer Research and Development Center's Materials Testing Center (MTC) is committed to quality testing and inspection services that are delivered on time and...

  10. Environmental, Health, and Safety Research Needs for Engineered Nanoscale Materials

    National Research Council Canada - National Science Library

    Alderson, Norris; Alexander, Catherine; Merzbacher, Celia; Chernicoff, William; Middendorf, Paul; Beck, Nancy; Chow, Flora; Poster, Dianne; Danello, Mary Ann; Barrera, Enriqueta

    2006-01-01

    ...) research and information needs related to understanding and management of potential risks of engineered nanoscale materials that may be used, for example, in commercial or consumer products, medical...

  11. Relating SLA Research to Language Teaching Materials

    Directory of Open Access Journals (Sweden)

    Vivian J. Cook

    1998-12-01

    Full Text Available Abstract This article discusses applications of Second Language Acquisition (SLA research to the preparation of language coursebooks. The author suggests a number of ways in which SLA research findings can help improve coursebooks and thereby enhance the learning of large numbers of students. Research leads us to consider learners as genuine speakers of the L2, as bilinguals who still have an L1 present in their minds and who do not all go about learning the L2 in the same way. Few coursebooks take into account these and other findings of SLA research, for example: that the acquisition of basic syntax precedes the acquisition of inflectional morphology, that most of the syntax to be learned is really part of the lexicon, or that vocabulary needs to be encountered in a structural and semantic context in order to be effectively acquired. Coursebook authors also need to bear in mind that pronunciation is necessary not only for communication but also for the actual learning of L2 forms, and that some aspects of the L2 writing system need to be explicitly taught. The author provides two sample lessons to illustrate how these research findings might be applied to the writing of a coursebook.

  12. Materials Research Department annual report 2000

    DEFF Research Database (Denmark)

    2001-01-01

    , educational activities and academic activities, such as collaboration with other research institutions, committee work and a list of publications. Furthermore, the main figures outlining the funding andexpenditures of the Department are given. Lists of staff members and visiting scientists are included....

  13. Process Research on Polycrystalline Silicon Material (PROPSM)

    Science.gov (United States)

    Culik, J. S.; Wrigley, C. Y.

    1985-01-01

    Results of hydrogen-passivated polycrysalline silicon solar cell research are summarized. The short-circuit current of solar cells fabricated from large-grain cast polycrystalline silicon is nearly equivalent to that of single-crystal cells, which indicates long bulk minority-carrier diffusion length. Treatments with molecular hydrogen showed no effect on large-grain cast polycrystalline silicon solar cells.

  14. Stability of ferritic steel to higher doses: Survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems

    International Nuclear Information System (INIS)

    Blagoeva, D.T.; Debarberis, L.; Jong, M.; Pierick, P. ten

    2014-01-01

    This paper is illustrating the potential of the well-known low alloyed clean steels, extensively used for the current light water Reactor Pressure Vessels (RPV) steels, for a likely use as a structural material also for the new generation nuclear systems. This option would provide, especially for large components, affordable, easily accessible and a technically more convenient solution in terms of manufacturing and joining techniques. A comprehensive comparison between several sets of surveillance and research data available for a number of RPV clean steels for doses up to 1.5 dpa, and up to 12 dpa for 9%Cr steels, is carried out in order to evaluate radiation stability of the currently used RPV clean steels even at higher doses. Based on the numerous data available, positive preliminary conclusions are drawn regarding the eventual use of clean RPV steels for the massive structural components of the new reactor systems. - Highlights: • Common embrittlement trend between RPV and advanced steels till intermediate doses. • For doses >1.5 dpa, damage rate saturation tendency is observed for RPV steels. • RPV steels might be conveniently utilised also outside their foreseen dose range

  15. Department of Materials Research by Computers - Overview

    International Nuclear Information System (INIS)

    Parlinski, K.

    2000-01-01

    Full text: During 1999 the main activity of the Department has been gradually moved to ab initio calculations. For that we have used the approach of density functional theory with either local density approximation (LDA) or generalized gradient approximation (GGA). This approach allows to find the structure and dynamics of any system which can be represented by a supercell with periodic boundary conditions. Our interests were limited to study of structure and dynamics of crystals. We have used two different packages of software: CASTEP and VASP and the pseudopotentials delivered with these programs. This method is parameter-free, which means that one needs to know only the physical constants, like Planck constant, element masses and electron charge, in order to get a quantitative result. We have concentrated our efforts around four subjects: calculation of phonon dispersion curves for polar crystals with LO/TO splitting, calculations of lattice dynamics of chalcopyrites, calculations of energy barriers in molecular crystals, and calculations of elastic properties and phase transitions in geologically important materials. We have calculated the phonon dispersion curves in ionic cubic MgO crystal. The phonon modes at Γ point are split to LO and TO modes. We have proposed a method to calculate this splitting by an elongated supercell. The results agree very well with the coherent inelastic neutron scattering data. Similar effects have been considered in hexagonal GaN, rhombohedral LiNbO 3 , and tetragonal Sn0 2 . In the two last crystals soft modes, responsible for the phase transitions, were found. Intensive calculations were carried out for tetragonal chalcopyrites structure. Each unit cell contains 16 atoms. By using enlarged supercell of 2 x 2 x 1 size with 64 atoms we could obtain valid phonon dispersion curves for CuInSe 2 , AgGaSe 2 , AgGaTe 2 , which agree with neutron data and Raman scattering results. Studies of the molecular motion in KSCN crystal were

  16. Nuclear materials research progress reports for 1977

    International Nuclear Information System (INIS)

    Olander, D.R.

    1977-12-01

    Research is reported concerning radiation enhancement of stress corrosion cracking of Zircaloy, surface chemistry of epitaxial Si deposited by thermal cracking of silane, thermal gradient migration of metallic inclusions in UO 2 , molecular beam studies of atomic H and reduction of oxides, mass transfer and reduction of UO 2 , kinetics of laser pulse vaporization of UO 2 , retention and release of water by UO 2 pellets, and solubility of H in UO 2

  17. Nuclear materials research progress reports for 1979

    International Nuclear Information System (INIS)

    Olander, D.R.

    1979-12-01

    Research is presented concerning iodide stress corrosion cracking of zircaloy, self-diffusion of oxygen in hypostoichiometric urania, surface chemistry of epitaxial silicon deposition by thermal cracking of silane, kinetics of laser pulse vaporization of UO 2 , gas laser model for laser induced evaporation, solubility of hydrogen in uranium dioxide, thermal gradient migration of metallic inclusions in UO 2 , molecular beam studies of atomic hydrogen reduction of oxides, and thermal gradient brine-inclusion migration in salt

  18. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    Energy Technology Data Exchange (ETDEWEB)

    Leclercq, Sylvain, E-mail: sylvain.leclercq@edf.f [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Lidbury, David [SERCO Assurance - Walton House, 404 Faraday Street, Birchwood Park, Warrington, Cheshire WA3 6GA (United Kingdom); Van Dyck, Steven [SCK-CEN, Nuclear Material Science, Boeretang 200, BE, 2400 Mol (Belgium); Moinereau, Dominique [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France); Alamo, Ana [CEA Saclay, DEN/DSOE, 91191 Gif-sur-Yvette (France); Mazouzi, Abdou Al [EDF R and D, Materials and Mechanics of Components, Avenue des Renardieres - Ecuelles, 77818 Moret sur Loing Cedex (France)

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach

  19. PERFORM 60 - Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    Science.gov (United States)

    Leclercq, Sylvain; Lidbury, David; Van Dyck, Steven; Moinereau, Dominique; Alamo, Ana; Mazouzi, Abdou Al

    2010-11-01

    In nuclear power plants, materials may undergo degradation due to severe irradiation conditions that may limit their operational life. Utilities that operate these reactors need to quantify the ageing and the potential degradations of some essential structures of the power plant to ensure safe and reliable plant operation. So far, the material databases needed to take account of these degradations in the design and safe operation of installations mainly rely on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and continuous progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the effects of irradiation on materials microstructure. A first step towards this goal has been successfully reached through the development of the RPV-2 and Toughness Module numerical tools by the scientific community created around the FP6 PERFECT project. These tools allow to simulate irradiation effects on the constitutive behaviour of the reactor pressure vessel low alloy steel, and also on its failure properties. Relying on the existing PERFECT Roadmap, the 4 years Collaborative Project PERFORM 60 has mainly for objective to develop multi-scale tools aimed at predicting the combined effects of irradiation and corrosion on internals (austenitic stainless steels) and also to improve existing ones on RPV (bainitic steels). PERFORM 60 is based on two technical sub-projects: (i) RPV and (ii) internals. In addition to these technical sub-projects, the Users' Group and Training sub-project shall allow representatives of constructors, utilities, research organizations… from Europe, USA and Japan to receive the information and training to get their own appraisal on limits and potentialities of the developed tools. An important effort will also be made to teach young

  20. Pressurized Vessel Slurry Pumping

    International Nuclear Information System (INIS)

    Pound, C.R.

    2001-01-01

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

  1. Radiation research of materials using irradiation capsules

    International Nuclear Information System (INIS)

    Chamrad, B.

    1976-01-01

    The methods are briefly characterized of radiation experiments on the WWR-S research reactor. The irradiation capsule installed in the reactor including the electronic instrumentation is described. Irradiated samples temperature is stabilized by an auxiliary heat source placed in the irradiation space. The electronic control equipment of the system is automated. In irradiation experiments, experimental and operating conditions are recorded by a digital measuring centre with electric typewriter and paper tape data recording and by an analog compensating recorder. The irradiation experiment control system controls irradiated sample temperature, the supply current size and the heating element temperature of the auxiliary stabilizing source, inert and technological pressures of the capsule atmosphere and the thermostat temperature of the thermocouple junctions. (O.K.)

  2. Thermogravimetric research of hydrogen storage materials

    International Nuclear Information System (INIS)

    Kleperis, J; Grinberga, L; Ergle, M; Chikvaidze, G; Klavins, J

    2007-01-01

    During thermogravimetric research of metal hydrides we noticed mass growth of samples above 200 deg. C even in an argon atmosphere. Further heating is leading to the growth of weight up to 2-7 weight% till 500 0 C. Second run of the same sample without taking out of DTA instrument gave only small mass changes, indicating that noticed mass increase during first run is permanent. Microscope and elemental analyses were made to determine the reason of mass growth. XRD inspection revealed the formation of new phase with bunsenite NiO structure with deformed cubic structure. The new phase is no more active to hydrogen sorption/desorption. Our results demonstrated that the usage of hydrogen storage alloys AB 5 must be taken with care - it is important not to exceed some critical temperature were irreversible structural, compositional and morphological changes will occur

  3. Applied solid state science advances in materials and device research

    CERN Document Server

    Wolfe, Raymond

    2013-01-01

    Applied Solid State Science: Advances in Materials and Device Research, Volume 4 covers articles on single crystal compound semiconductors and complex polycrystalline materials. The book discusses narrow gap semiconductors and solid state batteries. The text then describes the advantages of hot-pressed microcrystalline compacts of oxygen-octahedra ferroelectrics over single crystal materials, as well as heterostructure junction lasers. Solid state physicists, materials scientists, electrical engineers, and graduate students studying the subjects being discussed will find the book invaluable.

  4. 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

  5. Physical protection of radioactive materials in a University Research Institute

    International Nuclear Information System (INIS)

    Boeck, H.

    1998-01-01

    Although nuclear research centers attached to universities usually do not keep large inventories of radioactive or special nuclear material, the mentioned material has still to be under strict surveillance and safeguards if applicable. One problem in such research centers is the large and frequent fluctuation of persons - mainly students, scientists or visiting guest scientists - using such materials for basic or applied research. In the present paper an overview of protective actions in such a research institute will be given and experience of more than 36 years will be presented. (author)

  6. Materials and corrosion programs sponsored by the Gas Research Institute

    International Nuclear Information System (INIS)

    Flowers, A.

    1980-01-01

    The paper deals briefly with the Gas Research Institute and its research in materials and corrosion. As a not-for-profit organization, the Gas Research Institute plans, finances, and manages applied and basic research and technological development programs associated with gaseous fuels. These programs are in the general areas of production, transportation, storage, utilization and conservation of natural and manufactured gases and related products. Research results, whether experimental or analytical, are evaluated and publicly disseminated. Materials and corrosion research is concentrated in the SNG from Coal and Non-fossil Hydrogen subprograms

  7. Governing the postmortem procurement of human body material for research.

    Science.gov (United States)

    Van Assche, Kristof; Capitaine, Laura; Pennings, Guido; Sterckx, Sigrid

    2015-03-01

    Human body material removed post mortem is a particularly valuable resource for research. Considering the efforts that are currently being made to study the biochemical processes and possible genetic causes that underlie cancer and cardiovascular and neurodegenerative diseases, it is likely that this type of research will continue to gain in importance. However, post mortem procurement of human body material for research raises specific ethical concerns, more in particular with regard to the consent of the research participant. In this paper, we attempt to determine which consent regime should govern the post mortem procurement of body material for research. In order to do so, we assess the various arguments that could be put forward in support of a duty to make body material available for research purposes after death. We argue that this duty does in practice not support conscription but is sufficiently strong to defend a policy of presumed rather than explicit consent.

  8. 2004 research briefs :Materials and Process Sciences Center.

    Energy Technology Data Exchange (ETDEWEB)

    Cieslak, Michael J.

    2004-01-01

    This report is the latest in a continuing series that highlights the recent technical accomplishments associated with the work being performed within the Materials and Process Sciences Center. Our research and development activities primarily address the materials-engineering needs of Sandia's Nuclear-Weapons (NW) program. In addition, we have significant efforts that support programs managed by the other laboratory business units. Our wide range of activities occurs within six thematic areas: Materials Aging and Reliability, Scientifically Engineered Materials, Materials Processing, Materials Characterization, Materials for Microsystems, and Materials Modeling and Simulation. We believe these highlights collectively demonstrate the importance that a strong materials-science base has on the ultimate success of the NW program and the overall DOE technology portfolio.

  9. Response margins investigation of piping dynamic analyses using the independent support motion method and PVRC [Pressure Vessel Research Committee] damping

    International Nuclear Information System (INIS)

    Bezler, P.; Wang, Y.K.; Reich, M.

    1988-03-01

    An evaluation of Independent Support Motion (ISM) response spectrum methods of analysis coupled with the Pressure Vessel Research Committee (PVRC) recommendation for damping, to compute the dynamic component of the seismic response of piping systems, was completed. Response estimates for five piping/structural systems were developed using fourteen variants of the ISM response spectrum method, the Uniform Support Motions response spectrum method and the ISM time history analysis method, all based on the PVRC recommendations for damping. The ISM/PVRC calculational procedures were found to exhibit orderly characteristics with levels of conservatism comparable to those obtained with the ISM/uniform damping procedures. Using the ISM/PVRC response spectrum method with absolute combination between group contributions provided consistently conservative results while using the ISM/PVRC response spectrum method with square root sum of squares combination between group contributions provided estimates of response which were deemed to be acceptable

  10. Pressure vessel design manual

    CERN Document Server

    Moss, Dennis R

    2013-01-01

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

  11. Development of a dynamic in vitro model of a stented blood vessel to evaluate the effects of stent strut material selection and surface coating on smooth muscle cell response

    Science.gov (United States)

    Winn, Bradley Huegh

    formation of this new tissue, primarily consisting of VSMCs of the synthetic phenotype and their subsequent extracellular matrix, is the sole causation of in-stent restenosis since the stent serves to prevent elastic recoil and negative remodeling. This doctoral research program is focused on endovascular stent biomaterials science and engineering. Overall, this doctoral project is founded on the hypothesis that smooth muscle cell hyperplasia, as an important causative factor for vascular restenosis following endovascular stent deployment, is triggered by the various effects of stent strut contact on the vessel wall including contact forces and material biocompatibility. In this program, a dynamic in vitro model of a stented blood vessel aimed at evaluating the effect of stent strut material selection, and surface coating on smooth muscle cell response was developed. The in vitro stented artery model was validated through the proliferation of VSMC in contact with stent struts. Additionally, it was demonstrated that, with respect to known biocompatible materials such as Nitinol and 316L stainless steel, DNA synthesis and alpha-actin expression, as indicators of VSMC phenotype, are independent of stent material composition. Furthermore, hydroxyapatite was shown to be a biocompatible stent surface coating with acceptable post-strain integrity. This coating was shown in a feasibility study to be capable of serving as a favorable drug delivery platform able to reliably deliver locally therapeutic doses of bisphosphonates, such as alendronate, to control VSMC proliferation in an in vitro model of a stented blood vessel. This stent coating/drug combination may be effective for reducing restenosis as a result of VSMC hyperplasia in vivo.

  12. Magnetic materials in Japan research, applications and potential

    CERN Document Server

    2013-01-01

    Please note this is a Short Discount publication. This, the third report in Elsevier's Materials Technology in Japan series, concentrates on magnetic materials as a topic gaining worldwide attention, and each chapter looks not only at current research, but also describes the technology as it is being applied and its future potential. Magnetic-related research is the second largest field of research in Japan after semiconductors, with the estimated number of researchers and engineers engaged in magnetics-related activities currently at 20,000. This research report serves as both a review of

  13. Damage by radiation in structural materials of BWR reactor vessels; Dano por radiacion en materiales estructurales de vasijas de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E. [Departamento de Sintesis y Caracterizacion de Materiales, Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2002-07-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10{sup 18} n/cm{sup 2}) in the TRIGA Mark III Salazar reactor and separately with Ni{sup +3} ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A{sup 2}). (Author)

  14. Compilation of contract research for the Materials Engineering Branch, Division of Engineering: Annual report for FY 1988

    International Nuclear Information System (INIS)

    1989-05-01

    This compilation of annual reports by contractors to the Materials Engineering Branch of the NRC Office of Research concentrates on achievements in safety research for the primary system of commercial light water power reactors, particularly with regard to reactor vessels, primary system piping, steam generators, nondestructive examination of primary components, and in safety research for decommissioning and decontamination, on-site storage and engineered safety features. The Materials Engineering Branch assembles abbreviated reports from all the branch contractors and publishes them in a single annual report as soon after the end of the year as possible so that the information developed throughout the year can be promptly used in the safety-regulatory process. This report, covering research conducted during Fiscal Year 1988 is the seventh volume of the series of NUREG-0975, ''Compilation of Contractor Research for the Materials Engineering Branch, Division of Engineering.'' Individual projects are processed separately for the data bases

  15. PREFACE: MRS International Materials Research Conference (IMRC-2008)

    Science.gov (United States)

    Wang, Zhanguo; Qiu, Yong; Li, Yongxiang

    2009-03-01

    This volume contains selected papers presented at the MRS International Materials Research Conference (IMRC-2008) held in Chongqing, China, 9-12 June 2008. IMRC-2008 included 9 symposia of A. Eco/Environmental Materials, B. Sustainable Energy Materials, C. Electronic Packaging Materials, D. Electronic Materials, E. Materials and Processes for Flat-panel Displays, F. Functional Ceramics, G. Transportation Materials, H. Magnesium and I. Biomaterials for Medical Applications. Nearly 1200 participants from 33 countries attended the conference, and the conference organizers received more than 700 papers. After the peer review processes, 555 papers were selected to be published in 9 Journals or proceedings, including J. of Materials Research (JMR), Rare Metal Materials and Engineering, J. of Univ. Science and Technology Beijing, Biomedical Materials: Materials for Tissue Engineering and Regenerative Medicine, Chinese Journal of Aeronautics, Materials Science Forum, and Journal of Physics: Conference Series. Among the 555 selected papers, 91 papers are published in this volume, and the topics mainly cover electronic matrials, processes for flat-panel displays and functional ceramics. The editors would like to give special thanks to the graduate students Liwu Jiang, Ming Li and Di He from Beihang University for their hard work compiling and typesetting each paper in this volume. Zhanguo Wang, Yong Qiu and Yongxiang Li Editors

  16. Experiences of packaging research outputs into extension materials

    African Journals Online (AJOL)

    Mo

    Research dissemination is one component of research that still faces many hindrances, ... time-frames for dissemination activities going beyond project phase-out in order to maximise ..... Available or upcoming extension materials, with cost and availability ..... Renewable Natural Resources Research Strategy, Annual.

  17. Social justice and research using human biological material: A ...

    African Journals Online (AJOL)

    Social justice and research using human biological material: A response to Mahomed, Nöthling-Slabbert and Pepper. ... South African Medical Journal ... In a recent article, Mahomed, Nöthling-Slabbert and Pepper proposed that research participants should be entitled to share in the profits emanating from such research ...

  18. US Navy Transfers Research Vessel to Philippine Navy > U.S. Pacific Command

    Science.gov (United States)

    , 2016 EMAIL PRINT Photos 1 of 1 SAN DIEGO (April 27, 2016) - Vice Adm. Nora Tyson, left, commander of Philippines oceanographic research and study capabilities. SAN DIEGO (April 27, 2016) - Vice Adm. Nora Tyson part in that relationship," said Vice Adm. Nora Tyson, commander, U.S. 3rd Fleet. "Today, we

  19. High throughput materials research and development for lithium ion batteries

    Directory of Open Access Journals (Sweden)

    Parker Liu

    2017-09-01

    Full Text Available Development of next generation batteries requires a breakthrough in materials. Traditional one-by-one method, which is suitable for synthesizing large number of sing-composition material, is time-consuming and costly. High throughput and combinatorial experimentation, is an effective method to synthesize and characterize huge amount of materials over a broader compositional region in a short time, which enables to greatly speed up the discovery and optimization of materials with lower cost. In this work, high throughput and combinatorial materials synthesis technologies for lithium ion battery research are discussed, and our efforts on developing such instrumentations are introduced.

