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Sample records for iter candidate materials

  1. Compatibility of ITER candidate structural materials with static gallium

    International Nuclear Information System (INIS)

    Luebbers, P.R.; Michaud, W.F.; Chopra, O.K.

    1993-12-01

    Tests were conducted on the compatibility of gallium with candidate structural materials for the International Thermonuclear Experimental Reactor, e.g., Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, as well as Armco iron, Nickel 270, and pure chromium. Type 316 stainless steel is least resistant to corrosion in static gallium and Nb-5 Mo-1 Zr alloy is most resistant. At 400 degrees C, corrosion rates are ∼4.0, 0.5, and 0.03 mm/yr for type 316 SS, Inconel 625, and Nb-5 Mo- 1 Zr alloy, respectively. The pure metals react rapidly with gallium. In contrast to findings in earlier studies, pure iron shows greater corrosion than nickel. The corrosion rates at 400 degrees C are ≥88 and 18 mm/yr, respectively, for Armco iron and Nickel 270. The results indicate that at temperatures up to 400 degrees C, corrosion occurs primarily by dissolution and is accompanied by formation of metal/gallium intermetallic compounds. The solubility data for pure metals and oxygen in gallium are reviewed. The physical, chemical, and radioactive properties of gallium are also presented. The supply and availability of gallium, as well as price predictions through the year 2020, are summarized

  2. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1993-01-01

    Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

  3. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

    1994-01-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

  4. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  5. ITER status, design and material objectives

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

    During the ITER Engineering Design Activities (EDA), completed in July 2001, the Joint Central Team and Home Teams developed a robust design of ITER, summarised in this paper, with parameters which fully meet the required scientific and technological objectives, construction costs and safety requirements, with appropriate margins. The design is backed by R and D to qualify the technology, including materials R and D. Materials for ITER components have been selected largely because of their availability and well-established manufacturing technologies, taking account of the low fluence experienced during neutron irradiation, and the experimental nature of the device. Nevertheless, for specific needs relevant to a future fusion reactor, improved materials, in particular for magnet structures, in-vessel components, and joints between the different materials needed for plasma facing components, have been successfully developed. Now, with the technical readiness to decide on ITER construction, negotiations, supported by coordinated technical activities of an international team and teams from participant countries, are underway on joint construction of ITER with a view to the signature/ratification of an agreement in 2003

  6. Extending ITER materials design to welded joints

    Energy Technology Data Exchange (ETDEWEB)

    Tavassoli, A.-A.F. [DMN/Dir, CEA/Saclay, Commissariat a l' Energie Atomique, 91191 Gif sur Yvette cedex (France)]. E-mail: tavassoli@cea.fr

    2007-08-01

    This paper extends the ITER materials properties documentation to weld metals and incorporates the needs of Test Blanket Modules for higher temperature materials properties. Since the main structural material selected for ITER is type 316L(N)-IG, the paper is focused on weld metals and joining techniques for this steel. Materials properties data are analysed according to the French design and construction rules for nuclear components (RCC-MR) and design allowables are equally derived using the same rules. Particular attention is paid to the type of weld metal, to the type and position of welding and their influence on the materials properties data and design allowables. The primary goal of this work, starting with 19-12-2 weld metal, is to produce comprehensive materials properties documentations that when combined with codification and inspection documents would satisfy ITER licensing needs. As a result, structural stability and capability of welded joints during manufacturing of ITER components and their subsequent service, including the effects of irradiation and eventual incidental or accidental situations, are also covered.

  7. ITER plasma facing materials. Some critical considerations

    International Nuclear Information System (INIS)

    Barabash, V.; Dietz, K.J.; Federici, G.; Janeschitz, G.; Matera, R.; Tanaka, S.

    1995-01-01

    The description of current status with the choice of materials for ITER plasma facing components is presented. The main problem with lifetime of divertor elements is the particle and energy-induced erosion of armour materials. A solution for the first operation phase consists in using Be as an armour for the first wall and the divertor, however other possible materials (e.g. W) could be considered. (orig.)

  8. Neutron irradiation behavior of ITER candidate beryllium grades

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

    1998-01-01

    Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

  9. Materials issues in the design of the ITER first wall, blanket, and divertor

    International Nuclear Information System (INIS)

    Mattas, R.F.; Smith, D.L.; Wu, C.H.; Shatalov, G.

    1992-01-01

    During the ITER conceptual design study, a property data base was assembled, the key issues were identified, and a comprehensive R ampersand D plan was formulated to resolve these issues. The desired properties of candidate ITER divertor, first wall, and blanket materials are briefly reviewed, and the major materials issues are presented. Estimates of the influence of materials properties on the performance limits of the first wall, blanket, and divertor are presented

  10. Influence