  20. Advances in thermoelectric materials research: Looking back and moving forward.

    Science.gov (United States)

    He, Jian; Tritt, Terry M

    2017-09-29

    High-performance thermoelectric materials lie at the heart of thermoelectrics, the simplest technology applicable to direct thermal-to-electrical energy conversion. In its recent 60-year history, the field of thermoelectric materials research has stalled several times, but each time it was rejuvenated by new paradigms. This article reviews several potentially paradigm-changing mechanisms enabled by defects, size effects, critical phenomena, anharmonicity, and the spin degree of freedom. These mechanisms decouple the otherwise adversely interdependent physical quantities toward higher material performance. We also briefly discuss a number of promising materials, advanced material synthesis and preparation techniques, and new opportunities. The renewable energy landscape will be reshaped if the current trend in thermoelectric materials research is sustained into the foreseeable future. Copyright © 2017 The Authors, some rights reserved; exclusive licensee American Association for the Advancement of Science. No claim to original U.S. Government Works.

  1. Procurement of replacement pressure vessels for MURR

    International Nuclear Information System (INIS)

    Meyer, W.A. Jr.; Edwards, C.B. Jr.; McKibben, J.C.; Schoone, A.R.

    1989-01-01

    The University of Missouri Research Reactor Facility (MURR) located in Columbia, Missouri, is the highest powered, highest steady-state flux university research reactor in the United States. The reactor is a 10-MW pressurized loop, in-pool-type, light-water-moderated, beryllium-reflected, flux trap reactor. MURR has a compact core (0.033 m 3 ) composed of eight fuel elements of the materials test reactor type arranged as an annular right circular cylinder between the inner and outer aluminum pressure vessels. Conservative engineering judgment resulted in the decision in 1988 to purchase new inner and outer pressure vessels. This paper details the difficulties encountered in procuring replacements for aluminum pressure vessels built to standards that are no longer applicable in attempting to meet nuclear standards that are not applicable to nonferrous material

  2. Fossil Energy Advanced Research and Technology Development Materials Program

    Energy Technology Data Exchange (ETDEWEB)

    Cole, N.C.; Judkins, R.R. (comps.)

    1992-12-01

    Objective of this materials program is to conduct R and D on materials for fossil energy applications with focus on longer-term and generic needs of the various fossil fuel technologies. The projects are organized according to materials research areas: (1) ceramics, (2) new alloys: iron aluminides, advanced austenitics and chromium niobium alloys, and (3) technology development and transfer. Separate abstracts have been prepared.

  3. Research and development of advanced materials using ion beam

    Energy Technology Data Exchange (ETDEWEB)

    Namba, Susumu [Nagasaki Inst. of Applied Science, Nagasaki (Japan)

    1997-03-01

    A wide range of research and development activities of advanced material synthesis using ion beams will be discussed, including ion beam applications to the state-of-the-art electronics from giant to nano electronics. (author)

  4. Fundamental Research into Hyperelastic Materials for Flight Applications (FY15)

    Data.gov (United States)

    National Aeronautics and Space Administration — This research project is working to develop methods to characterize elastomer materials for flight applications as well as instrumentation methods to monitor their...

  5. On-line repository of audiovisual material feminist research methodology

    Directory of Open Access Journals (Sweden)

    Lena Prado

    2014-12-01

    Full Text Available This paper includes a collection of audiovisual material available in the repository of the Interdisciplinary Seminar of Feminist Research Methodology SIMReF (http://www.simref.net.

  6. Film Music: The Material, Literature and Present State of Research.

    Science.gov (United States)

    Marks, Martin

    1982-01-01

    A comprehensive look at the neglected art of film music. Examines the nature of the medium, the literature (how others have wrestled with film music's recalcitrant materials), and the present state of research into film music. Includes a bibliography. (PD)

  7. Research Tools and Materials | NCI Technology Transfer Center | TTC

    Science.gov (United States)

    Research Tools can be found in TTC's Available Technologies and in scientific publications. They are freely available to non-profits and universities through a Material Transfer Agreement (or other appropriate mechanism), and available via licensing to companies.

  8. Materials and Molecular Research Division annual report 1980

    International Nuclear Information System (INIS)

    1981-06-01

    Progress made in the following research areas is reported: materials sciences (metallurgy and ceramics, solid state physics, materials chemistry); chemical sciences (fundamental interactions, processes and techniques); nuclear sciences; fossil energy; advanced isotope separation technology; energy storage; magnetic fusion energy; and nuclear waste management

  9. Materials and Molecular Research Division annual report 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-06-01

    Progress made in the following research areas is reported: materials sciences (metallurgy and ceramics, solid state physics, materials chemistry); chemical sciences (fundamental interactions, processes and techniques); nuclear sciences; fossil energy; advanced isotope separation technology; energy storage; magnetic fusion energy; and nuclear waste management.

  10. Materials research and development for nuclear weapons applications

    International Nuclear Information System (INIS)

    Anon.

    1977-01-01

    Highlights of a comprehensive summary of materials research and development being conducted at Sandia in support of the nuclear weapons development programs are presented. The developments include foams, encapsulants, metals with memories, material equations-of-state, composites, glass-to-metal bonds, and design processes

  11. Evolution of materials research within the AINSE portfolio

    International Nuclear Information System (INIS)

    Jostsons, A.

    1998-01-01

    Full text: The main materials research interactions between ANSTO/AAEC and the AINSE member universities are reviewed and linked to the main thrust of contemporary ANSTO/AAEC programs. The AINSE portfolio encompasses the previous AAEC research contracts, which represent an earlier example of public sector outsourcing, until re-discovered during the present decade, as well as AINSE studentships and Research and Training Projects. Collectively these mechanisms did much to foster the maintenance of effective materials research teams in Australian universities. Selective examples will illustrate the success of the AINSE family in training to help provide engineers and scientists of high ability for the future

  12. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    International Nuclear Information System (INIS)

    Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

    2012-01-01

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

  13. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  14. Japanese program of materials research for fusion reactors

    International Nuclear Information System (INIS)

    Hasiguti, R.R.

    1982-01-01

    The Japanese program of materials research for fusion reactors is described based on the report to the Nuclear Fusion Council, the project research program of the Ministry of Education, Science and Culture, and other official documents. The alloy development for the first wall and its radiation damage are the main topics discussed in this paper. Materials viewpoints for the Japanese Tokamak facilities and the problems of irradiation facilities are also discussed. (orig.)

  15. Summary of NRC LWR safety research programs on fuel behavior, metallurgy/materials and operational safety

    International Nuclear Information System (INIS)

    Bennett, G.L.

    1979-09-01

    The NRC light-water reactor safety-research program is part of the NRC regulatory program for ensuring the safety of nuclear power plants. This paper summarizes the results of NRC-sponsored research into fuel behavior, metallurgy and materials, and operational safety. The fuel behavior research program provides a detailed understanding of the response of nuclear fuel assemblies to postulated off-normal or accident conditions. Fuel behavior research includes studies of basic fuel rod properties, in-reactor tests, computer code development, fission product release and fuel meltdown. The metallurgy and materials research program provides independent confirmation of the safe design of reactor vessels and piping. This program includes studies on fracture mechanics, irradiation embrittlement, stress corrosion, crack growth, and nondestructive examination. The operational safety research provides direct assistance to NRC officials concerned with the operational and operational-safety aspects of nuclear power plants. The topics currently being addressed include qualification testing evaluation, fire protection, human factors, and noise diagnostics

  16. PREFACE: 7th EEIGM International Conference on Advanced Materials Research

    Science.gov (United States)

    Joffe, Roberts

    2013-12-01

    The 7th EEIGM Conference on Advanced Materials Research (AMR 2013) was held at Luleå University of Technology on the 21-22 March 2013 in Luleå, SWEDEN. This conference is intended as a meeting place for researchers involved in the EEIGM programme, in the 'Erasmus Mundus' Advanced Materials Science and Engineering Master programme (AMASE) and the 'Erasmus Mundus' Doctoral Programme in Materials Science and Engineering (DocMASE). This is great opportunity to present their on-going research in the various fields of Materials Science and Engineering, exchange ideas, strengthen co-operation as well as establish new contacts. More than 60 participants representing six countries attended the meeting, in total 26 oral talks and 19 posters were presented during two days. This issue of IOP Conference Series: Materials Science and Engineering presents a selection of articles from EEIGM-7 conference. Following tradition from previous EEIGM conferences, it represents the interdisciplinary nature of Materials Science and Engineering. The papers presented in this issue deal not only with basic research but also with applied problems of materials science. The presented topics include theoretical and experimental investigations on polymer composite materials (synthetic and bio-based), metallic materials and ceramics, as well as nano-materials of different kind. Special thanks should be directed to the senior staff of Division of Materials Science at LTU who agreed to review submitted papers and thus ensured high scientific level of content of this collection of papers. The following colleagues participated in the review process: Professor Lennart Walström, Professor Roberts Joffe, Professor Janis Varna, Associate Professor Marta-Lena Antti, Dr Esa Vuorinen, Professor Aji Mathew, Professor Alexander Soldatov, Dr Andrejs Purpurs, Dr Yvonne Aitomäki, Dr Robert Pederson. Roberts Joffe October 2013, Luleå Conference photograph EEIGM7 conference participants, 22 March 2013 The PDF

  17. Ministerial Order of 31 December 1982 made in implementation of Section 3(2)(a) of the Royal Order of 5 November 1982 on the training certificate for drivers of vehicles carrying radioactive materials in cisterns or vessels

    International Nuclear Information System (INIS)

    1983-01-01

    This Ministerial Order fixes the conditions for obtaining the ADR training certificate for driving transport vehicles which contain radioactive materials in cisterns and vessels. It is the second of four Orders made under the 1982 Royal Order. (NEA) [fr

  18. PERFORM 60: Prediction of the effects of radiation for reactor pressure vessel and in-core materials using multi-scale modelling - 60 years foreseen plant lifetime

    International Nuclear Information System (INIS)

    Al Mazouzi, A.; Alamo, A.; Lidbury, D.; Moinereau, D.; Van Dyck, S.

    2011-01-01

    Highlights: → Multi-scale and multi-physics modelling are adopted by PERFORM 60 to predict irradiation damage in nuclear structural materials. → PERFORM 60 allows to Consolidate the community and improve the interaction between universities/industries and safety authorities. → Experimental validation at the relevant scale is a key for developing the multi-scale modelling methodology. - Abstract: In nuclear power plants, materials undergo degradation due to severe irradiation conditions that may limit their operational lifetime. Utilities that operate these reactors need to quantify the ageing and potential degradation of certain essential structures of the power plant to ensure their safe and reliable operation. So far, the monitoring and mitigation of these degradation phenomena rely mainly on long-term irradiation programs in test reactors as well as on mechanical or corrosion testing in specialized hot cells. Continuous progress in the physical understanding of the phenomena involved in irradiation damage and progress in computer sciences have now made possible the development of multi-scale numerical tools able to simulate the materials behaviour in a nuclear environment. Indeed, within the PERFECT project of the EURATOM framework program (FP6), a first step has been successfully reached through the development of a simulation platform that contains several advanced numerical tools aiming at the prediction of irradiation damage in both the reactor pressure vessel (RPV) and its internals using available, state-of-the-art-knowledge. These tools allow simulation of irradiation effects on the nanostructure and the constitutive behaviour of the RPV low alloy steels, as well as their fracture mechanics properties. For the more highly irradiated reactor internals, which are commonly produced using austenitic stainless steels, the first partial models were established, describing radiation effects on the nanostructure and providing a first description of the

  19. Novel Fiber Optic Sensor Probe with a Pair of Highly Reflected Connectors and a Vessel of Water Absorption Material for Water Leak Detection

    Directory of Open Access Journals (Sweden)

    Tae-Sik Cho

    2012-08-01

    Full Text Available The use of a fiber optic quasi-distributed sensing technique for detecting the location and severity of water leakage is suggested. A novel fiber optic sensor probe is devised with a vessel of water absorption material called as water combination soil (WCS located between two highly reflected connectors: one is a reference connector and the other is a sensing connector. In this study, the sensing output is calculated from the reflected light signals of the two connectors. The first reflected light signal is a reference and the second is a sensing signal which is attenuated by the optical fiber bending loss due to the WCS expansion absorbing water. Also, the bending loss of each sensor probe is determined by referring to the total number of sensor probes and the total power budget of an entire system. We have investigated several probe characteristics to show the design feasibility of the novel fiber sensor probe. The effects of vessel sizes of the probes on the water detection sensitivity are studied. The largest vessel probe provides the highest sensitivity of 0.267 dB/mL, while the smallest shows relatively low sensitivity of 0.067 dB/mL, and unstable response. The sensor probe with a high output value provides a high sensitivity with various detection levels while the number of total installable sensor probes decreases.

  20. Novel fiber optic sensor probe with a pair of highly reflected connectors and a vessel of water absorption material for water leak detection.

    Science.gov (United States)

    Cho, Tae-Sik; Choi, Ki-Sun; Seo, Dae-Cheol; Kwon, Il-Bum; Lee, Jung-Ryul

    2012-01-01

    The use of a fiber optic quasi-distributed sensing technique for detecting the location and severity of water leakage is suggested. A novel fiber optic sensor probe is devised with a vessel of water absorption material called as water combination soil (WCS) located between two highly reflected connectors: one is a reference connector and the other is a sensing connector. In this study, the sensing output is calculated from the reflected light signals of the two connectors. The first reflected light signal is a reference and the second is a sensing signal which is attenuated by the optical fiber bending loss due to the WCS expansion absorbing water. Also, the bending loss of each sensor probe is determined by referring to the total number of sensor probes and the total power budget of an entire system. We have investigated several probe characteristics to show the design feasibility of the novel fiber sensor probe. The effects of vessel sizes of the probes on the water detection sensitivity are studied. The largest vessel probe provides the highest sensitivity of 0.267 dB/mL, while the smallest shows relatively low sensitivity of 0.067 dB/mL, and unstable response. The sensor probe with a high output value provides a high sensitivity with various detection levels while the number of total installable sensor probes decreases.

  1. Crystal Growth and Other Materials Physical Researches in Space Environment

    Science.gov (United States)

    Pan, Mingxiang

    Material science researches in space environment are based on reducing the effects of buoyancy driven transport, the effects of atomic oxygen, radiation, extremes of heat and cold and the ultrahigh vacuum, so as to unveil the underlying fundamental phenomena, lead maybe to new potential materials or new industrial processes and develop space techniques. Currently, research program on materials sciences in Chinese Manned Space Engineering (CMSE) is going on. More than ten projects related to crystal growth and materials processes are selected as candidates to be executed in Shenzhou spacecraft, Tiangong Space Laboratory and Chinese Space Station. In this talk, we will present some examples of the projects, which are being prepared and executed in the near future flight tasks. They are both basic and applied research, from discovery to technology.

  2. First Materials Science Research Rack Capabilities and Design Features

    Science.gov (United States)

    Schaefer, D.; King, R.; Cobb, S.; Whitaker, Ann F. (Technical Monitor)

    2001-01-01

    The first Materials Science Research Rack (MSRR-1) will accommodate dual Experiment Modules (EM's) and provide simultaneous on-orbit processing operations capability. The first international Materials Science Experiment Module for the MSRR-1 is an international cooperative research activity between NASA's Marshall Space Flight Center (MSFC) and the European Space Agency's (ESA) European Space Research and Technology Center. (ESTEC). This International Standard Payload Rack (ISPR) will contain the Materials Science Laboratory (MSL) developed by ESA as an Experiment Module. The MSL Experiment Module will accommodate several on-orbit exchangeable experiment-specific Module Inserts. Module Inserts currently planned are a Quench Module Insert, Low Gradient Furnace, Solidification with Quench Furnace, and Diffusion Module Insert. The second Experiment Module for the MSRR-1 configuration is a commercial device supplied by MSFC's Space Products Department (SPD). It includes capabilities for vapor transport processes and liquid metal sintering. This Experiment Module will be replaced on-orbit with other NASA Materials Science EMs.

  3. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    International Nuclear Information System (INIS)

    Duysen, J.C. van; Meric de Bellefon, G.

    2017-01-01

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  4. 60th Anniversary of electricity production from light water reactors: Historical review of the contribution of materials science to the safety of the pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Duysen, J.C. van, E-mail: jean-claude.van-duysen@ensc-lille.fr [Department of Nuclear Engineering University of Tennessee Knoxville (United States); Unité Matériaux et Transformation (UMET) CNRS, Université de Lille 1 (France); Meric de Bellefon, G., E-mail: mericdebelle@wisc.edu [Department of Nuclear Engineering, University of Wisconsin, Madison (United States)

    2017-02-15

    The first light water nuclear reactor dedicated to electricity production was commissioned in Shippingport, Pennsylvania in the United States in 1957. Sixty years after the event, it is clear that this type of reactor will be a major source of electricity and one of the key solutions to limit climate change in the 21st century. This article pays homage to the teams that contributed to this achievement by their involvement in research and development and their determination to push back the frontiers of knowledge. Via a few examples of scientific or technological milestones, it describes the evolution of ideas, models, and techniques during the last 60 years, and gives the current state-of-the-art in areas related to the safety of the reactor pressure vessel. Among other topics, it focuses on vessel manufacturing, steel fracture mechanics analysis, and understanding of irradiation-induced damage.

  5. Radioactive waste processing vessel

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  6. 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

  7. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective ...

  8. Assessment of research needs for wind turbine rotor materials technology

    National Research Council Canada - National Science Library

    National Research Council Staff; Commission on Engineering and Technical Systems; Division on Engineering and Physical Sciences; National Research Council; National Academy of Sciences

    1991-01-01

    ... on Assessment of Research Needs for Wind Turbine Rotor Materials Technology Energy Engineering Board Commission on Engineering and Technical Systems National Research Council NATIONAL ACADEMY PRESS Washington, D.C. 1991 Copyrightthe true use are Please breaks Page inserted. accidentally typesetting been have may original the from errors not...

  9. Life Science Research Facility materials management requirements and concepts

    Science.gov (United States)

    Johnson, Catherine C.

    1986-01-01

    The Advanced Programs Office at NASA Ames Research Center has defined hypothetical experiments for a 90-day mission on Space Station to allow analysis of the materials necessary to conduct the experiments and to assess the impact on waste processing of recyclable materials and storage requirements of samples to be returned to earth for analysis as well as of nonrecyclable materials. The materials include the specimens themselves, the food, water, and gases necessary to maintain them, the expendables necessary to conduct the experiments, and the metabolic products of the specimens. This study defines the volumes, flow rates, and states of these materials. Process concepts for materials handling will include a cage cleaner, trash compactor, biological stabilizer, and various recycling devices.

  10. General problems specific to hot nuclear materials research facilities

    International Nuclear Information System (INIS)

    Bart, G.

    1996-01-01

    During the sixties, governments have installed hot nuclear materials research facilities to characterize highly radioactive materials, to describe their in-pile behaviour, to develop and test new reactor core components, and to provide the industry with radioisotopes. Since then, the attitude towards the nuclear option has drastically changed and resources have become very tight. Within the changed political environment, the national research centres have defined new objectives. Given budgetary constraints, nuclear facilities have to co-operate internationally and to look for third party research assignments. The paper discusses the problems and needs within experimental nuclear research facilities as well as industrial requirements. Special emphasis is on cultural topics (definition of the scope of nuclear research facilities, the search for competitive advantages, and operational requirements), social aspects (overageing of personnel, recruitment, and training of new staff), safety related administrative and technical issues, and research needs for expertise and state of the art analytical infrastructure

  11. Materials Science Research Rack Onboard the International Space Station

    Science.gov (United States)

    Reagan, Shawn; Frazier, Natalie; Lehman, John

    2016-01-01

    The Materials Science Research Rack (MSRR) is a research facility developed under a cooperative research agreement between NASA and ESA for materials science investigations on the International Space Station (ISS). MSRR was launched on STS-128 in August 2009 and currently resides in the U.S. Destiny Laboratory Module. Since that time, MSRR has logged more than 1400 hours of operating time. The MSRR accommodates advanced investigations in the microgravity environment on the ISS for basic materials science research in areas such as solidification of metals and alloys. The purpose is to advance the scientific understanding of materials processing as affected by microgravity and to gain insight into the physical behavior of materials processing. MSRR allows for the study of a variety of materials, including metals, ceramics, semiconductor crystals, and glasses. Materials science research benefits from the microgravity environment of space, where the researcher can better isolate chemical and thermal properties of materials from the effects of gravity. With this knowledge, reliable predictions can be made about the conditions required on Earth to achieve improved materials. MSRR is a highly automated facility with a modular design capable of supporting multiple types of investigations. The NASA-provided Rack Support Subsystem provides services (power, thermal control, vacuum access, and command and data handling) to the ESA-developed Materials Science Laboratory (MSL) that accommodates interchangeable Furnace Inserts (FI). Two ESA-developed FIs are presently available on the ISS: the Low Gradient Furnace (LGF) and the Solidification and Quenching Furnace (SQF). Sample Cartridge Assemblies (SCAs), each containing one or more material samples, are installed in the FI by the crew and can be processed at temperatures up to 1400?C. ESA continues to develop samples with 14 planned for launch and processing in the near future. Additionally NASA has begun developing SCAs to

  12. Molecularly Engineered Energy Materials, an Energy Frontier Research Center

    Energy Technology Data Exchange (ETDEWEB)

    Ozolins, Vidvuds [Univ. of California, Los Angeles, CA (United States). Materials Science and Engineering Dept.

    2016-09-28

    Molecularly Engineered Energy Materials (MEEM) was established as an interdisciplinary cutting-edge UCLA-based research center uniquely equipped to attack the challenge of rationally designing, synthesizing and testing revolutionary new energy materials. Our mission was to achieve transformational improvements in the performance of materials via controlling the nano-and mesoscale structure using selectively designed, earth-abundant, inexpensive molecular building blocks. MEEM has focused on materials that are inherently abundant, can be easily assembled from intelligently designed building blocks (molecules, nanoparticles), and have the potential to deliver transformative economic benefits in comparison with the current crystalline-and polycrystalline-based energy technologies. MEEM addressed basic science issues related to the fundamental mechanisms of carrier generation, energy conversion, as well as transport and storage of charge and mass in tunable, architectonically complex materials. Fundamental understanding of these processes will enable rational design, efficient synthesis and effective deployment of novel three-dimensional material architectures for energy applications. Three interrelated research directions were initially identified where these novel architectures hold great promise for high-reward research: solar energy generation, electrochemical energy storage, and materials for CO2 capture. Of these, the first two remained throughout the project performance period, while carbon capture was been phased out in consultation and with approval from BES program manager.

  13. PREFACE: 6th EEIGM International Conference on Advanced Materials Research

    Science.gov (United States)

    Horwat, David; Ayadi, Zoubir; Jamart, Brigitte

    2012-02-01

    The 6th EEIGM Conference on Advanced Materials Research (AMR 2011) was held at the European School of Materials Engineering (EEIGM) on the 7-8 November 2011 in Nancy, France. This biennial conference organized by the EEIGM is a wonderful opportunity for all scientists involved in the EEIGM programme, in the 'Erasmus Mundus' Advanced Materials Science and Engineering Master programme (AMASE) and the 'Erasmus Mundus' Doctoral Programme in Materials Science and Engineering (DocMASE), to present their research in the various fields of Materials Science and Engineering. This conference is also open to other universities who have strong links with the EEIGM and provides a forum for the exchange of ideas, co-operation and future orientations by means of regular presentations, posters and a round-table discussion. This edition of the conference included a round-table discussion on composite materials within the Interreg IVA project '+Composite'. Following the publication of the proceedings of AMR 2009 in Volume 5 of this journal, it is with great pleasure that we present this selection of articles to the readers of IOP Conference Series: Materials Science and Engineering. Once again it represents the interdisciplinary nature of Materials Science and Engineering, covering basic and applicative research on organic and composite materials, metallic materials and ceramics, and characterization methods. The editors are indebted to all the reviewers for reviewing the papers at very short notice. Special thanks are offered to the sponsors of the conference including EEIGM-Université de Lorraine, AMASE, DocMASE, Grand Nancy, Ville de Nancy, Region Lorraine, Fédération Jacques Villermaux, Conseil Général de Meurthe et Moselle, Casden and '+Composite'. Zoubir Ayadi, David Horwat and Brigitte Jamart

  14. Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

    International Nuclear Information System (INIS)

    Vlasichev, G.N.

    1994-01-01

    Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

  15. User's manual and analysis methodology of probabilistic fracture mechanics analysis code PASCAL Ver.2 for reactor pressure vessel (Contract research)

    International Nuclear Information System (INIS)

    Osakabe, Kazuya; Onizawa, Kunio; Shibata, Katsuyuki; Kato, Daisuke

    2006-09-01

    As a part of the aging structural integrity research for LWR components, the probabilistic fracture mechanics (PFM) analysis code PASCAL (PFM Analysis of Structural Components in Aging LWR) has been developed in JAEA. This code evaluates the conditional probabilities of crack initiation and fracture of a reactor pressure vessel (RPV) under transient conditions such as pressurized thermal shock (PTS). The development of the code has been aimed to improve the accuracy and reliability of analysis by introducing new analysis methodologies and algorithms considering the recent development in the fracture mechanics and computer performance. PASCAL Ver.1 has functions of optimized sampling in the stratified Monte Carlo simulation, elastic-plastic fracture criterion of the R6 method, crack growth analysis models for a semi-elliptical crack, recovery of fracture toughness due to thermal annealing and so on. Since then, under the contract between the Ministry of Economy, Trading and Industry of Japan and JAEA, we have continued to develop and introduce new functions into PASCAL Ver.2 such as the evaluation method for an embedded crack, K I database for a semi-elliptical crack considering stress discontinuity at the base/cladding interface, PTS transient database, and others. A generalized analysis method is proposed on the basis of the development of PASCAL Ver.2 and results of sensitivity analyses. Graphical user interface (GUI) including a generalized method as default values has been also developed for PASCAL Ver.2. This report provides the user's manual and theoretical background of PASCAL Ver.2. (author)

  16. Development of outcome measures for large-vessel vasculitis for use in clinical trials: opportunities, challenges, and research agenda.

    Science.gov (United States)

    Direskeneli, Haner; Aydin, Sibel Z; Kermani, Tanaz A; Matteson, Eric L; Boers, Maarten; Herlyn, Karen; Luqmani, Raashid A; Neogi, Tuhina; Seo, Philip; Suppiah, Ravi; Tomasson, Gunnar; Merkel, Peter A

    2011-07-01

    Giant cell (GCA) and Takayasu's arteritis (TAK) are 2 forms of large-vessel vasculitis (LVV) that involve the aorta and its major branches. GCA has a predilection for the cranial branches, while TAK tends to affect the extracranial branches. Both disorders may also cause nonspecific constitutional symptoms. Although some clinical features are more common in one or the other disorder and the ages of initial presentation differ substantially, there is enough clinical and histopathologic overlap between these disorders that some investigators suggest GCA and TAK may be 2 processes within the spectrum of a single disease. There have been few randomized therapeutic trials completed in GCA, and none in TAK. The lack of therapeutic trials in LVV is only partially explained by the rarity of these diseases. It is likely that the lack of well validated outcome measures for LVV and uncertainties regarding trial design contribute to the paucity of trials for these diseases. An initiative to develop a core set of outcome measures for use in clinical trials of LVV was launched by the international OMERACT Vasculitis Working Group in 2009 and subsequently endorsed by the OMERACT community at the OMERACT 10 meeting. Aims of this initiative include: (1) to review the literature and existing data related to outcome assessments in LVV; (2) to obtain the opinion of experts and patients on disease content; and (3) to formulate a research agenda to facilitate a more data-based approach to outcomes development.

  17. Testing a polarimetric cloud imager aboard research vessel Polarstern: comparison of color-based and polarimetric cloud detection algorithms.

    Science.gov (United States)

    Barta, András; Horváth, Gábor; Horváth, Ákos; Egri, Ádám; Blahó, Miklós; Barta, Pál; Bumke, Karl; Macke, Andreas

    2015-02-10

    Cloud cover estimation is an important part of routine meteorological observations. Cloudiness measurements are used in climate model evaluation, nowcasting solar radiation, parameterizing the fluctuations of sea surface insolation, and building energy transfer models of the atmosphere. Currently, the most widespread ground-based method to measure cloudiness is based on analyzing the unpolarized intensity and color distribution of the sky obtained by digital cameras. As a new approach, we propose that cloud detection can be aided by the additional use of skylight polarization measured by 180° field-of-view imaging polarimetry. In the fall of 2010, we tested such a novel polarimetric cloud detector aboard the research vessel Polarstern during expedition ANT-XXVII/1. One of our goals was to test the durability of the measurement hardware under the extreme conditions of a trans-Atlantic cruise. Here, we describe the instrument and compare the results of several different cloud detection algorithms, some conventional and some newly developed. We also discuss the weaknesses of our design and its possible improvements. The comparison with cloud detection algorithms developed for traditional nonpolarimetric full-sky imagers allowed us to evaluate the added value of polarimetric quantities. We found that (1) neural-network-based algorithms perform the best among the investigated schemes and (2) global information (the mean and variance of intensity), nonoptical information (e.g., sun-view geometry), and polarimetric information (e.g., the degree of polarization) improve the accuracy of cloud detection, albeit slightly.

  18. Materials and Components Technology Division research summary, 1992

    International Nuclear Information System (INIS)

    1992-11-01

    The Materials and Components Technology Division (MCT) provides a research and development capability for the design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs related to nuclear energy support the development of the Integral Fast Reactor (IFR): life extension and accident analyses for light water reactors (LWRs); fuels development for research and test reactors; fusion reactor first-wall and blanket technology; and safe shipment of hazardous materials. MCT Conservation and Renewables programs include major efforts in high-temperature superconductivity, tribology, nondestructive evaluation (NDE), and thermal sciences. Fossil Energy Programs in MCT include materials development, NDE technology, and Instrumentation design. The division also has a complementary instrumentation effort in support of Arms Control Technology. Individual abstracts have been prepared for the database

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

    International Nuclear Information System (INIS)

    1997-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

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

  1. QUALITATIVE FACTORS OF MATERIALITY - A REVIEW OF EMPIRICAL RESEARCH

    OpenAIRE

    Emil Irimie Popa; Ancuta Georgeta Span; Timea Melinda Fulop

    2010-01-01

    Determination of materiality is a crucial step in an audit mission because it affects theentire audit process. The incorrect application of materiality can have serious negativerepercussions on both the audited entity and the auditor (Enron-Anderson). Researches conductedover time revealed the complexity of this element in an audit mission and the need to emergencesome generally accepted rules and regulations to provide support in defining and substantiating theprofessional judgment applied i...

  2. Research Note: Inside an Indonesian Online Library for Radical Materials

    Directory of Open Access Journals (Sweden)

    Muhammad Haniff Hassan

    2012-12-01

    Full Text Available This Research Note provides a review of an Indonesian online library for radical materials. The objective of this review is to compile data and information that will contribute to the understanding of the online radicalisation phenomenon as well as the extremists themselves. Based on data found on the online library, this Research Note reports findings on the influence of Al-Maqdisi’s website; the emphasis on translation work of Arabic materials to Indonesian language by radicals and the value of Arabic materials to them. It also covers influential thinkers and ideologues and the use of the Wikipedia modus operandi to hasten the development of the website and effect mobilisation and recruitment, among others things. Based on the data found, this Research Note concludes that ideas matter to radicals.

  3. Heat dissipation research on the water-cooling channel of HL-2M in-vessel coils

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J., E-mail: jiangjiaming@swip.ac.cn; Liu, Y.; Chen, Q.; Ji, X.Q.

    2017-04-15

    Highlights: • The joule heat of in-vessel coils is very difficult to dissipate inside HL-2M vacuum vessel. • Heat dissipation model of the coil includes the joule heat model, the heat conduction model and the heat transfer model. • The CFD analysis has been done for the coil-water cooling, with comparison with the date of theoretical analysis and experiment. • The result shows water-cooling channel is good for the joule heat transfer and taken away. - Abstract: HL-2M in-vessel coils are positioned in high vacuum circumstance, and they will generate joule heat when they carry 15 kA electrical current, but joule heat is very difficult to dissipate in vacuum, so a hollow cable with 8 mm inner diameter is design as water-cooling channel for heat convection. By using the methods of the theoretical derivation, together with CFD numeric simulation method and the experiment of the heat transfer, the water channel of HL-2M in-vessel coils has been studied, and the temperature of HL-2M in-vessel coils under different cooling water flow rates is obtained and acceptable. Simultaneously, the external cooling water supply system parameters for the water-cooling channel of the coils are estimated. Three methods’ results are in good agreement; the theoretical model is verified and could be popularized for predicting the temperature rise of HL-2M in-vessel coils.

  4. Materials and Components Technology Division research summary, 1991

    International Nuclear Information System (INIS)

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base

  5. NATO Advanced Research Workshop on New Materials for Thermoelectric Applications

    CERN Document Server

    Hewson, Alex

    2013-01-01

    Thermoelectric devices could play an important role in making efficient use of our energy resources but their efficiency would need to be increased for their wide scale application. There is a multidisciplinary search for materials with an enhanced thermoelectric responses for use in such devices. This volume covers the latest ideas and developments in this research field, covering topics ranging from the fabrication and characterization of new materials, particularly those with strong electron correlation, use of nanostructured, layered materials and composites, through to theoretical work to gain a deeper understanding of thermoelectric behavior. It should be a useful guide and stimulus to all working in this very topical field.

  6. Accelerator-driven neutron sources for materials research

    International Nuclear Information System (INIS)

    Jameson, R.A.

    1990-01-01

    Particle accelerators are important tools for materials research and production. Advances in high-intensity linear accelerator technology make it possible to consider enhanced neutron sources for fusion material studies or as a source of spallation neutrons. Energy variability, uniformity of target dose distribution, target bombardment from multiple directions, time-scheduled dose patterns, and other features can be provided, opening new experimental opportunities. New designs have also been used to ensure hands-on maintenance on the accelerator in these factory-type facilities. Designs suitable for proposals such as the Japanese Energy-Selective Intense Neutron Source, and the international Fusion Materials Irradiation Facility are discussed

  7. Materials and Components Technology Division research summary, 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-04-01

    This division has the purpose of providing a R and D capability for design, fabrication, and testing of high-reliability materials, components, and instrumentation. Current divisional programs are in support of the Integral Fast Reactor, life extension for light water reactors, fuels development for the new production reactor and research and test reactors, fusion reactor first-wall and blanket technology, safe shipment of hazardous materials, fluid mechanics/materials/instrumentation for fossile energy systems, and energy conservation and renewables (including tribology, high- temperature superconductivity). Separate abstracts have been prepared for the data base.

  8. Application of ion scattering spectrometers for the observation of process of cleaning of surfaces of materials for vacuum vessel

    International Nuclear Information System (INIS)

    Akashi, Ken-ya; Miyahara, Akira; Sagara, Akio.

    1978-01-01

    The impurity gas emitted from the surfaces of vacuum vessels was investigated by using the shadowing effect of the covering atoms. The ion scattering spectrometer used for the experiment consists of an ion source, a test sample, an energy analyzer and an ion detector. The evacuation system comprises a turbomolecular pump, a Ti-sublimation pump and an ion pump. The achieved final gas pressure is 5 x 10 -10 Torr. The ion beam intensity to a sample is 10 micro ampere/cm 2 , and the ion energy is about 1 to 1.5 keV. The quantity of oxygen on the surface of a sample molybdenum was measured in the process of evacuation. The concentration of surface oxygen decreased with the gas pressure of the system. It was found that residual oxygen was observed after the sputter etching with Ar ion impact on the surface. The reason of this residual oxygen was considered. (Kato, T.)

  9. Final IAEA research coordination meeting on plasma-interaction induced erosion of fusion reactor materials. October 9-11, 1995, Vienna, Austria. Summary report

    International Nuclear Information System (INIS)

    Langley, R.A.

    1995-12-01

    The proceedings and results of the Final IAEA Research Coordination Meeting on ''Plasma-interaction Induced Erosion of Fusion Reactor Materials'' held on October 9, 10 and 11, 1995 at the IAEA Headquarters in Vienna are briefly described. This report includes a summary of presentations made by the meeting participants, the results of a data survey and needs assessment for the erosion of plasma facing components and in-vessel materials, and recommendations regarding future work. (author). Refs, figs, tabs

  10. Is there a future for material fatigue research?

    International Nuclear Information System (INIS)

    Joosse, P.; Bulder, B.

    1996-01-01

    Due to the fact that it is quite difficult to get new funding for (fundamental) wind turbine material related fatigue research the authors started a discussion with the following title: Are there still wind turbine engineering specific fatigue problems? and What are the research goals for the fatigue experts in wind engineering for the second half of the 90 ies . In this paper the present status of the fatigue issue and the discussion following is reported. (au)

  11. Is there a future for material fatigue research?

    Energy Technology Data Exchange (ETDEWEB)

    Joosse, P [Stork Product Engineering b.v., Amsterdam (Netherlands); Bulder, B [ECN-Renewable Energy, Petten (Netherlands)

    1996-09-01

    Due to the fact that it is quite difficult to get new funding for (fundamental) wind turbine material related fatigue research the authors started a discussion with the following title: Are there still wind turbine engineering specific fatigue problems? and What are the research goals for the fatigue experts in wind engineering for the second half of the 90{sup ies}. In this paper the present status of the fatigue issue and the discussion following is reported. (au)

  12. Computational micromechanics of wind blade materials: recent activities at the Materials Research Division, Risoe DTU

    Energy Technology Data Exchange (ETDEWEB)

    Mishnaevsky Jr., L.; Broendsted, P.; Qing, H.; Wang, H.; Soerensen, Bent F. (Technical Univ. of Denmark, Riso National Lab. for Sustainable Energy. Materials Research Div., Roskilde (Denmark)); OEstergaard, R.C. (LM Wind Power Blades, Composite Mechanics, Roskilde (Denmark))

    2010-10-22

    Recent research works in the area of 3D computational microstructural modelling, virtual testing and numerical optimization of wind blade materials, carried out at the Materials Research Division, Rise DTU (Programme Composites and Materials Mechanics) are summarized. The works presented here have been carried out in the framework of several research projects: EU FP6 Upwind, Danida project 'Development of wind energy technologies in Nepal' and SinoDanish project '3D Virtual Testing of composites for wind energy applications' as well as the Framework Program 'Interface design of composite materials' and recently established Danish Centre for Composite Structures and Materials for Wind Turbines. Different groups of materials, which are used or have a potential for use for the wind turbine blades, are modelled with the use of the methods of the computational micromechanics, in particular: (1) glass and carbon fiber reinforced polymer composites used in the large wind turbine blades, (2) different sorts of timber, used in small wind turbines (first of all, in developing countries) and (3) nanoparticle reinforced polymer matrix composites (which have a potential to be used as components for future high strength wind blades). On the basis of the developed 3D microstructural finite element models of these materials, we analyzed the effect of their microstructures on damage resistance, strength and stiffness. The methods of the 3D model design and results of the simulations are discussed in this paper. (Author)

  13. Development of an Extreme Environment Materials Research Facility at Princeton

    International Nuclear Information System (INIS)

    Cohen, A.B.; Gentile, C.A.; Tully, C.G.; Austin, R.; Calaprice, F.; McDonald, K.; Ascione, G.; Baker, G.; Davidson, R.; Dudek, L.; Grisham, L.; Kugel, H.; Pagdon, K.; Stevenson, T.; Woolley, R.; Zwicker, A.

    2010-01-01

    The need for a fundamental understanding of material response to a neutron and/or high heat flux environment can yield development of improved materials and operations with existing materials. Such understanding has numerous applications in fields such as nuclear power (for the current fleet and future fission and fusion reactors), aerospace, and other research fields (e.g., high-intensity proton accelerator facilities for high energy physics research). A proposal has been advanced to develop a facility for testing various materials under extreme heat and neutron exposure conditions at Princeton. The Extreme Environment Materials Research Facility comprises an environmentally controlled chamber (48 m 3 ) capable of high vacuum conditions, with extreme flux beams and probe beams accessing a central, large volume target. The facility will have the capability to expose large surface areas (1 m 2 ) to 14 MeV neutrons at a fluence in excess of 10 13 n/s. Depending on the operating mode. Additionally beam line power on the order of 15-75 MW/m 2 for durations of 1-15 seconds are planned. The multi-second duration of exposure can be repeated every 2-10 minutes for periods of 10-12 hours. The facility will be housed in the test cell that held the Tokamak Fusion Test Reactor (TFTR), which has the desired radiation and safety controls as well as the necessary loading and assembly infrastructure. The facility will allow testing of various materials to their physical limit of thermal endurance and allow for exploring the interplay between radiation-induced embrittlement, swelling and deformation of materials, and the fatigue and fracturing that occur in response to thermal shocks. The combination of high neutron energies and intense fluences will enable accelerated time scale studies. The results will make contributions for refining predictive failure modes (modeling) in extreme environments, as well as providing a technical platform for the development of new alloys, new

  14. RUPS: Research Utilizing Problem Solving. Administrators Version. Participant Materials.

    Science.gov (United States)

    Jung, Charles; And Others

    These materials are the handouts for school administrators participating in RUPS (Research Utilizing Problem Solving) workshops. The purposes of the workshops are to develop skills for improving schools and to increase teamwork skills. The handouts correspond to the 16 subsets that make up the five-day workshop: (1) orientation; (2) identifying…

  15. Action Research to Support Teachers' Classroom Materials Development

    Science.gov (United States)

    Edwards, Emily; Burns, Anne

    2016-01-01

    Language teachers constantly create, adapt and evaluate classroom materials to develop new curricula and meet their learners' needs. It has long been argued (e.g. by Stenhouse, L. [1975]. "An Introduction to Curriculum Research and Development." London: Heinemann) that teachers themselves, as opposed to managers or course book writers,…

  16. 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.)

  17. [Research on the aging of all-ceramics restoration materials].

    Science.gov (United States)

    Zhang, Dongjiao; Chen, Xinmin

    2011-10-01

    All-ceramic crowns and bridges have been widely used for dental restorations owing to their excellent functionality, aesthetics and biocompatibility. However, the premature clinical failure of all-ceramic crowns and bridges may easily occur when they are subjected to the complex environment of oral cavity. In the oral environment, all-ceramic materials are prone to aging. Aging can lead all-ceramic materials to change color, to lower bending strength, and to reduce anti-fracture toughness. There are many factors affecting the aging of the all-ceramic materials, for example, the grain size, the type of stabilizer, the residual stress and the water environment. In order to analyze the aging behavior, to optimize the design of all-ceramic crowns and bridges, and to evaluate the reliability and durability, we review in this paper recent research progress of aging behavior for all-ceramics restoration materials.

  18. Developing a Computational Environment for Coupling MOR Data, Maps, and Models: The Virtual Research Vessel (VRV) Prototype

    Science.gov (United States)

    Wright, D. J.; O'Dea, E.; Cushing, J. B.; Cuny, J. E.; Toomey, D. R.; Hackett, K.; Tikekar, R.

    2001-12-01

    The East Pacific Rise (EPR) from 9-10deg. N is currently our best-studied section of fast-spreading mid-ocean ridge. During several decades of investigation it has been explored by the full spectrum of ridge investigators, including chemists, biologists, geologists and geophysicists. These studies, and those that are ongoing, provide a wealth of observational data, results and data-driven theoretical (often numerical) studies that have not yet been fully utilized either by research scientists or by professional educators. While the situation is improving, a large amount of data, results, and related theoretical models still exist either in an inert, non-interactive form (e.g., journal publications) or as unlinked and currently incompatible computer data or algorithms. Infrastructure is needed not just for ready access to data, but linkage of disparate data sets (data to data) as well as data to models in order quantitatively evaluate hypotheses, refine numerical simulations, and explore new relations between observables. The prototype of a computational environment and toolset, called the Virtual Research Vessel (VRV), is being developed to provide scientists and educators with ready access to data, results and numerical models. While this effort is focused on the EPR 9N region, the resulting software tools and infrastructure should be helpful in establishing similar systems for other sections of the global mid-ocean ridge. Work in progress includes efforts to develop: (1) virtual database to incorporate diverse data types with domain-specific metadata into a global schema that allows web-query across different marine geology data sets, and an analogous declarative (database available) description of tools and models; (2) the ability to move data between GIS and the above DBMS, and tools to encourage data submission to archivesl (3) tools for finding and viewing archives, and translating between formats; (4) support for "computational steering" (tool composition

  19. Role of radiation embrittlement in reactor vessel integrity assessment

    International Nuclear Information System (INIS)

    Marston, T.U.; Chexal, V.K.; Wyckoff, M.

    1982-01-01

    Reactor vessel integrity calculations are complex. The effect of radiation embrittlement on vessel material properties is a very important aspect of any vessel integrity evaluation. The importance of realistic (based on surveillance capsule results) rather than conservative estimates of the material properties (based on regulatory curves) cannot be overestimated. It is also important to make realistic thermal hydraulic and system operations assumptions. In addition, use of actual flaw sizes from in-service inspections (versus hypothetical flaw size selection) will promote realism. Important research results exist that need to be incorporated into the regulatory process. The authors believe results from current research and development efforts will demonstrate that, with reasonable assumptions and best estimate calculations, the safety of even the older reactor vessels with high copper content welds can be assured over their design lifetimes without the need for major fixes. The utilities, through EPRI and the vendors, have dedicated a significant effort to solving the pressurized thermal shock problem

  20. Casting materials and their application in research and teaching.

    Science.gov (United States)

    Haenssgen, Kati; Makanya, Andrew N; Djonov, Valentin

    2014-04-01

    From a biological point of view, casting refers to filling of anatomical and/or pathological spaces with extraneous material that reproduces a three-dimensional replica of the space. Casting may be accompanied by additional procedures such as corrosion, in which the soft tissue is digested out, leaving a clean cast, or the material may be mixed with radiopaque substances to allow x-ray photography or micro computed topography (µCT) scanning. Alternatively, clearing of the surrounding soft tissue increases transparency and allows visualization of the casted cavities. Combination of casting with tissue fixation allows anatomical dissection and didactic surgical procedures on the tissue. Casting materials fall into three categories namely, aqueous substances (India ink, Prussian blue ink), pliable materials (gelatins, latex, and silicone rubber), or hard materials (methyl methacrylates, polyurethanes, polyesters, and epoxy resins). Casting has proved invaluable in both teaching and research and many phenomenal biological processes have been discovered through casting. The choice of a particular material depends inter alia on the targeted use and the intended subsequent investigative procedures, such as dissection, microscopy, or µCT. The casting material needs to be pliable where anatomical and surgical manipulations are intended, and capillary-passable for ultrastructural investigations.

  1. Thermoelectric materials -- New directions and approaches. Materials Research Society symposium proceedings, Volume 478

    Energy Technology Data Exchange (ETDEWEB)

    Tritt, T M; Kanatzidis, M G; Lyon, Jr, H B; Mahan, G D [eds.

    1997-07-01

    Thermoelectric materials are utilized in a wide variety of applications related to solid-state refrigeration or small-scale power generation. Thermoelectric cooling is an environmentally friendly method of small-scale cooling in specific applications such as cooling computer chips and laser diodes. Thermoelectric materials are used in a wide range of applications from beverage coolers to power generation for deep-space probes such as the Voyager missions. Over the past thirty years, alloys based on the Bi-Te systems {l{underscore}brace}(Bi{sub 1{minus}x}Sb{sub x}){sub 2} (Te{sub 1{minus}x}Se{sub x}){sub 3}{r{underscore}brace} and Si{sub 1{minus}x}Ge{sub x} systems have been extensively studied and optimized for their use as thermoelectric materials to perform a variety of solid-state thermoelectric refrigeration and power generation tasks. Despite this extensive investigation of the traditional thermoelectric materials, there is still a substantial need and room for improvement, and thus, entirely new classes of compounds will have to be investigated. Over the past two-to-three years, research in the field of thermoelectric materials has been undergoing a rapid rebirth. The enhanced interest in better thermoelectric materials has been driven by the need for much higher performance and new temperature regimes for thermoelectric devices in many applications. The essence of a good thermoelectric is given by the determination of the material's dimensionless figure of merit, ZT = ({alpha}{sup 2}{sigma}/{lambda})T, where {alpha} is the Seebeck coefficient, {sigma} the electrical conductivity and {lambda} the total thermal conductivity. The best thermoelectric materials have a value of ZT = 1. This ZT = 1 has been an upper limit for more than 30 years, yet no theoretical or thermodynamic reason exits for why it can not be larger. The focus of the symposium is embodied in the title, Thermoelectric Materials: New Directions and Approaches. Many of the researchers in the

  2. General principles of researching the lexicon of traditional material culture

    Directory of Open Access Journals (Sweden)

    Nedeljkov Ljiljana

    2009-01-01

    Full Text Available The paper discusses a linguistic research of terminological systems connected with basic fields of human life and work which, in modern conditions, are either transformed into contemporary modern forms or gradually disappear due to changes in the way of life and work. The lexicon of material culture of native inhabitants of Vojvodina is examined, resulting in monographs on the terminologies of fishing, cartwrighting, shepherding and houses and furniture, all of which have in common the fact that the starting point was the research of the lexicon in question by semantic fields. The paper shows the lexicological and lexicographical procedures used while researching these terminological systems.

  3. Use of Reactor Pressure Vessel Surveillance Materials for Extended Life Evaluations Using Power and Test Reactor Irradiations

    International Nuclear Information System (INIS)

    Server, W.L.; Nanstad, R.K.; Odette, G.R.

    2012-01-01

    The most important component in assuring safety of the nuclear power plant is the reactor pressure (RPV). Surveillance programs have been designed to cover the licensed life of operating nuclear RPVs. The original surveillance programs were designed when the licensed life was 40 years. More than one-half of the operating nuclear plants in the USA have an extended license out to 60 years, and there are plans to continue to operate many plants out to 80 years. Therefore, the surveillance programs have had to be adjusted or enhanced to generate key data for 60 years, and now consideration must be given for 80 or more years. To generate the necessary data to assure safe operation out to these extended license lives, test reactor irradiations have been initiated with key RPV and model alloy steels, which include several steels irradiated in the current power reactor surveillance programs out to relatively high fluence levels. These data are crucial in understanding the radiation embrittlement mechanisms and to enable extrapolation of the irradiation effects on mechanical properties for these extended time periods. This paper describes the potential radiation embrittlement mechanisms and effects when assessing much longer operating times and higher neutron fluence levels. Potential methods for adjusting higher neutron flux test reactor data for use in predicting power reactor vessel conditions are discussed. (author)

  4. Radiation resistance of concrete of nuclear reactor vessel

    International Nuclear Information System (INIS)

    Belyakov, V.V.; Denisov, A.V.; Korenevskij, V.V.; Muzalevskij, L.P.; Dubrovskij, V.B.; Ivanov, D.A.; Nazarov, I.L.; Sashin, N.L.

    1992-01-01

    Results of calculational-experimental determination of radiation resistance for concrete bases on limestone gravel and quartz sand, which are the most perspective materials for manufacturing prestressed concrete of the VG-400 reactor vessel are considered. Material samples under investigation were irradiated in the channels of the IBR-2 research reactor for the purpose of the calcultional result verification

  5. The Materiality of Exclusion and the Ideology of Research

    DEFF Research Database (Denmark)

    Pais, Alexandre

    2013-01-01

    of resources, teacher formation, mathematical content for social justice, etc.). In this paper I shall argue that such dissemination of the problem of inequity disavows its materiality. Mathematics education as a research field will be used to illustrate how postmodern moves in educational research, and its...... the contemporary reading of Lacan made by Žižek, to show how the real of schools—that is, its materiality, the fact that schools are economical places—has to be repressed by existing postmodern educational research—and its emphasis on discourse and identity politics— in order for research to be possible...... education, valorization of different cultures, useful and critical mathematics. However, an ideology critique sees these obstacles as symptomatic points which allow one to grasp the political and economical relevance of mathematics in the school system. Baldino, R., & Cabral, T. (2006). Inclusion...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  7. Research and development of spent-fuel shipping casks and the criteria for sea-going vessels carrying them

    International Nuclear Information System (INIS)

    Aoki, S.; Ando, Y.

    1977-01-01

    Since the transport of spent fuel will increase rapidly and extensively in the near future, the Japanese Atomic Energy Committee enacted the Technical Standard for Transportation of Radioactive Materials, based on the IAEA Regulation for the Safe Transport of Radioactive Materials, 1973 Revised Edition. The authorities concerned have begun to review the former ordinances for transporting radioactive materials and to develop a unified system of relevant laws and standards. For ten years the Atomic Energy Bureau has invested in research and development to obtain data for the design and licence of a spent-fuel shipping cask. Different scale models of a prototype weighing 80t were used to clarify the scale effect of drop, puncture and fire tests, which are a feature of Japanese research and development. Also an immersion test in water at pressures up to about 500 bar is now carried out to investigate the integrity of the cask body and sealing structure to prevent leakage of radioactive contents to the surroundings should the cask fall into deep sea. In Japan, depending on the site of nuclear plants, almost all transport of unirradiated and spent fuels is by sea. Therefore, to secure safe transport, the design criteria of ships for the exclusive transport of spent-fuel shipping casks, namely full-load shipping, have been enacted, which aim to make the likelihood of sinking on collision, stranding, and other unforeseen accidents at sea highly improbable and also to keep radiation exposure of the crew as low as possible. (author)

  8. Stress analysis of liners for prestressed concrete reactor pressure vessels with regard to non-linear behaviour of liner material and of anchor-characteristics

    International Nuclear Information System (INIS)

    Oberpichler, R.; Schnellenbach, G.

    1975-01-01

    The thin liner attached by anchors like a membrane to the interior wall of a prestressed concrete reactor pressure vessel (PCRV) has to provide the leak-tightness of the vessel. Furthermore the liner may serve as internal shuttering for placing of concrete as well as a support for the cooling system. The two-dimensional behaviour of the liner is investigated with regard to non-linear anchor-characteristics and non-linear material behaviour of the liner. The analysis is based on a plane stress model under the assumption of a membrane state of the liner. Calculations are performed by the dynamic relaxation method. With the aid of available non-linear stress-strain diagrams, describing the post-buckling behaviour, individual panels are considered as buckled ones. The adjacent unbuckled panels are calculated on other non-linear diagrams. Strains and stresses in the liner and additional shear loads in the anchors can be calculated with arbitrary sizing and spacing of the anchors. With respect to the parameters they are easily controlled. Since actual loads on the liner are defined by the PCRV-behaviour, an economical and safe design is possible. Finally an extreme case is calculated to assess the maximum value of the shear-forces assuming zero post-buckling capacity for the buckled panel. (Auth.)

  9. Polymer materials basic research needs for energy applications

    Energy Technology Data Exchange (ETDEWEB)

    Macknight, W.J.; Baer, E.; Nelson, R.D. (eds.)

    1978-08-01

    The larger field covered in the workshop consists of (1) synthesis and characterization, (2) physical chemistry, (3) physics, and (4) engineering. Polymeric materials are properly regarded as new materials in their own right, not as replacements for existing materials. As such they need to be studied to understand the properties which are unique to them by virtue of their particular molecular structures. Technological applications will rationally follow from such studies. It is the objective of this report to point out basic research needs in polymer materials related to energy. The development of sophisticated instrumentation makes the task of molecular characterization possible on a level hitherto unattainable. Many of these instruments because of their size and complexity must of necessity be located at the DOE National Laboratories. The importance of personnel trained in the polymer field located at these facilities is emphasized. In the past there has been relatively little concerted polymer research within the energy community. This report attempts to describe the present situation and point out some needs and future research directions. (GHT)

  10. Enhancement of the quality of the reactor pressure vessel used in light water power plants by advanced material, fabrication and testing technologies

    International Nuclear Information System (INIS)

    Kussmaul, K.; Ewald, J.; Maier, G.; Schellhammer, W.

    1980-01-01

    Fracture safe assessment of nuclear reactor pressure vessels (RPV) is based upon an adequate stress analysis, reliable material characteristics, and acceptable defect sizes. Problems may arise concerning inhomogeneties, low toughness and crack phenomena as observed in the base material and heat affected zone (HAZ). Therefore, efforts have been made to develop a steel which would be both non-susceptible to embrittlement and/or cracking in the HAZ, and have a higher upper-shelf toughness of base and HAZ material. Tests have been made on inhomogeneties and defects and also on improvement of chemical composition, the steel-making process, welding procedures and the optimum temperature cycle and level for stress-relief heat treatment. To solve these problems, common testing methods were supplemented by tangential-cut techniques, small HAZ-tensile test procedures and HAZ-simulation techniques. Results indicate that 50 per cent of 100 investigated component-strength welds are affected by micro stress-relief cracking (SRC) on a micro-and millimetre scale. The 22 NiMoCr 37 steel with optimised chemical composition, and the 20 MnMoNi 55 steel are both resistant to stress-relief embrittlement and SRC. Specific welding techniques are found to limit SRC and proposals for optimum stress-relief temperatures are given. For the generation of new components, the fracture-safe analysis can now be based completely upon homogeneous and high upper-shelf base materials including the HAZ. (author)

  11. Mixed plasma-facing materials research at INEEL

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.

    2001-01-01

    Mixed-materials research at the Idaho National Engineering and Environmental Laboratory (INEEL) has focused on Be-C and W-C systems. The purpose of this work was to investigate hydrogen isotope retention in these systems. Plasma-mixed material layers using carbon coated Be and W specimens that were heat-treated and tungsten carbide specimens prepared by chemical vapor deposition (CVD) were simulated. Hydrogen isotope retention was investigated by means of thermal desorption spectroscopy (TDS) measurements on deuterium implanted samples

  12. Applications of neutron powder diffraction in materials research

    International Nuclear Information System (INIS)

    Kennedy, S.J.

    1996-01-01

    The aim of this article is to provide an overview of the applications of neutron powder diffraction in materials science. The technique is introduced with particular attention to comparison with the X-ray powder diffraction technique to which it is complementary. The diffractometers and special environment ancillaries operating around the HIFAR research reactor at the Australian Nuclear Science and Technology Organisation (ANSTO) are described. Applications of the technique which the advantage of the unique properties of thermal neutrons have been selected from recent materials studies undertaken at ANSTO

  13. Structural materials performance research at JRC-Institute for Energy

    International Nuclear Information System (INIS)

    Haehner, P.

    2009-01-01

    The DG-JRC structure and activities are presented in the paper. The Generation IV reactor concepts Very High Temperature Reactor (VHTR), Supercritical Water Reactor (SCWR) and Lead Cooled Reactor (LCR) are currently under study at the JRC. Requirements for innovative nuclear systems and material-related operational condition are under investigation. Considering the operational experience with current nuclear industry, these conditions imply demanding challenges from the structural materials point of view. The European Projects and initiatives and coordinated research programs are also presented

  14. Long Term Validation of High Precision RTK Positioning Onboard a Ferry Vessel Using the MGBAS in the Research Port of Rostock

    Directory of Open Access Journals (Sweden)

    Ralf Ziebold

    2017-09-01

    Full Text Available In order to enable port operations, which require an accuracy of about 10cm, the German Aerospace Center (DLR operates the Maritime Ground Based Augmentation Service (MGBAS in the Research Port of Rostock. The MGBAS reference station provides GPS dual frequency code + phase correction data, which are continuously transmitted via an ultra-high frequency (UHF modem. Up to now the validation of the MGBAS was rather limited. Either a second shore based station was used as an artificial user, or measurement campaigns on a vessel with duration of a few hours have been conducted. In order to overcome this, we have installed three separate dual frequency antennas and receivers and a UHF modem on the Stena Line ferry vessel Mecklenburg-Vorpommern which is plying between Rostock and Trelleborg. This paper concentrates on the analysis of the highly accurate phase based positioning with a Real Time Kinematic (RTK algorithm, using correction data received by the UHF modem onboard the vessel. We analyzed the availability and accuracy of RTK fix solutions for several days, whenever the ferry vessel was inside the service area of the MGBAS.

  15. First Materials Science Research Facility Rack Capabilities and Design Features

    Science.gov (United States)

    Cobb, S.; Higgins, D.; Kitchens, L.; Curreri, Peter (Technical Monitor)

    2002-01-01

    The first Materials Science Research Rack (MSRR-1) is the primary facility for U.S. sponsored materials science research on the International Space Station. MSRR-1 is contained in an International Standard Payload Rack (ISPR) equipped with the Active Rack Isolation System (ARIS) for the best possible microgravity environment. MSRR-1 will accommodate dual Experiment Modules and provide simultaneous on-orbit processing operations capability. The first Experiment Module for the MSRR-1, the Materials Science Laboratory (MSL), is an international cooperative activity between NASA's Marshall Space Flight Center (MSFC) and the European Space Agency's (ESA) European Space Research and Technology Center (ESTEC). The MSL Experiment Module will accommodate several on-orbit exchangeable experiment-specific Module Inserts which provide distinct thermal processing capabilities. Module Inserts currently planned for the MSL are a Quench Module Insert, Low Gradient Furnace, and a Solidification with Quench Furnace. The second Experiment Module for the MSRR-1 configuration is a commercial device supplied by MSFC's Space Products Development (SPD) Group. Transparent furnace assemblies include capabilities for vapor transport processes and annealing of glass fiber preforms. This Experiment Module is replaceable on-orbit. This paper will describe facility capabilities, schedule to flight and research opportunities.

  16. LANL Robotic Vessel Scanning

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-25

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

  17. Irradiation facilities for materials research: IFMIF and small scale installations

    International Nuclear Information System (INIS)

    Perlado, J. M.; Victoria, M.

    2007-01-01

    The research of advance materials in nuclear fields such as new fission reactors (Generation-IV), Accelerator Driven Systems for Transmutation of Radioactive Wastes and Nuclear Fusion, is becoming very much common in the types of low activation and radiation resistant Materials. Ferritic-Martensitic Steels (based in 9-12 Cr) with or without Oxide Dispersion Techniques (Ytria Nanoparticles), Composites materials are becoming the new generation to answer requirements of high temperature, high radiation resistance of structural materials. Special dedication is appearing in general research programmes to this area of Materials. The understanding of their final performance needs a wider knowledge of the mechanisms of radiation damage in these materials from the atomistic scale to the macroscopic responses. New extensive campaigns are being funded to irradiate from simple elements to model alloys and finally the complex materials themselves. That sequence and its state of art will be presented One clear technique for that understanding is the Multi scale Modelling which includes simulation techniques from quantum mechanics, molecular dynamics, defects diffusion, mesoscopic modelling and finally the macroscopic constitutive relations for macroscopic analysis. However, in each one of these steps is necessary a systematic and well established program of experiments that combines the irradiation and the very detailed analysis with techniques such as Transmission Electron Microscope, Positron Annihilation, SIMS, Atom Probe, Nanoindebntation. A key aspect that wants to be presented in this work is the state of art and discussion of Irradiation Facilities for Materials studies. Those facilities goes from ion implantation sources, small accelerator, Experimental Reactors such High Flux Reactor, sophisticated Triple Beams Sources as JANNUS in France to generate at the same time displacements-hydrogen-helium, and projected very large neutron installation such as IFMIF. The role to

  18. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    International Nuclear Information System (INIS)

    Ura, Tamaki; Takamasa, Tomoji; Nishimura, Hajime

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  19. Study on operation conditions and an operation system of a nuclear powered submersible research vessel, 'report of working group on application of a very small nuclear reactor to an ocean research'

    Energy Technology Data Exchange (ETDEWEB)

    Ura, Tamaki [Tokyo Univ., Tokyo (Japan); Takamasa, Tomoji [Tokyo Univ. of Mercantile Marine, Tokyo (Japan); Nishimura, Hajime [Japan Marine Science and Technology Center, Yokosuka, Kanagawa (JP)] [and others

    2001-07-01

    JAERI has studied on design of a nuclear powered submersible research vessel, which will navigate under sea mainly in the Arctic Ocean, as a part of the design activity of advanced marine reactors. This report describes operation conditions and an operating system of the vessel, which were discussed by the specialists of hull design, sound positioning, ship motions and oceanography, etc. The design conditions on ship motions for submersible vessels were surveyed considering regulations in our country, and ship motions were evaluated in the cases of underwater and surface navigations taking account of observation activities in the Arctic Ocean. The effect of ship motions on the compact nuclear reactor SCR was assessed. A submarine transponder system and an on-ice communication buoy system were examined as a positioning and communication system, supposing the activity under ice. The interval between transponders or communication buoys was recommended as 130 km. Procedures to secure safety of nuclear powered submersible research vessel were discussed according to accidents on the hull or the nuclear reactor. These results were reflected to the concept of the nuclear powered submersible research vessel, and subjects to be settled in the next step were clarified. (author)

  20. Modelling of in-vessel retention after relocation of corium into the lower plenum - Evaluation of the temperature field and of the viscoplastic deformation of the vessel wall. Reactor safety research, project No.:150 1254 - Final report; Beitrag zur Modellierung der Schmelzerueckhaltung im RDB nach Verlagerung von Corium in das untere Plenum - Berechnung des Temperaturfeldes und der viskoplastischen Verformung der Behaelterwand. Reaktorsicherheitsforschung, Vorhaben-Nr.: 150 1254 - Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E.; Willschuetz, H.G. [Forschungszentrum Rossendorf e.V. (FZR), Dresden (Germany)

    2005-01-01

    Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute Of Safety Research of the FZR a finite element model has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal hydraulic and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre- and post test calculations for the FOREVER test series representing the RPV of a PWR in the scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. The results of the calculations can be summarised as follows: The creeping process is caused by the simultaneous presence of high temperature (>600 C) and pressure (>1 MPa). The hot focus region is the most endangered zone exhibiting the highest creep strain rates. The exact level of temperature and pressure has an influence on the vessel failure time but not on the failure position. The failure time can be predicted with an uncertainty of 20 to 25%. This uncertainty is caused by the large scatter and the high temperature sensitivity of the viscoplastic properties of the RPV steel. Contrary to the hot focus region, the lower centre of the vessel head exhibits a higher strength because of the lower temperatures in this zone. The lower part moves down without significant deformation. Therefore it can be assumed, that the vessel failure can be retarded or prevented by supporting this range. The

  1. Antibody-linked drug destroys tumor cells and tumor blood vessels in many types of cancer | Center for Cancer Research

    Science.gov (United States)

    A team led by Brad St. Croix, Ph.D., Senior Associate Scientist, Mouse Cancer Genetics Program, has developed an antibody-drug conjugate (ADC) that destroys both tumor cells and the blood vessels that nourish them. The drug significantly shrank breast tumors, colon tumors and several other types of cancer and prolonged survival. Learn more...  

  2. Determination of J-integral R-curves for the pressure vessel material A 533 B1 using the potential drop technique and the multi-specimen method

    International Nuclear Information System (INIS)

    Krompholz, K.; Ullrich, G.

    1985-01-01

    J-integral experiments at room temperature were performed on three point bend type specimens of the nuclear pressure vessel material A 533 B1 with a/w-ratios of 0.3 and 0.5. Following the ASTM-proposal for the multi-specimen technique a value is obtained close to the value obtained in the HSST round robin test. On the other hand, from the measurement of the Jsub(IC)-value by means of the potential drop technique there is an indication that a lower value of Jsub(IC) is correct. This is in agreement with the multi-specimen technique using linear regression lines without excluding 'invalid' points. That is reasonable if fractographic investigations gives clear indications that stable crack growth has occurred as is the case in this work. (Auth.)

  3. Basic materials research programs at the U.S. Air Force Office of Scientific Research

    International Nuclear Information System (INIS)

    Carlson, Herbert C.; Goretta, K.C.

    2006-01-01

    The Air Force Office of Scientific Research (AFOSR) annually sponsors approximately 5000 research scientists at 1000 universities and laboratories, generating about 10,000 Ph.D. graduates per decade, all expected to publish their basic research findings in peer-reviewed journals. After a brief introduction of the nature of AFOSR's support to basic research in the U.S. and international scientific communities, work it supports at the frontiers of materials science is highlighted. One focused research theme that drives our investment is the MEANS program. It begins with the end in mind; materials are designed with practicable manufacture as an explicit initial goal. AFOSR's broad research portfolio comprises many materials. Nanotechnology efforts include optical materials that reduce distortion to the scale of the nanoparticles themselves. Advances in semiconductors include breakthroughs in Group III nitrides, some of which emanated from Asia under sponsorship from AFOSR's Asian office. Advances in structural materials include those for use at ultra-high temperatures and self-healing composites. The growing role of high-performance computing in design and study of functional, biological, and structural materials is also discussed

  4. Electrostatic Levitation: A Tool to Support Materials Research in Microgravity

    Science.gov (United States)

    Rogers, Jan; SanSoucie, Mike

    2012-01-01

    Containerless processing represents an important topic for materials research in microgravity. Levitated specimens are free from contact with a container, which permits studies of deeply undercooled melts, and high-temperature, highly reactive materials. Containerless processing provides data for studies of thermophysical properties, phase equilibria, metastable state formation, microstructure formation, undercooling, and nucleation. The European Space Agency (ESA) and the German Aerospace Center (DLR) jointly developed an electromagnetic levitator facility (MSL-EML) for containerless materials processing in space. The electrostatic levitator (ESL) facility at the Marshall Space Flight Center provides support for the development of containerless processing studies for the ISS. Apparatus and techniques have been developed to use the ESL to provide data for phase diagram determination, creep resistance, emissivity, specific heat, density/thermal expansion, viscosity, surface tension and triggered nucleation of melts. The capabilities and results from selected ESL-based characterization studies performed at NASA's Marshall Space Flight Center will be presented.

  5. Rethinking Socialization Research through the Lens of New Materialism

    Directory of Open Access Journals (Sweden)

    Grit Höppner

    2017-09-01

    Full Text Available In recent decades, socialization research appears to have suffered the loss of its former capacity to explain the processes of becoming a socialized subject in a social environment. In this article, I review socialization theories taking into account assumptions regarding human subjects and their social environments. I confront them with the idea of rethinking dualisms, ontologies, and agencies addressed by the field of new materialism. I propose a new materialist-inspired socialization theory that assumes that humans, knowledge, and material environments become inseparable parts of (gendered socialization processes in a world of constant change. This approach contributes to socialization theory and methodology because it illustrates precisely how humans and non-humans coproduce socialization in situated material-discursive processes.

  6. NDT studies of laser cladding defects of pure copper on SS316L for in vessel materials for fusion reactor applications

    International Nuclear Information System (INIS)

    Shaikh, S.; Buddu, Ramesh Kumar; Raole, P.M.; Sarkar, B.

    2015-01-01

    The pure thick copper coatings of 1-3 mm are required for the in-vessel materials for the plasma facing components in fusion reactor systems to extract the very high heat flux in shorter durations (like VDEs) and to protect the in vessel components. Laser cladding technique is one of the potential technique for thick coatings on substrate materials. The present study reports the NDT characterization studies carried on samples of pure copper powder cladded on SS316L substrates of thickness 1 mm - 3 mm , fabricated by CO_2 laser system. Process parameters optimization like laser power, laser travel speed, spot size, powder feed rate and shield gas flow show the effect on quality of final cladding on steel substrates. X-ray radiography and Ultrasonic testing has been carried out thoroughly on the fabricated samples and defects are analyzed. Ultrasonic scan tests using different probes are employed as the interface defects are not thoroughly revealed by radiography. The calibration has been carried out by the test sample plate with known defect size created and various process parameters like amplitude, gain and metal velocity, relevant to specimen are chosen for probes calibration. The interface defects of porosity, lack of penetration, cracks or group porosities are observed in few set of samples developed. Radiography examination revealed the porosity at extreme edges and distributed porosity in the middle for thick cladding. Ultrasonic manual A-scanning with TR probe provides qualitative information about flaw and broadly gives its location of the defects. Samples of 1 mm thick cladding have shown relatively less porosity defects at the interface compared to 3 mm thick samples. (author)

  7. Ion backscattering techniques applied in materials science research

    International Nuclear Information System (INIS)

    Sood, D.K.

    1978-01-01

    The applications of Ion Backscattering Technique (IBT) to material analysis have expanded rapidly during the last decade. It is now regarded as an analysis tool indispensable for a versatile materials research program. The technique consists of simply shooting a beam of monoenergetic ions (usually 4 He + ions at about 2 MeV) onto a target, and measuring their energy distribution after backscattering at a fixed angle. Simple Rutherford scattering analysis of the backscattered ion spectrum yields information on the mass, the absolute amount and the depth profile of elements present upto a few microns of the target surface. The technique is nondestructive, quick, quantitative and the only known method of analysis which gives quantitative results without recourse to calibration standards. Its major limitations are the inability to separate elements of similar mass and a complete absence of chemical-binding information. A typical experimental set up and spectrum analysis have been described. Examples, some of them based on the work at the Bhabha Atomic Research Centre, Bombay, have been given to illustrate the applications of this technique to semiconductor technology, thin film materials science and nuclear energy materials. Limitations of IBT have been illustrated and a few remedies to partly overcome these limitations are presented. (auth.)

  8. Research on superconducting generator and materials in Japan

    International Nuclear Information System (INIS)

    Uyeda, K.; Maki, N.; Kurihara, S.; Ueda, A.; Hirose, S.; Itoh, K.

    1988-01-01

    As a first step of application of superconducting technology to electric power equipment, the practical use of superconducting generator is sucessfully developed, enhanced generation efficiency, reduced construction cost, improved stability limit. For the development, it is required to integrated such technical assets as new generator design technology based on detailed analysis of techniques and high strength material for with standing intensive electro-magnetic force. This paper describes history and results of research and development of superconducting generator for experimental machines, the results of feasibility study of pilot generator, and master plan for research and development of superconducting technology for applications to generator and the other power apparatus

  9. Technical Education Outreach in Materials Science and Technology Based on NASA's Materials Research

    Science.gov (United States)

    Jacobs, James A.

    2003-01-01

    The grant NAG-1 -2125, Technical Education Outreach in Materials Science and Technology, based on NASA s Materials Research, involves collaborative effort among the National Aeronautics and Space Administration s Langley Research Center (NASA-LaRC), Norfolk State University (NSU), national research centers, private industry, technical societies, colleges and universities. The collaboration aims to strengthen math, science and technology education by providing outreach related to materials science and technology (MST). The goal of the project is to transfer new developments from LaRC s Center for Excellence for Structures and Materials and other NASA materials research into technical education across the nation to provide educational outreach and strengthen technical education. To achieve this goal we are employing two main strategies: 1) development of the gateway website and 2) using the National Educators Workshop: Update in Engineering Materials, Science and Technology (NEW:Updates). We have also participated in a number of national projects, presented talks at technical meetings and published articles aimed at improving k-12 technical education. Through the three years of this project the NSU team developed the successful MST-Online site and continued to upgrade and update it as our limited resources permitted. Three annual NEW:Updates conducted from 2000 though 2002 overcame the challenges presented first by the September 11,2001 terrorist attacks and the slow U.S. economy and still managed to conduct very effective workshops and expand our outreach efforts. Plans began on NEW:Update 2003 to be hosted by NASA Langley as a part of the celebration of the Centennial of Controlled Flight.

  10. Applied solid state science advances in materials and device research

    CERN Document Server

    Wolfe, Raymond

    2013-01-01

    Applied Solid State Science: Advances in Materials and Device Research, Volume 1 presents articles about junction electroluminescence; metal-insulator-semiconductor (MIS) physics; ion implantation in semiconductors; and electron transport through insulating thin films. The book describes the basic physics of carrier injection; energy transfer and recombination mechanisms; state of the art efficiencies; and future prospects for light emitting diodes. The text then discusses solid state spectroscopy, which is the pair spectra observed in gallium phosphide photoluminescence. The extensive studies

  11. Current results of coal gasification materials research at GRI

    International Nuclear Information System (INIS)

    Hill, V.L.; Barone, S.P.; Meyer, H.S.

    1984-01-01

    Corrosion, erosion/corrosion and mechanical property testing of commercial available materials in coal gasification atmospheres has been supported by the Gas Research Institute (GRI) since 1978. Recent corrosion data developed in the program for gasification and methanation technologies under development by GRI are presented. A brief discussion of typical results of long-term stress-rupture tests in coal gasification atmospheres is included

  12. Rethinking Socialization Research through the Lens of New Materialism

    OpenAIRE

    Höppner, Grit

    2017-01-01

    In recent decades, socialization research appears to have suffered the loss of its former capacity to explain the processes of becoming a socialized subject in a social environment. In this article, I review socialization theories taking into account assumptions regarding human subjects and their social environments. I confront them with the idea of rethinking dualisms, ontologies, and agencies addressed by the field of new materialism. I propose a new materialist-inspired socialization theor...

  13. Annual report 1992 on research and development work by the IMF, Institute for Materials Research

    International Nuclear Information System (INIS)

    1993-03-01

    The present annual report describes the activities undertaken by the IMF in the following areas: 1. Low-pollutant and low-waste techniques (treatment and utilization of special wastes); 2. Nuclear fusion (studies for NET/ITER; structural materials for fusion devices; superconducting magnets; plasmas heating technique; blanket development; component-related safety investigations); 3. Nuclear safety research (safety and materials of fast breeders; transient behaviour of fast breeder fuel elements; LWR-oriented safety research; containment concepts for PWR-plants); 4. Nuclear waste management (materials studies of waste forms); 5. Superconductivity (superconductor developments); 6. Microsystems engineering (development and testing of compact and laminated materials of microsystems engineering); 7. Handling technique (remote handling components for invasive surgery); 8. Materials and interfaces (inter alia high-performance ceramics, failure behaviour, LCP, biomechanics). The appendix lists all publications or primary reports by the IMF in 1992. (orig./HP) [de

  14. Enhancement of the quality of the reactor pressure vessel used in light water power plants by advanced material fabrication and testing technoligies

    International Nuclear Information System (INIS)

    Kussmaul, K.; Ewald, J.

    1977-01-01

    The fracture safe assessment of nuclear pressure vessels (RPV) is based upon: (i) an adequate stress analysis, (ii) reliable material characteristics, (iii) acceptable defects sizes. There may arise problems which are related to the inhomogeneity, low toughness and crack phenomena sometimes observed in the base material and heat affected zone (HAZ). Due to this it is difficult and in some respects even impossible to measure the decisive values of (fracture-) toughness and defects. Apart from the significance of those facts for existing RPVs, all efforts were directed to provide a steel which should be non-susceptible to embrittlement and/or cracking in the HAZ and simultaneously yielding in a higher upper shelf toughness of base and HAZ material. These objections were pursued in cooperation with manufacturers, vendors and inspection authorities by the following activities. (i) Detailed investigations to obtain information on: occurrence and size of inhomogeneities and defects, especially stress relief cracking (SCR), toughness properties adjacent to defects; (ii) improvement of: chemical composition, steel making processes, welding procedures, optimum temperature cycle and level for stress relief heat-treatment. In order to solve these tasks it was necessary to develop additional tools and to correlate all partial results which were newly elaborated. (Auth.)

  15. Neuromorphic Computing – From Materials Research to Systems Architecture Roundtable

    Energy Technology Data Exchange (ETDEWEB)

    Schuller, Ivan K. [Univ. of California, San Diego, CA (United States); Stevens, Rick [Argonne National Lab. (ANL), Argonne, IL (United States); Univ. of Chicago, IL (United States); Pino, Robinson [Dept. of Energy (DOE) Office of Science, Washington, DC (United States); Pechan, Michael [Dept. of Energy (DOE) Office of Science, Washington, DC (United States)

    2015-10-29

    Computation in its many forms is the engine that fuels our modern civilization. Modern computation—based on the von Neumann architecture—has allowed, until now, the development of continuous improvements, as predicted by Moore’s law. However, computation using current architectures and materials will inevitably—within the next 10 years—reach a limit because of fundamental scientific reasons. DOE convened a roundtable of experts in neuromorphic computing systems, materials science, and computer science in Washington on October 29-30, 2015 to address the following basic questions: Can brain-like (“neuromorphic”) computing devices based on new material concepts and systems be developed to dramatically outperform conventional CMOS based technology? If so, what are the basic research challenges for materials sicence and computing? The overarching answer that emerged was: The development of novel functional materials and devices incorporated into unique architectures will allow a revolutionary technological leap toward the implementation of a fully “neuromorphic” computer. To address this challenge, the following issues were considered: The main differences between neuromorphic and conventional computing as related to: signaling models, timing/clock, non-volatile memory, architecture, fault tolerance, integrated memory and compute, noise tolerance, analog vs. digital, and in situ learning New neuromorphic architectures needed to: produce lower energy consumption, potential novel nanostructured materials, and enhanced computation Device and materials properties needed to implement functions such as: hysteresis, stability, and fault tolerance Comparisons of different implementations: spin torque, memristors, resistive switching, phase change, and optical schemes for enhanced breakthroughs in performance, cost, fault tolerance, and/or manufacturability.

  16. Failure probability analysis on mercury target vessel

    International Nuclear Information System (INIS)

    Ishikura, Syuichi; Futakawa, Masatoshi; Kogawa, Hiroyuki; Sato, Hiroshi; Haga, Katsuhiro; Ikeda, Yujiro

    2005-03-01

    Failure probability analysis was carried out to estimate the lifetime of the mercury target which will be installed into the JSNS (Japan spallation neutron source) in J-PARC (Japan Proton Accelerator Research Complex). The lifetime was estimated as taking loading condition and materials degradation into account. Considered loads imposed on the target vessel were the static stresses due to thermal expansion and static pre-pressure on He-gas and mercury and the dynamic stresses due to the thermally shocked pressure waves generated repeatedly at 25 Hz. Materials used in target vessel will be degraded by the fatigue, neutron and proton irradiation, mercury immersion and pitting damages, etc. The imposed stresses were evaluated through static and dynamic structural analyses. The material-degradations were deduced based on published experimental data. As a result, it was quantitatively confirmed that the failure probability for the lifetime expected in the design is very much lower, 10 -11 in the safety hull, meaning that it will be hardly failed during the design lifetime. On the other hand, the beam window of mercury vessel suffered with high-pressure waves exhibits the failure probability of 12%. It was concluded, therefore, that the leaked mercury from the failed area at the beam window is adequately kept in the space between the safety hull and the mercury vessel by using mercury-leakage sensors. (author)

  17. Compilation of contract research for the Materials Engineering Branch, Division of Engineering Technology. Annual report for FY 1983. Vol.2

    International Nuclear Information System (INIS)

    1984-03-01

    This report presents summaries of the research work performed during Fiscal Year 1983 by laboratories and organizations under contracts administered by the NRC's Materials Engineering Branch, Office of Nuclear Regulatory Research. Each contractor has written a more complete and detailed annual report of their work which can be obtained by writing to NRC. The contractor reports are organized into the major areas of concern to Primary System Integrity, which is the main focus for the branch's research. These areas are: Vessel and Piping Fracture Mechanics; Pressure Vesel Surveillance Dosimetry; Steam Generators, Aging, and Environmental Cracking; and Non-Destructive Examination. The research programs reported provide information on the overall program objectives, a more limited scope of work for FY 1983, a technical description of the year's work, and a brief forecast of the plans for continuing work

  18. An accountancy system for nuclear materials control in research centres

    International Nuclear Information System (INIS)

    Buttler, R.; Bueker, H.; Vallee, J.

    1979-01-01

    The Nuclear Accountancy and Control System (NACS) was developed at KFA Juelich in accordance with the requirements of the Non-Proliferation Treaty. The main features are (1) recording of nuclear material in inventory items. These are combined to form batches wherever suitable; (2) extrapolation of accounting data as a replacement for detailed measurement of inventory items data. Recording and control of nuclear material are carried out on two levels with access to a common data bank. The lower level deals with nuclear materials handling plus internal management while on the upper level there is a central control point which is responsible for nuclear safeguarding within the entire research centre. By keeping the organizational and technical infrastructure it was possible to develop a system which is both economical and operator-oriented. In this system the emphasis of nuclear safeguarding is placed on the acquisition of the nuclear material inventory. As much consideration has been given to the interests of the various operational levels and organizational units as to internal and national regulations. Since it is part of the safeguarding and control system, access to the NACS must be restricted to a limited number of users only. Furthermore, it must include facilities for manual control in the form of records. Authorization for access must correspond with the various tasks of different user groups. All necessary data are acquired decentrally in the organizational units and entered via a terminal. It is available to the user groups on both levels through a central data bank. To meet all requirements, the NACS has been designed as an integrated, computer-assisted information system for the automated processing of extensive and multi-level nuclear materials data. As part of the preventive measures entailed with nuclear safeguarding, the accountancy system enables the operator of a nuclear plant to furnish proof of non-diversion of nuclear material. (author)

  19. 1D/2D analyses of the lower head vessel in contact with high temperature melt

    International Nuclear Information System (INIS)

    Chang, Jong Eun; Cho, Jae Seon; Suh, Kune Y.; Chung, Chang H.

    1998-01-01

    One- and two-dimensional analyses were performed for the ceramic/metal melt and the vessel to interpret the temperature history of the outer surface of the vessel wall measured from typical Al 2 O 3 /Fe thermite melt tests LAVA (Lower-plenum Arrested Vessel Attack) spanning heatup and cooldown periods. The LAVA tests were conducted at the Korea Atomic Energy Research Institute (KAERI) during the process of high temperature molten material relocation from the delivery duct down into the water in the test vessel pressurized to 2.0 MPa. Both analyses demonstrated reasonable predictions of the temperature history of the LHV (Lower Head Vessel). The comparison sheds light on the thermal hydraulic and material behavior of the high temperature melt within the hemispherical vessel

  20. Base technology approaches in materials research for future nuclear applications

    International Nuclear Information System (INIS)

    Kondo, Tatsuo

    1992-01-01

    In the development of advanced nuclear systems for future, majority of critical issues in material research and development are more or less related with the effects of neutron irradiation. The approaches to those issues in the past have been mainly concerned with interpretation of the facts and minor modification of existing materials, having been inevitably of passive nature. In combating against predicted complex effects arising from variety of critical parameters, approaches must be reviewed more strategically. Some attempts of shifting research programs to such a direction have been made at JAERI in the Base (Common) Technology Programs either by adding to or restructuring the existing tasks. Major tasks currently in progress after the reorientation are categorized in several disciplines including new tasks for material innovation and concept development for neutron sources. The efforts have been set forth since 1988, and a few of them are now mature to transfer to the tasks in the projects of advanced reactors. The paper reviews the status of some typical activities emphasizing the effects of the reorientation and possible extensions of the outcomes to future applications. (author)

  1. High energy neutron source for materials research and development

    International Nuclear Information System (INIS)

    Odera, M.

    1989-01-01

    Requirements for neutron source for nuclear materials research are reviewed and ESNIT, Energy Selective Neutron Irradiation Test facility proposed by JAERI is discussed. Its principal aims of a wide neutron energy tunability and spectra peaking at each energy to enable characterization of material damage process are demanding but attractive goals which deserve detailed study. It is also to be noted that the requirements make a difference in facility design from those of FMIT, IFMIF and other high energy intense neutron sources built or planned to date. Areas of technologies to be addressed to realize the ESNIT facility are defined and discussed. In order to get neutron source having desired spectral characteristics keeping moderate intensity, projectile and target combinations must be examined including experimentation if necessary. It is also desired to minimize change of flux density and energy spectrum according to location inside irradiation chamber. Extended target or multiple targets configuration might be a solution as well as specimen rotation and choice of combination of projectile and target which has minimum velocity of the center of mass. Though relevant accelerator technology exists, it is to be stressed that considerable efforts must be paid, especially in the area of target and irradiation devices to get ESNIT goal. Design considerations to allow hands-on maintenance and future upgrading possibility are important either, in order to exploit the facility fully for nuclear materials research and development. (author)

  2. Micro-Scale Experiments and Models for Composite Materials with Materials Research

    DEFF Research Database (Denmark)

    Zike, Sanita

    Numerical models are frequently implemented to study micro-mechanical processes in polymer/fibre composites. To ensure that these models are accurate, the length scale dependent properties of the fibre and polymer matrix have to be taken into account. Most often this is not the case, and material...... properties acquired at macro-scale are used for micro-mechanical models. This is because material properties at the macro-scale are much more available and the test procedures to obtain them are well defined. The aim of this research was to find methods to extract the micro-mechanical properties of the epoxy...... resin used in polymer/fibre composites for wind turbine blades combining experimental, numerical, and analytical approaches. Experimentally, in order to mimic the stress state created by a void in a bulk material, test samples with finite root radii were made and subjected to a double cantilever beam...

  3. NATO Advanced Research Workshop on Molecular Engineering for Advanced Materials

    CERN Document Server

    Schaumburg, Kjeld

    1995-01-01

    An important aspect of molecular engineering is the `property directed' synthesis of large molecules and molecular assemblies. Synthetic expertise has advanced to a state which allows the assembly of supramolecules containing thousands of atoms using a `construction kit' of molecular building blocks. Expansion in the field is driven by the appearance of new building blocks and by an improved understanding of the rules for joining them in the design of nanometer-sized devices. Another aspect is the transition from supramolecules to materials. At present no single molecule (however large) has been demonstrated to function as a device, but this appears to be only a matter of time. In all of this research, which has a strongly multidisciplinary character, both existing and yet to be developed analytical techniques are and will remain indispensable. All this and more is discussed in Molecular Engineering for Advanced Materials, which provides a masterly and up to date summary of one of the most challenging researc...

  4. Process research on non-CZ silicon material

    Science.gov (United States)

    1982-01-01

    High risk, high payoff research areas associated with he process for producing photovoltaic modules using non-CZ sheet material are investigated. All investigations are being performed using dendritic web silicon, but all processes are directly applicable to other ribbon forms of sheet material. The technical feasibility of forming front and back junctions in non-CZ silicon using liquid dopant techniques was determined. Numerous commercially available liquid phosphorus and boron dopant solutions are investigated. Temperature-time profiles to achieve N(+) and P(+) sheet resistivities of 60 + or - 10 and 40 + or - s10 ohms per square centimeter respectively are established. A study of the optimal method of liquid dopant application is performed. The technical feasibility of forming a liquid applied diffusion mask to replace the more costly chemical vapor deposited SiO2 diffusion mask was also determined.

  5. Electrical research on solar cells and photovoltaic materials

    Science.gov (United States)

    Orehotsky, J.

    1985-01-01

    A systematic study of the properties of various polymer pottant materials and of the electrochemical corrosion mechanisms in solar cell materials is required for advancing the technology of terrestrial photovoltaic modules. The items of specific concern in this sponsored research activity involve: (1) kinetics of plasticizer loss in PVB, (2) kinetics of water absorption and desorption in PVB, (3) kinetics of water absorption and desorption in EVA, (4) the electrical properties at PVB as a function of temperature and humidity, (5) the electrical properties of EVA as a function of temperature and humidity, (6) solar cell corrosion characteristics, (7) water absorption effects in PVB and EVA, and (8) ion implantation and radiation effects in PVB and EVA.

  6. Manufacturing radioactive material for medical, research and industrial applications

    International Nuclear Information System (INIS)

    Seidel, C.W.

    1992-01-01

    Hospitals, clinics and other medical complexes are among the most extensive users of radioactive material. Nuclear medicine uses radioactive solutions of Tc-99m, Tl-201, Ga-67, I-123, Xe-133 and other radiopharmaceuticals as diagnostic tools to evaluate dynamic functions of various organs in the body, detect cancerous tumors, sites of infection or other bodily dysfunctions. Examples of monitoring blood flow to the brain of a cocaine addict will be shown. Many different radionuclides are also produced for life science research and industrial applications. Some require long irradiations and are needed only periodically. Radiopharmaceutical manufactures look for reliable suppliers that can produce quality product at a reasonable cost. Worldwide production of the processed and unprocessed radionuclides and the enriched stable nuclides that are the target materials used in the accelerators and reactors around the world will be discussed. (author)

  7. The vessel fluence; Fluence cuve

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  8. Development and application of a material law for steel-fibre-reinforced concrete with regard to its use for pre-stressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.; Borgerhoff, M.

    1995-01-01

    On the basis of the evaluation of many publications on the mechanical behaviour of steel fibre reinforced concrete (SFRC) and on the results of experiments using an SFRC especially developed for pre-stressed concrete reactor vessels (PCRVs), a material law for SFRC including general multiaxial stress conditions has been developed. From fibre pull-out tests described in the literature and by use of the experimental results, relations describing the capable tensile stress in SFRC after cracking, as a function of crack width, have been derived. There is a significant increase in the biaxial compressive strength of SFRC compared with plain concrete. The improved behaviour under multiaxial stress conditions, with one of the principal stresses being tensile, is outlined in comparison with different formulations of failure envelopes of plain concrete. For the purpose of verifying the material law implemented in the computer program used, analyses have been carried out for experiments with SFRC beams. After some modification concerning the shear behaviour, load-displacement curves and realistic crack propagations which correspond well have been obtained. In the stand-tube area in the centre of a PCRV top cap the use of SFRC is advantageous because of the difficulties concerning the arrangement of reinforcement in the concrete between the tubes. (orig.)

  9. Next Generation Nuclear Plant Materials Research and Development Program Plan

    International Nuclear Information System (INIS)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-01-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R and D) Program is responsible for performing R and D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R and D Program includes the following elements: (1) Developing a specific approach, program plan and other project management

  10. Next Generation Nuclear Plant Materials Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for

  11. Filling the gaps in SCWR materials research: advanced nuclear corrosion research facilities in Hamilton

    International Nuclear Information System (INIS)

    Krausher, J.L.; Zheng, W.; Li, J.; Guzonas, D.; Botton, G.

    2011-01-01

    Research efforts on materials selection and development in support of the design of supercritical water-cooled reactors (SCWRs) have produced a considerable amount of data on corrosion, creep and other related properties. Summaries of the data on corrosion [1] and stress corrosion cracking [2] have recently been produced. As research on the SCWR advances, gaps and limitations in the published data are being identified. In terms of corrosion properties, these gaps can be seen in several areas, including: 1) the test environment, 2) the physical and chemical severity of the tests conducted as compared with likely reactor service/operating conditions, and 3) the test methods used. While some of these gaps can be filled readily using existing facilities, others require the availability of advanced test facilities for specific tests and assessments. In this paper, highlights of the new materials research facilities jointly established in Hamilton by CANMET Materials Technology Laboratory and McMaster University are presented. (author)

  12. The role of material evidence in architectural research

    DEFF Research Database (Denmark)

    2011-01-01

    The following texts explore the production of knowledge in architectural research. Focussing on a wide definition of practice led research, the aim for these texts is to discuss how the practices of architectural design; drawing, modelling, prototyping and building embody a particular set of know...... emerges are the fundamental crossovers between these practices. What we see here is that drawing is as much a practice of theoretical reflection as one of detailing, prototyping as much one of physical as conceptual testing.......The following texts explore the production of knowledge in architectural research. Focussing on a wide definition of practice led research, the aim for these texts is to discuss how the practices of architectural design; drawing, modelling, prototyping and building embody a particular set...... of knowledges that inform architectural thinking. Architectural reflection is allied with it media. It is through the drawing, the model and the built that architecture is conceived and developed. In practice based research working through design means reflecting through the production of material evidence...

  13. Karlsruhe Nuclear Research Center, Institute of Materials Research. Progress report on research and development work in 1993

    International Nuclear Information System (INIS)

    1994-03-01

    The Institute consists of three parts IMF I, IMF II and IMF III. The tasks are divided into applied material physics (IMF I), material and structural mechanics (IMF II) and material process technology (IMF III). IMF I works preferably on the development of metallic, non-metallic and compound materials and on questions of the structure and properties of boundary surfaces and surface protection coatings. The main work of IMF II is the reliability of components, failure mechanics and the science of damage. IMF III examines process technology questions in the context of the manufacture of ceramic materials and fusion materials and the design of nuclear components. The Institute works on various main points of the Kernforschungszentrum in its research work, particularly in nuclear fusion, micro-system technique, nuclear safety research, superconductivity and in processes with little harmful substances and waste. Material and strength problems for future fusion reactors and fission reactors, in powerful micro systems and safety-related questions of nuclear technology are examined. Also, research not bound to projects in the field of metallic, ceramic and polymer materials for high stresses is carried out. (orig.) [de

  14. Process research of non-CZ silicon material

    Science.gov (United States)

    Campbell, R. B.

    1984-01-01

    Advanced processing techniques for non-CZ silicon sheet material that might improve the cost effectiveness of photovoltaic module production were investigated. Specifically, the simultaneous diffusion of liquid boron and liquid phosphorus organometallic precursors into n-type dendritic silicon web was examined. The simultaneous junction formation method for solar cells was compared with the sequential junction formation method. The electrical resistivity of the n-n and p-n junctions was discussed. Further research activities for this program along with a program documentation schedule are given.

  15. New era of neutron scattering research on advanced materials

    International Nuclear Information System (INIS)

    Ikeda, Susumu

    2001-01-01

    The projects of the next generation of pulsed spallation neutron source are planed in USA, Europe and Japan. They are one order of magnitude more powerful than the most powerful existing neutron source, ISIS in UK. They offer the exciting prospects for the future, and will open the new era of neutron scattering research on advanced materials. The Japanese project is named as the 'Joint project' between JAERI and KEK on high-intensity proton accelerators. The details of the neutron science facility in the 'Joint project' and the sciences to be developed are summarized. (author)

  16. Materials Science and Engineering-1989 Publications (Naval Research Laboratory)

    Science.gov (United States)

    1991-03-29

    Antamanide J.H. Konnert, P. D’Antonio, J.M. Cowley, and Analog. Crystal Structure of A. Higgs , H-J. Ou Perhydrosymmetric antamanide, Ultramicroscopy, 30, 371...Paired Boson Superconductor" Molecular Beam Epitaxy" W. Jin, S.D. Mahanti, A.K. Rajagopal A. Christou, N. Flevaris, A. Georgakilas, Solid State...33(3), 347-358 Si(100)" "Neutron Scattering from Fermion and S.M. Prokes, W.F. Tseng, A- Christou Boson Superconductors" Materials Research Society

  17. Containment vessel

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  18. Application of Chemistry in Materials Research at NASA GRC

    Science.gov (United States)

    Kavandi, Janet L.

    2016-01-01

    Overview of NASA GRC Materials Development. New materials enabled by new chemistries offering unique properties and chemical processing techniques. Durability of materials in harsh environments requires understanding and modeling of chemical interaction of materials with the environment.

  19. Tungsten alloy research at the US Army Materials Technology Laboratory

    International Nuclear Information System (INIS)

    Dowding, R.J.

    1991-01-01

    This paper reports that recent research into tungsten heavy alloys at the U. S. Army Materials Technology Laboratory (MTL) has explored many areas of processing and process development. The recrystallization and respheroidization of tungsten grains in a heavily cold worked heavy alloy has been examined and resulted in the identification of a method of grain refinement. Another area of investigation has been lightly cold worked. It was determined that it was possible to increase the strength and hardness of the tungsten grains by proper hat treatment. MTL has been involved in the Army's small business innovative research (SBIR) program and several programs have been funded. Included among these are a method of coating the tungsten powders with the alloying elements and the development of techniques of powder injection molding of heavy alloys

  20. Innovation in use and research on cementitious material

    International Nuclear Information System (INIS)

    Scrivener, Karen L.; Kirkpatrick, R. James

    2008-01-01

    In this paper we discuss innovations in concrete technology which are currently being applied in the field-namely high and ultra high performance (strength), and self consolidating concrete. We discuss the factors which have enabled these developments and ongoing needs in these areas. The importance of sustainability as the major driver for future innovations and prospects for development of new cementitious materials with lower environmental impact is briefly discussed. Finally the importance of innovation in research is examined. The dramatic development in experimental and computational techniques over recent years opens up wide-ranging possibilities for understanding the micro- and nano- scale chemical and physical processes which underlie performance at a macroscopic level. The example of computational approaches at the atomic and molecular scale is presented in detail. In order to exploit the opportunities presented by such new techniques, there needs to be greater efforts to structure interdisciplinary, multi-group research

  1. Radiation materials science. V. 7

    International Nuclear Information System (INIS)

    Zelenskij, V.F.

    1990-01-01

    This volume includes the papers of the international conference on radiation materials in Alushta, Ukraine in May 1990. The main topics are: basic research in radiation damage physics, a study of the structural materials for reactor cores; irradiation effect on reactor vessel, fuel, super- and semiconductor materials; investigation damage research methods

  2. Radiation materials science. V. 6

    International Nuclear Information System (INIS)

    Zelenskij, V.F.

    1990-01-01

    This volume includes the papers of the international conference on radiation materials in Alushta, Ukraine in May 1990. The main topics are: basic research in radiation damage physics, a study of the structural materials for reactor cores; irradiation effect on reactor vessel, fuel, super- and semiconductor materials; investigation damage research methods

  3. Research and survey of structural materials for fast breeder reactor

    International Nuclear Information System (INIS)

    Baba, Kyozi

    1986-01-01

    In the development of FBRs, the selection of the materials for high temperature use is an important factor which determines the reliability of plants. The materials for secondary sodium system equipment centering around steam generators are affected by the type of steam generators, economical efficiency, aseismatic ability, fuel design and the method of removing core decay heat. At present, the conceptual design of demonstration FBRs (tank type, loop type) is in progress, and the research on the materials for steam generator tubes was completed in fiscal year 1984 by 10 electric power companies and 4 other companies. The four kinds of the steel tested were modified 9Cr-1Mo steel, 9Cr-2Mo steel, 12Cr-1Mo-V-Nb steel and Alloy 800. The specifications of the modified 9Cr-1Mo steel and Alloy 800 are shown. The results of tensile strength, creep strength, fatique strength, the characteristics after high temperature heating, weldability, and the strength of welded joints are reported. Also the weight of heating tubes was compared. The results of the general evaluation showed that 9Cr group steels were most promising. The matters to be examined hereafter are pointed out. (Kako, I.)

  4. A study on corium melt pool behavior under external vessel cooling : investigation of the first phase research results in the OECD RASPLAV project

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Kim, Sang Baik; Kim, Hee Dong; Yoo, Kun Joong

    1998-04-01

    The scope and contents of the OECD RASPLAV program are to investigate natural convection heat transfer in the corium, chemical and mechanical interaction between the corium and the reactor vessel, crust formation of the corium, and thermal behaviour of the corium by experiments and model development during external vessel cooling to prevent reactor vessel failure in severe accidents of nuclear power plant. This study includes evaluation and analysis of the RASPLAV V phase I results for three years between July 1, 1994 and June 30, 1997. These results supply technical basis for our experimental program on severe accident research. Two large-scale experiments of RASPLAV-AW-between the corium and the reactor vessel. Several small-scale experiments were conducted to analyze thermal stratification in the corium. The salt experiments were conducted to estimate the crust and the mushy region formation, and natural convection heat transfer in the corium. In the analytical studies, pre and post analysis of the RASPLAV-AW-200 experiments and evaluation of the salt test results have been performed using CONV 2 and 3D computer codes, which were developed during RASPLAV program phase I. Low density corium was separated from the high density corium during the RASPLAV-AW-200 tests and the TULPAN test, which was a new finding in the RASPLAV project phase I. From the salts test, heat flux distribution in the side wall heating case is similar to the direct internal heat generation case, and the crust formation is a little effect on heat transfer rate. The results of CONV 2 and 3 D were very well with with the experimental results. The results of RASLAV project phase I, such as furnace design and the techniques on fuel melting, are very helpful to our severe accident experimental program. (author). 57 refs., 13 tabs., 52 figs.

  5. Visualization of vessel traffic

    NARCIS (Netherlands)

    Willems, C.M.E.

    2011-01-01

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

  6. Acoustoelastic evaluation of the 20 MnMoNi 55 structural material of the pressure vessels of Angra 2 and 3 nuclear reactors

    International Nuclear Information System (INIS)

    Dutra, Marco Aurelio Monteiro

    2009-01-01

    The pressure vessels of the Angra II and Angra III nuclear power plants, which are of large thickness, have as one of their structural components the 20 MnMo Ni 55 steel. The acoustoelastic evaluation carried out through non-destructive ultrasonic tests was based on the variation of the speed of ultrasonic shear waves of normal incidence as a function of the stress applied in the material using the acoustic birefringence technique, which considers the fractional difference of the speeds of two ultrasonic waves propagating in orthogonal directions. In this work two experiments were carried out. In the first one, a new method for acquisition of ultrasonic signals for joint application with this technique of stress analysis was evaluated by comparison with the conventional one. The aim was to determine its potential and limitations for application in materials up to 120 mm of thickness. In the second experiment, the acoustoelastic behavior of the 20 MnMo Ni 55 steel was studied. The new method of acquiring ultrasonic signals led to satisfactory results and was used in the study of the acoustoelastic behavior of the 20 MnMoNi 55 steel. The acoustoelastic analysis indicated that the material has an anisotropic and heterogeneous behavior. The acoustoelastic constant of this material, obtained experimentally from the analysis of the graphs of the material behavior (birefringence x stress) under compression tests performed on specimens of the 20 MnMo Ni 55 steel, was used for the quantitative stress analysis. A specimen of the 20 MnMo Ni 55 steel was subjected to a bending test to carry out a qualitative stress analysis through the difference between the values of the birefringence obtained during loading (B) and before loading (Bo) in every point studied. With the obtained results it was possible to identify qualitative and quantitatively the regions of the specimen under compressive and tensile stresses. The execution of these tests allowed to verify the efficiency of

  7. Reactor pressure vessel status report

    International Nuclear Information System (INIS)

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

    1994-12-01

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

  8. Sealing method and sealing device for radioactive waste containing vessel

    International Nuclear Information System (INIS)

    Ishiwatari, Koji; Otsuki, Akira

    1998-01-01

    A radioactive waste-containing body is hoisted down into a strong-material vessel opened upwardly, and a strong-material lid is hoisted down to the opening of the strong-material-vessel and welded. The strong material vessel is hoisted up and loaded on a corrosion resistant-material bottom plate placed horizontally. A corrosion resistant-material vessel having one opening end and having a corrosion resistant-material flange on the other end and previously agreed with the strong material-vessel main body is hoisted up by a hoisting device having an inserting device so that the opening of the corrosion resistant vessel is directed downwardly. The corrosion resistant vessel is press-fitted to the outside of the strong material-vessel by the inserting device while being heated by a preheater to shrink. Subsequently, the lower end of the corrosion resistant-material vessel and the corrosion resistant-material bottom plate are welded to constitute a corrosion resistant-material vessel. Then, the radioactive waste containing body can be sealed in a sealing vessel comprising the strong-material vessel and the corrosion resistant-material vessel. (N.H.)

  9. Environmental safety issues for semiconductors (research on scarce materials recycling)

    International Nuclear Information System (INIS)

    Izumi, Shigekazu

    2004-01-01

    In the 21st century, in the fabrication of various industrial parts, particularly, current and future electronics devices in the semiconductor industry, environmental safety issues should be carefully considered. We coined a new term, environmental safety issues for semiconductors, considering our semiconductor research and technology which include environmental and ecological factors. The main object of this analysis is to address the present situation of environmental safety problems in the semiconductor industry; some of which are: (1) the generation and use of hazardous toxic gases in the crystal growth procedure such as molecular beam epitaxy (MBE) and metalorganic chemical vapor deposition (MOCVD), (2) the generation of industrial toxic wastes in the semiconductor process and (3) scarce materials recycling from wastes in the MBE and MOCVD growth procedure

  10. Present status of decommissioning materials reuse research at JAERI

    International Nuclear Information System (INIS)

    Fujiki, Kazuo; Nakamura, Hisashi; Kanazawa, Katsuo

    1991-01-01

    Rational treatment and disposal of a large volume of the dismantling wastes resulting from the reactor dismantling are the key to success to carry out the decommissioning smoothly. From this viewpoint, the Japan Atomic Energy Research Institute (JAERI) has been conducting development of the recycling technology for metal waste and a investigation study on the rational recycling system for the dismantling wastes recycling. With respect to the development of the recycling technology, melting tests using non-contaminated metals have been conducted and the basic characteristics of experimental facility and material balances understood. In the investigation study on the rational recycling system, review and discussion were made on the amount of waste arising from decommissioning a nuclear power plant, a scenario of recycling the wastes, and the necessary processing facilities. (author)

  11. Reactor pressure vessel steels

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  12. The Change of the Seebeck Coefficient Due to Neutron Irradiation and Thermal Fatigue of Nuclear Reactor Pressure Vessel Steel and its Application to the Monitoring of Material Degradation

    International Nuclear Information System (INIS)

    Niffenegger, M.; Reichlin, K.; Kalkhof, D.

    2002-05-01

    The monitoring of material degradation, that might be caused by neutron irradiation and thermal fatigue, is an important topic in lifetime extension of nuclear power plants. We therefore investigated the application of the Seebeck effect for determining material degradation of common reactor pressure vessel steel. The Seebeck coefficient (SC) of several irradiated Charpy specimens made from Japanese JRQ-steel were measured. The specimens suffered a fluence from 0 up to 4.5 x 10 19 neutrons per cm 2 with energies higher than 1 MeV. The measured changes of the SC within this range were about 500 nV, increasing continuously in the range under investigation. Some indications of saturation appeared at fluencies larger than 4.55 x 10 19 neutrons per cm 2 . We obtained a linear dependency between the SC and the temperature shift ΔT 41 of the Charpy-Energy- Temperature curve which is widely used to characterize material embrittlement. Similar measurements were performed on specimens made from the widely used austenitic steel X6CrNiTi18-10 (according to DIN 1.4541) that were fatigued by applying a cyclic strain amplitude of 0.28%. For this kind of fatigue the observed change of SC was somewhat smaller than for the irradiated specimens. Further investigations were made to quantify the size of the gage volume in which the thermoelectric power is generated. It appeared that the information gathered from a Thermo Electric Power (TEP) measurement is very local. To overcome this problem we propose a novel TEP-method using a Thermoelectric Scanning Microscope (TSM). We finally conclude that the change of the SC has a potential for monitoring of material degradation due to neutron irradiation and thermal fatigue, but it has to be taken into account that several influencing parameters could contribute to the TEP in either an additional or extinguishing manner. A disadvantage of the method is the requirement of a clean surface without any oxide layer. A part of this disadvantage can

  13. Research on utilization of isotopes for metallic materials

    International Nuclear Information System (INIS)

    Maebashi, Yoichi; Kagaya, Yutaka; Kametani, Hiroshi

    1983-01-01

    As the research on the utilization of unsealed radioisotopes for metallic materials, among the refining of nonferrous metals already carried out in the National Research Institute for Metals, the refining reaction of copper sulfide was taken up. In this refining reaction, it is important to know the oxidation behavior of sulfur in copper sulfide for improving the refining method. However in the oxidation of sulfur, the kinds of the oxides formed are many, and when copper and iron ions coexist as in this case, their separation and analysis are very difficult. The utilization of radioisotopes is required for identifying the oxidation products and the oxides in melt, and for identifying various compound ions. The solvent for thin layer chromatography was selected, and the effects exerted by the moving rate, concentration and coexisting elements of various sulfur acid ions on the thin layer of silica gel were clarified. In the suspension reaction of copper sulfide without a power source, it was elucidated that S 2 O 3 2- arose consistently from the initial stage of reaction, and the reaction equation was forecast. The melting state of sulfur in anode oxidation reaction was studied. (Kako, I.)

  14. Pressure vessel integrity 1991

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  15. NATO Advanced Research Workshop on Brilliant Light Facilities and Research in Life and Material Sciences

    CERN Document Server

    Tsakanov, Vasili; Brilliant Light in Life and Material Sciences

    2007-01-01

    The present book contains an excellent overview of the status and highlights of brilliant light facilities and their applications in biology, chemistry, medicine, materials and environmental sciences. Overview papers on diverse fields of research by leading experts are accompanied by the highlights in the near and long-term perspectives of brilliant X-Ray photon beam usage for fundamental and applied research. The book includes advanced topics in the fields of high brightness photon beams, instrumentation, the spectroscopy, microscopy, scattering and imaging experimental techniques and their applications. The book is strongly recommended for students, engineers and scientists in the field of accelerator physics, X-ray optics and instrumentation, life, materials and environmental sciences, bio and nanotechnology.

  16. Prediction of the Effects of Radiation FOr Reactor pressure vessel and in-core Materials using multi-scale modeling - 60 years foreseen plant lifetime (PERFORM-60 project)

    International Nuclear Information System (INIS)

    Al Mazouzi, A.; Bugat, S.; Leclercq, S.; Massoud, J.-P.; Moinereau, D.; Lidbury, D.; Van Dyck, S.; Marini, B.; Alamo, Ana

    2010-01-01

    The PERFECT project of the EURATOM framework program (FP6) is a first step through the development of a simulation platform that contains several advanced numerical tools aiming at the prediction of irradiation damage in both the reactor pressure vessel (RPV) and its internals using state-of-the-art knowledge. These tools allow simulation of irradiation effects on the microstructure and the constitutive behavior of the RPV low alloy steels, as well as their fracture mechanics properties. For the reactor internals, the first partial models were established, describing radiation damage to the microstructure and providing a first description of the stress corrosion behaviour of austenitic steels in primary environment, without physical linking of the radiation and corrosion effects. Thus, relying on the existing PERFECT Roadmap, the FP7 Collaborative Project PERFORM 60 has mainly for objective to develop similar tools that would allow the simulation of the combined effects of irradiation and corrosion on internals, in addition to a further improvement of the existing ones on RPV made of bainitic steels. From the managerial view point, PERFORM 60 is based on two technical sub-projects, namely (i) RPV and (ii) Internals. In addition, a Users' Group and a training scheme have been adopted in order to allow representatives of constructors, utilities, research organizations... from Europe, USA and Japan to participate actively in the process of appraising the limits and potentialities of the developed tools as well as their validation against qualified experimental data

  17. Materials and Molecular Research Division annual report 1983

    Energy Technology Data Exchange (ETDEWEB)

    Searcy, A.W.; Muller, R.H.; Peterson, C.V.

    1984-07-01

    Progress is reported in the following fields: materials sciences (metallurgy and ceramics, solid-state physics, materials chemistry), chemical sciences (fundamental interactions, processes and techniques), actinide chemistry, fossil energy, electrochemical energy storage systems, superconducting magnets, semiconductor materials and devices, and work for others. (DLC)

  18. Materials and Molecular Research Division annual report 1983

    International Nuclear Information System (INIS)

    Searcy, A.W.; Muller, R.H.; Peterson, C.V.

    1984-07-01

    Progress is reported in the following fields: materials sciences (metallurgy and ceramics, solid-state physics, materials chemistry), chemical sciences (fundamental interactions, processes and techniques), actinide chemistry, fossil energy, electrochemical energy storage systems, superconducting magnets, semiconductor materials and devices, and work for others

  19. Phenomenological vessel burst investigations

    International Nuclear Information System (INIS)

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

    1985-07-01

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

  20. Covalently bonded disordered thin-film materials. Materials Research Society symposium proceedings Volume 498

    International Nuclear Information System (INIS)

    Siegal, M.P.; Milne, W.I.; Jaskie, J.E.

    1998-01-01

    The current and potential impact of covalently bonded disordered thin films is enormous. These materials are amorphous-to-nanocrystalline structures made from light atomic weight elements from the first row of the periodic table. Examples include amorphous tetrahedral diamond-like carbon, boron nitride, carbon nitride, boron carbide, and boron-carbon-nitride. These materials are under development for use as novel low-power, high-visibility elements in flat-panel display technologies, cold-cathode sources for microsensors and vacuum microelectronics, encapsulants for both environmental protection and microelectronics, optical coatings for laser windows, and ultra-hard tribological coatings. researchers from 17 countries and a broad range of academic institutions, national laboratories and industrial organizations come together in this volume to report on the status of key areas and recent discoveries. More specifically, the volume is organized into five sections. The first four highlight ongoing work primarily in the area of amorphous/nanocrystalline (disordered) carbon thin films; theoretical and experimental structural characterization; electrical and optical characterizations; growth methods; and cold-cathode electron emission results. The fifth section describes the growth, characterization and application of boron- and carbon-nitride thin films

  1. Progress on research of materials science and biotechnology by ion beam application

    Energy Technology Data Exchange (ETDEWEB)

    Ishigaki, Isao [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment

    1997-03-01

    Research of materials science and biotechnology by ion beam application in Takasaki Establishment was reviewed. Especially, the recent progresses of research on semiconductors in space, creation of new functional materials and topics in biotechnology were reported. (author)

  2. Measurement of materialism and spiritualism in substance abuse research.

    Science.gov (United States)

    Mathew, R J; Mathew, V G; Wilson, W H; Georgi, J M

    1995-07-01

    A modified version of an instrument called the Mathew Materialism-Spiritualism Scale (MMSS), originally developed in India, was evaluated for possible use in substance abuse research in the U.S. The scale was administered to 62 individuals recovering from substance use, 20 clergy people and 61 general controls. Test-retest reliability for the MMSS was verified by administering it to 18 control subjects on two separate occasions, 7 days apart. The Pearson correlation for the MMSS total scores was 0.83 (p < .0001). Internal consistency was examined with Cronbach's alpha in the entire sample of 143 subjects; the result for the total score was .93. Factor analysis showed a factor structure compatible with the subscales proposed by the developer. Women, in general, obtained higher spirituality scores. Members of the recovering group obtained significantly higher scores on "character" and "mysticism" than the general controls. When general controls were divided into MAST positive and MAST negative individuals, the MAST positive group obtained lower scores than the recovering group for "God," "mysticism" and "character." MAST negative individuals had lower scores on "mysticism" than the recovering group. Christians had higher scores on "God" and "religion" subscales than did nonChristians and agnostics. The results of this study need confirmation using an improved methodology and larger sample sizes. However, they suggest that the scale may be useful for the study of spirituality in the U.S.

  3. Research process of nondestructive testing pitting corrosion in metal material

    Directory of Open Access Journals (Sweden)

    Bo ZHANG

    2017-12-01

    Full Text Available Pitting corrosion directly affects the usability and service life of metal material, so the effective nondestructive testing and evaluation on pitting corrosion is of great significance for fatigue life prediction because of data supporting. The features of pitting corrosion are elaborated, and the relation between the pitting corrosion parameters and fatigue performance is pointed out. Through introducing the fundamental principles of pitting corrosion including mainly magnetic flux leakage inspection, pulsed eddy current and guided waves, the research status of nondestructive testing technology for pitting corrosion is summarized, and the key steps of nondestructive testing technologies are compared and analyzed from the theoretical model, signal processing to industrial applications. Based on the analysis of the signal processing specificity of different nondestructive testing technologies in detecting pitting corrosion, the visualization combined with image processing and signal analysis are indicated as the critical problems of accurate extraction of pitting defect information and quantitative characterization for pitting corrosion. The study on non-contact nondestructive testing technologies is important for improving the detection precision and its application in industries.

  4. ISS Material Science Research Rack HWIL Interface Simulation

    Science.gov (United States)

    Williams, Philip J.; Ballard, Gary H.; Crumbley, Robert T. (Technical Monitor)

    2002-01-01

    In this paper, the first Material Science Research Rack (MSRR-1) hardware-in-the-loop (HWIL) interface simulation is described. Dynamic Concepts developed this HWIL simulation system with funding and management provided by the Flight Software group (ED14) of NASA-MSFC's Avionics Department. The HWIL system has been used both as a flight software development environment and as a software qualification tool. To fulfill these roles, the HWIL simulator accurately models the system dynamics of many MSRR-1 subsystems and emulates most of the internal interface signals. The modeled subsystems include the Experiment Modules, the Thermal Environment Control System, the Vacuum Access System, the Solid State Power Controller Module, and the Active Rack Isolation Systems. The emulated signals reside on three separate MIL-STD-1553B digital communication buses, the ISS Medium Rate Data Link, and several analog controller and sensor signals. To enhance the range of testing, it was necessary to simulate several off-nominal conditions that may occur in the interfacing subsystems.

  5. State of opening the cover and carrying out the checkup of the reactor vessel of the nuclear-powered ship 'Mutsu' by Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    1989-01-01

    In the checkup by opening the cover of the reactor vessel of the nuclear-powered ship 'Mutsu', Japan Atomic Energy Research Institute carried out the checkup and maintenance for the reactor proper, control system and primary coolant facilities including the secondary side of steam generators and the pressure balancing valve of the containment vessel. The works were classified into the opening of the reactor, checkup, maintenance and restoration. The opening was begun on August 4, 1988, and finished on December 5. The checkup and maintenance were begun on September 22, and are still continued now. The maximum radiation dose rate on the surfaces of fuel assemblies and control rods and at the positions 1 m distant from them was measured. The results of the checkup of various components are reported. In 290 absorbent rods of control rods, spot corrosion and discoloration were observed, of which the spot corrosion penetrated the walls of 4 rods. Also in 12 fuel rods, spot corrosion was observed near the welded end plugs, but leak was not observed. (K.I.)

  6. Expanded Fermilab pressure vessel directory program

    Energy Technology Data Exchange (ETDEWEB)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect.

  7. Expanded Fermilab pressure vessel directory program

    International Nuclear Information System (INIS)

    Tanner, A.

    1983-01-01

    Several procedures have been written to manage the information pertaining to the vacuum tanks and pressure vessels for which the laboratory is responsible. These procedures have been named TANK1 for the vessels belonging to the Accelerator Division, TANK2 and TANK3 for the vessels belonging to the Research Division and to Technical Support respectively, and TANK4 for the vessels belonging to the Business Division. The operating procedures are otherwise identical in every respect

  8. New Optical Sensing Materials for Application in Marine Research

    Science.gov (United States)

    Borisov, S.; Klimant, I.

    2012-04-01

    Optical chemosensors are versatile analytical tools which find application in numerous fields of science and technology. They proved to be a promising alternative to electrochemical methods and are applied increasingly often in marine research. However, not all state-of-the- art optical chemosensors are suitable for these demanding applications since they do not fully fulfil the requirements of high luminescence brightness, high chemical- and photochemical stability or their spectral properties are not adequate. Therefore, development of new advanced sensing materials is still of utmost importance. Here we present a set of novel optical sensing materials recently developed in the Institute of Analytical Chemistry and Food Chemistry which are optimized for marine applications. Particularly, we present new NIR indicators and sensors for oxygen and pH which feature high brightness and low level of autofluorescence. The oxygen sensors rely on highly photostable metal complexes of benzoporphyrins and azabenzoporphyrins and enable several important applications such as simultaneous monitoring of oxygen and chlorophyll or ultra-fast oxygen monitoring (Eddy correlation). We also developed ulta-sensitive oxygen optodes which enable monitoring in nM range and are primary designed for investigation of oxygen minimum zones. The dynamic range of our new NIR pH indicators based on aza-BODIPY dyes is optimized for the marine environment. A highly sensitive NIR luminescent phosphor (chromium(III) doped yttrium aluminium borate) can be used for non-invasive temperature measurements. Notably, the oxygen, pH sensors and temperature sensors are fully compatible with the commercially available fiber-optic readers (Firesting from PyroScience). An optical CO2 sensor for marine applications employs novel diketopyrrolopyrrol indicators and enables ratiometric imaging using a CCD camera. Oxygen, pH and temperature sensors suitable for lifetime and ratiometric imaging of analytes

  9. Materials research symposium 1988 of the Federal German Ministry of Research and Technology (BMFT). Proceedings and posters. Vol. 1

    International Nuclear Information System (INIS)

    1988-01-01

    In the context of concentrating the research activities on key areas of technology, the West German Ministry of Research and Technology started the materials research program in 1985. Long-term and risky questions of modern materials research were and are being tackled, using the instrument of combined project work, i.e.: the partnership of industry and scientific institutions. Three years after the start of the program, the technological state in West Germany in the field of new materials is to be documented and balanced by the 'Symposium on Materials Research'. Results of basic research to application orientated material developments are introduced by survey and detailed articles. The following subjects are dealt with in the first two volumes: 1. Functional polymers; 2. Structural polymers; 3. Metal materials; 4. Ceramics. 22 articles are listed separately in the 'ENERGY' databank. (orig./MM) [de

  10. 78 FR 35638 - Certificate of Alternative Compliance for the NOAA Research Vessel FSV-6 RUBEN LASKER, 9664988

    Science.gov (United States)

    2013-06-13

    ...'' box, and then clicking ``Search.'' FOR FURTHER INFORMATION CONTACT: If you have questions on this.... If you have questions on viewing or submitting material to the docket, call Barbara Hairston, Program... with Annex I of the Inland Rules Act. The Commandant, U.S. Coast Guard, certifies that full compliance...

  11. Scientific Applications of Optical Instruments to Materials Research

    Science.gov (United States)

    Witherow, William K.

    1997-01-01

    Microgravity is a unique environment for materials and biotechnology processing. Microgravity minimizes or eliminates some of the effects that occur in one g. This can lead to the production of new materials or crystal structures. It is important to understand the processes that create these new materials. Thus, experiments are designed so that optical data collection can take place during the formation of the material. This presentation will discuss scientific application of optical instruments at MSFC. These instruments include a near-field scanning optical microscope, a miniaturized holographic system, and a phase-shifting interferometer.

  12. The analysis of reactor vessel surveillance program data

    International Nuclear Information System (INIS)

    Norris, E.B.

    1979-01-01

    Commercial nuclear power reactor vessel surveillance programs are provided by the reactor supplier and are designed to meet the requirements of ASTM Method E 185. (3). Each surveillance capsule contains sets of Charpy V-notch (Csub(v)) specimens representing selected materials from the vessel beltline region and some reference steel, tension test specimens machined from selected beltline materials, temperature monitors, and neutron flux dosimeters. Surveillance capsules may also contain fracture mechanics specimens machined from selected vessel beltline materials. The major steps in the conduct of a surveillance program include (1) the testing of the surveillance specimens to determine the exposure conditions at the capsule location and the resulting embrittlement of the vessel steel, (2) the extrapolation of the capsule results to the pressure vessel wall, and (3) the determination of the heatup and cooldown limits for normal, upset, and test operation. This paper will present data obtained from commercial light water reactor surveillance programs to illustrate the methods of analysis currently in use at Southwest Research Institute and to demonstrate some of the limitations imposed by the data available. Details concerning the procedures for testing the surveillance capsule specimens will not be included because they are considered to be outside of the scope of this paper

  13. Materials Science Research Rack-1 Fire Suppressant Distribution Test Report

    Science.gov (United States)

    Wieland, P. O.

    2002-01-01

    Fire suppressant distribution testing was performed on the Materials Science Research Rack-1 (MSRR-1), a furnace facility payload that will be installed in the U.S. Lab module of the International Space Station. Unlike racks that were tested previously, the MSRR-1 uses the Active Rack Isolation System (ARIS) to reduce vibration on experiments, so the effects of ARIS on fire suppressant distribution were unknown. Two tests were performed to map the distribution of CO2 fire suppressant throughout a mockup of the MSRR-1 designed to have the same component volumes and flowpath restrictions as the flight rack. For the first test, the average maximum CO2 concentration for the rack was 60 percent, achieved within 45 s of discharge initiation, meeting the requirement to reach 50 percent throughout the rack within 1 min. For the second test, one of the experiment mockups was removed to provide a worst-case configuration, and the average maximum CO2 concentration for the rack was 58 percent. Comparing the results of this testing with results from previous testing leads to several general conclusions that can be used to evaluate future racks. The MSRR-1 will meet the requirements for fire suppressant distribution. Primary factors that affect the ability to meet the CO2 distribution requirements are the free air volume in the rack and the total area and distribution of openings in the rack shell. The length of the suppressant flowpath and degree of tortuousness has little correlation with CO2 concentration. The total area of holes in the rack shell could be significantly increased. The free air volume could be significantly increased. To ensure the highest maximum CO2 concentration, the PFE nozzle should be inserted to the stop on the nozzle.

  14. R.M.S Titanic 2003 Expedition on the Russian Research Vessel Akademik Mstislav Keldysh between 20030622 and 20030702

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — As the leading ocean agency, and as per the Guidelines for Research, Exploration and Salvage of RMS Titanic, issued under the authority of the RMS Titanic Maritime...

  15. 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...

  16. Needs in Research and Development on materials for the gas coolant nuclear system: HTR/VHTR and GFR

    International Nuclear Information System (INIS)

    Billot, Ph.

    2003-01-01

    This presentation takes stock on the materials for high temperature reactors HTR (850 C), very high temperature VHTR(>1000 C) and fast neutrons high temperature GGF(850 C). It concerns the welding materials for the vessel, Ni-based superalloys for gas turbines, coatings, graphite, ceramics and corrosion studies. (A.L.B.)

  17. Materials Science: A Spin-off in Research and Development

    International Nuclear Information System (INIS)

    Aduda, B.O.

    2006-01-01

    The scope materials science is wide since it is a multi/trans-disciplinary subject, and is based on physics and chemistry of solid state. It embraces all aspects of engineering materials, from the most basic to the most novel, and is concerned with how a material is assembled from the basic units, can be used, can be modified or improved to perform specific tasks. Further, it is concerned with proper selection of materials for specific applications, and development of new and improved materials with unique properties for the ever increasing and more demanding applications, e.g., aerogels, ceramic membranes for fuel cells, bioceramics for hip bone replacements, nanostructured photoactive thin films for solar cell, sensors and photocatalysis applications etc

  18. PIE technology on mechanical tests for HTTR core component and structural materials developed at Research Hot Laboratory

    International Nuclear Information System (INIS)

    Kizaki, Minoru; Honda, Junichi; Usami, Kouji; Ouchi, Asao; Oeda, Etsuro; Matsumoto, Seiichiro

    2001-02-01

    The high temperature engineering test reactor (HTTR) with the target operation temperature of 950degC established the first criticality on November, 1998 based on a large amount of R and D results on fuel and materials. In such R and D works, the development of reactor materials are one of the key issues from the view point of reactor environments such as extremely high temperature, neutron irradiation and so on for the HTTR. The Research Hot Laboratory (RHL) had carried out much kind of post irradiation examinations (PIEs) on core component and pressure vessel materials for during more than a quarter century. And obtained data played an important role in development, characterization and licensing of those materials for the HTTR. This paper describes the PIE technology developed at RHL and typical results on mechanical tests such as elevated temperature tensile and creep rupture tests for Hasteloy-X, Incolloy 800H and so on, and Charpy impact, J IC fracture toughness, K Id fracture toughness and small punch tests for normalized and tempered 2 1/4Cr-1Mo steel from historical view. In addition, an electrochemical test technique established for investigating the irradiation embrittlement mechanism on 2 1/4Cr-1Mo steel is also mentioned. (author)

  19. University of Illinois at Urbana-Champaign, Materials Research Laboratory progress report for FY 1991

    International Nuclear Information System (INIS)

    1991-10-01

    The Materials Research Laboratory at the University of Illinois is an interdisciplinary laboratory operated in the College of Engineering. Its focus is the science of materials and it supports research in the areas of condensed matter physics, solid state chemistry, and materials science. This report addresses topics such as: an MRL overview; budget; general programmatic and institutional issues; new programs; research summaries for metallurgy, ceramics, solid state physics, and materials chemistry

  20. University of Illinois at Urbana-Champaign, Materials Research Laboratory progress report for FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    1991-10-01

    The Materials Research Laboratory at the University of Illinois is an interdisciplinary laboratory operated in the College of Engineering. Its focus is the science of materials and it supports research in the areas of condensed matter physics, solid state chemistry, and materials science. This report addresses topics such as: an MRL overview; budget; general programmatic and institutional issues; new programs; research summaries for metallurgy, ceramics, solid state physics, and materials chemistry.

  1. Material test reactor fuel research at the BR2 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

    2012-03-15

    The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

  2. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  3. Application of annealing for WWER vessels life extension

    International Nuclear Information System (INIS)

    Badanin, V.I.; Gorynin, I.V.; Nickolaev, V.A.; Dragunov, Y.G.; Fedorov, V.G.

    1989-01-01

    Safe operation of NPP is greatly dependent on the guarantee of reactor vessel brittle failure strength with account for the effect of radiation embrittlement of vessel material. Recovery of irradiated material properties is principally important way to extend radiation life of reactor vessel. The aim of this report is to demonstrate the efficiency of annealing for recovery of vessel material properties and extension of its service-life

  4. Achievement report for fiscal 1981 on Sunshine Program-assisted project. Research and development of materials for coal liquefaction plant (Research on materials); 1981 nendo sekitan ekika plant zairyo no kenkyu kaihatsu (zairyo no kenkyu) seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-03-01

    Research and development is started of materials for constituting a liquefaction reactor which is the primary apparatus in the coal liquefaction process. Tested in the basic research on the designing of alloys for the reactor are Cr-Mo-V-Ti based alloys prepared by adding V, Ti, and B to Cr-Mo based alloys presently in use as materials for pressure vessels. They are melted and 50kg test ingots of steel are prepared, and an alloy with a constitution of Low Si-3%Cr-1%Mo-0.25%V-0.02%Ti-0.002%B is selected. In the designing of alloys to serve as stainless steel overlay welding materials and in the development of basic welding technologies, studies are conducted in search for materials that exhibit excellent durability under the conditions of liquefaction reaction and for technologies in this connection. These are accomplished using a test device installed this fiscal year, which is capable of reproducing high-temperature high-pressure hydrogen atmosphere. As the result, basic technologies are acquired, which will enable the achievement of the goal. In the study of on-site production technologies using medium-size ingots as the materials, an 85-ton medium size steel ingot of a new chemical constitution of Low Si-3Cr-1Mo-1/4V-Ti-B is melted, and a shell which is 400mm in thickness and 2m in internal diameter is forged, and the shell is subjected to a property and performance qualification test. (NEDO)

  5. Pressure vessel lid

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. Research and applications of N-halamine biocidal materials

    Directory of Open Access Journals (Sweden)

    KANG Zhenzhen

    2012-10-01

    Full Text Available N-halamines,a new class of biocides,overcome some of the disadvantages caused by the traditional biocides in practical applications.They are environmentally friendly germicides due to their fast and efficient sterilization,storage stability,and regeneration.Earlier studies on N-halamines mainly focused on the syntheses and applications of small molecular organic N-halamines such as fivemembered and six-membered heterocyclic N-halamine compounds.Compared to traditional inorganic halogen-containing disinfectants such as chlorine gas,sodium hypochlorite,chlorine dioxide,these heterocyclic N-halamines can maintain disinfection capacity in the water for longer time due to their better stability.Since the late 20th century,non-leaching biocial N-halamine materials have received much attention.Some novel N-halmine precursors with binding groups have been covalently bounded to various materials such as cellulose fiber,silica gel,polystyrene,polyethylene,and polyurethane to produce nonleaching biocidal materials.Specially,the successful development of macroporous cross-linked N-halamine polymer resin materials (Halopure and related technologies created a new era for the applications of N-halamine materials in the disinfection of drinking water.In this review paper,the antibacterial mechanism and synthetic methods of N-halamine biocidal materials and their application prospects in various fields of daily life were introduced.Their development prospects were also made.

  7. Compilation of reports from research supported by the Materials Engineering Branch, Division of Engineering: 1965--1990

    International Nuclear Information System (INIS)

    Hiser, A.L.

    1991-05-01

    Since 1965, the Materials Engineering Branch, Division of Engineering, of the Nuclear Regulatory Commission's Office of Nuclear Regulatory Research, and its predecessors dating back to the Atomic Energy Commission (AEC), has sponsored research programs concerning the integrity of the primary system pressure boundary of light water reactors. The components of concern in these research programs have included the reactor pressure vessel (RPV), steam generators, and the piping. These research programs have covered a broad range of topics, including fracture mechanics analysis and experimental work for RPV and piping applications, inspection method development and qualification, and evaluation of irradiation effects to RPV steels. This report provides as complete a listing as practical of formal technical reports submitted to the NRC by the investigators working on these research programs. This listing includes topical, final and progress reports, and is segmented by topic area. In many cases a report will cover several topics (such as in the case of progress reports of multi-faceted programs), but is listed under only one topic. Therefore, in searching for reports on a specific topic, other related topic areas should be checked also

  8. Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. Final Report of a Coordinated Research Project 2008-2012

    International Nuclear Information System (INIS)

    2014-11-01

    The IAEA supports Member States in the area of advanced fast reactor technology development by providing a major fulcrum for information exchange and collaborative research programmes. The IAEA’s activities in this field are mainly carried out within the framework of the Technical Working Group on Fast Reactors (TWG-FR), which assists in the implementation of corresponding IAEA support, and ensures that all technical activities are in line with expressed needs of Member States. Among this broad range, the IAEA proposes and establishes coordinated research projects (CRPs), aimed at improving Member State capability in fast reactor design and analysis. An important opportunity to perform collaborative research activities was provided by the system startup tests carried out by the Japan Atomic Energy Agency (JAEA) in the prototype loop type sodium cooled fast reactor Monju, in particular a turbine trip test performed in December 1995. As the JAEA opened the experimental dataset to international collaboration in 2008, the IAEA launched the CRP on Benchmark Analyses of Sodium Natural Convection in the Upper Plenum of the Monju Reactor Vessel. The CRP, together with eight institutes from seven States, has contributed to improving capabilities in sodium cooled fast reactors simulation through code verification and validation, with particular emphasis on thermal stratification and natural circulation phenomena

  9. The research of establishing reactor materials thermophysical properties data base

    International Nuclear Information System (INIS)

    Luo Danhui; Zhong Jianguo; Zhang Lili; Zhao Yongming

    1992-01-01

    In the process of nuclear reactor design and safety analysis, the reactor materials thermophysical properties parameters are very important as the main input data of reactor design and calculation. The goal of this work is to establish a practical, reliable data base of reactor materials thermophysical properties parameters with obvious function in reactor design, operation and safety analysis. At present phase, the focal point of this data base is to collect the materials thermophysical properties data based on the need of safety analysis in light water reactor and heavy water reactor. The materials to be chosen are as follows: Uranium, U-Al alloy, UO 2 , UO 2 -PuO 2 mixture, Zr-2, Zr-4, Zr-1% Ni alloy, Inconel-625, ZrO 2 (oxidic layer), boron carbide, cadmium in stainless steel, silver-indium-cadmium alloy, light water and heavy water, etc. The following thermophysical properties parameters are mainly included in the data base: thermal conductivity, thermal diffusivity, specific heat capacity, heat of melting, coefficient of thermal expansion, emittance, density, heat of vaporization, kinematic viscosity etc. The first phase of this work has been finished, which includes the method of establishing reactor materials thermophysical properties data base, the requirement of data collection, the requirement of establishing data base and the method of the data evaluation. This data base has been established and used on PC computer

  10. Research Update: Computational materials discovery in soft matter

    Directory of Open Access Journals (Sweden)

    Tristan Bereau

    2016-05-01

    Full Text Available Soft matter embodies a wide range of materials, which all share the common characteristics of weak interaction energies determining their supramolecular structure. This complicates structure-property predictions and hampers the direct application of data-driven approaches to their modeling. We present several aspects in which these methods play a role in designing soft-matter materials: drug design as well as information-driven computer simulations, e.g., histogram reweighting. We also discuss recent examples of rational design of soft-matter materials fostered by physical insight and assisted by data-driven approaches. We foresee the combination of data-driven and physical approaches a promising strategy to move the field forward.

  11. Safeguards research: assessing material control and accounting systems

    International Nuclear Information System (INIS)

    Maimoni, A.

    1977-01-01

    The Laboratory is working for the Nuclear Regulatory Commission to improve the safeguarding of special nuclear material at nuclear fuel processing facilities, to provide a basis for improved regulations for material control and accounting systems, and to develop an assessment procedure for verifying compliance with these regulations. Early work included setting up a hierarchy of safeguard objectives and a set of measurable parameters with which systems performance to meet those objectives can be measured. Present work has focused on developing a computerized assessment procedure. We have also completed a test bed (based on a plutonium nitrate storage area) to identify and correct problems in the procedure and to show how this procedure can be used to evaluate the performance of an applicant's material control and accounting system

  12. On The Research Of The Materials Corrosion Stability

    International Nuclear Information System (INIS)

    Bolotnikov, A.

    1998-01-01

    The contact of chemically active working medium with machine parts results in their corrosion damage. The problem is especially actual when a plant destined to work with another medium is used. Protracted reliable operation of these machines can be guaranteed only by a correct selection of part materials which ensures both their high corrosion stability in the given medium and necessary strength under working conditions. Resistance of materials to corrosion (including that are known as rust-resisting ones) essentially depends on the reagent type. Literature contains limited amount of information about materials behavior in the given medium. Necessity of such information even on the initial stage of design demands an effective method of the fast corrosion stability examination. The low rate of the chemical reaction under normal conditions leads to difficulties while discovering such method. This paper is dedicated to solution of the foregoing problem. Theoretical grounds, descriptions of experimental plant, and results of the test are adduced

  13. Neutron Assay System for Confinement Vessel Disposition

    International Nuclear Information System (INIS)

    Frame, Katherine C.; Bourne, Mark M.; Crooks, William J.; Evans, Louise; Mayo, Douglas R.; Miko, David K.; Salazar, William R.; Stange, Sy; Valdez, Jose I.; Vigil, Georgiana M.

    2012-01-01

    Los Alamos National Laboratory has a number of spherical confinement vessels (CVs) remaining from tests involving nuclear materials. These vessels have an inner diameter of 6 feet with 1-inch thick steel walls. The goal of the Confinement Vessel Disposition (CVD) project is to remove debris and reduce contamination inside the CVs. The Confinement Vessel Assay System (CVAS) was developed to measure the amount of special nuclear material (SNM) in CVs before and after cleanout. Prior to cleanout, the system will be used to perform a verification measurement of each vessel. After cleanout, the system will be used to perform safeguards-quality assays of (le)100-g 239 Pu equivalent in a vessel for safeguards termination. The CVAS has been tested and calibrated in preparation for verification and safeguards measurements.

  14. Combinatorial methods for advanced materials research and development

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, R.; Dondorf, S.; Hauck, M.; Horbach, D.; Kaiser, M.; Krysta, S.; Kyrylov, O.; Muenstermann, E.; Philipps, M.; Reichert, K.; Strauch, G. [Rheinisch-Westfaelische Technische Hochschule Aachen (Germany). Lehrstuhl fuer Theoretische Huettenkunde

    2001-10-01

    The applicability of combinatorial methods in developing advanced materials is illustrated presenting four examples for the deposition and characterization of one- and two-dimensionally laterally graded coatings, which were deposited by means of (reactive) magnetron sputtering and plasma-enhanced chemical vapor deposition. To emphasize the advantages of combinatorial approaches, metastable hard coatings like (Ti,Al)N and (Ti,Al,Hf)N respectively, as well as Ge-Sb-Te based films for rewritable optical data storage were investigated with respect to the relations between structure, composition, and the desired materials properties. (orig.)

  15. The role of materials in controlled thermonuclear research

    Energy Technology Data Exchange (ETDEWEB)

    Craston, J L; Hancox, R; Robson, A E [U.K. Atomic Energy Authority, AERE, Harwell (United Kingdom); Kaufman, S; Miles, H T; Ware, A A; Wesson, J A [AEI Research Laboratory, Aldermaston (United Kingdom)

    1958-07-01

    It is the purpose of this paper to examine the processes occurring at the wall and to discuss their importance in the choice of materials both for present equipment and for future designs. The emphasis is laid primarily on plasma contamination but other effects are considered, such as thermal stress fatigue and radiation damage of the wall. The principal problems associated with the choice of wall material for a high current discharge tube have been discussed, both under the conditions which exist in present systems and under the conditions which are anticipated in a thermonuclear reactor.

  16. Report on current research into organic materials in radioactive waste

    International Nuclear Information System (INIS)

    Norris, G.H.

    1987-11-01

    A preliminary review of relevant recent papers on organic materials in radioactive waste is presented. In particular, the effects of chelating or complexing agents, the influence of bacteria and the role of colloids are assessed. The requirement for further radioactive waste inventory detail is indicated. Potential problem areas associated with the presence of organic materials in radioactive waste are identified and appropriate experimental work to assess their significance is proposed. Recommendations for specific further work are made. A list and diagrams of some of the more important polymer structures likely to be present in radioactive waste and their possible degradation products are appended. (author)

  17. Research on the preparation, biocompatibility and bioactivity of magnesium matrix hydroxyapatite composite material.

    Science.gov (United States)

    Linsheng, Li; Guoxiang, Lin; Lihui, Li

    2016-08-12

    In this paper, magnesium matrix hydroxyapatite composite material was prepared by electrophoretic deposition method. The optimal process parameters of electrophoretic deposition were HA suspension concentration of 0.02 kg/L, aging time of 10 days and voltage of 60 V. Animal experiment and SBF immersion experiment were used to test the biocompatibility and bioactivity of this material respectively. The SD rats were divided into control group and implant group. The implant surrounding tissue was taken to do tissue biopsy, HE dyed and organizational analysis after a certain amount of time in the SD rat body. The biological composite material was soaked in SBF solution under homeothermic condition. After 40 days, the bioactivity of the biological composite material was evaluated by testing the growth ability of apatite on composite material. The experiment results showed that magnesium matrix hydroxyapatite biological composite material was successfully prepared by electrophoretic deposition method. Tissue hyperplasia, connective tissue and new blood vessels appeared in the implant surrounding soft tissue. No infiltration of inflammatory cells of lymphocytes and megakaryocytes around the implant was found. After soaked in SBF solution, a layer bone-like apatite was found on the surface of magnesium matrix hydroxyapatite biological composite material. The magnesium matrix hydroxyapatite biological composite material could promot calcium deposition and induce bone-like apatite formation with no cytotoxicity and good biocompatibility and bioactivity.

  18. Opening of new field in material science and technology by materials irradiation research

    Energy Technology Data Exchange (ETDEWEB)

    Kurishita, Hiroaki [Tohoku Univ., Sendai (Japan). Inst. for Materials Research

    1998-03-01

    It is believed that high energy particle irradiation causes severe degradation of materials, and great efforts have been made to reveal the underlying mechanism of such degradation. However, recent progress of the developments of irradiation rigs performed in the Japan Materials Testing Reactor (JMTR) and materials fabrication techniques has enabled to change our understanding of radiation effects on materials from the above pessimistic one to the very challenging one, i.e., irradiation has the beneficial effect of producing new phenomena and/or innovative materials that will not be available without irradiation. An example to be noted is that irradiation with neutrons in JMTR greatly improved the ductility of less ductile metals. This ductility improvement due to irradiation is directly opposite to irradiation embrittlement and is called radiation induced ductilization (RIDU). In this presentation the significance of RIDU and its mechanism will be stated. (author)

  19. 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)

  20. Materials and Molecular Research Division. Annual report 1981

    International Nuclear Information System (INIS)

    1982-08-01

    Progress is reported in the areas of materials sciences, chemical sciences, nuclear sciences, fossil energy, advanced (laser) isotope separation technology, energy storage, superconducting magnets, and nuclear waste management. Work for others included phase equilibria for coal gasification products and β-alumina electrolytes for storage batteries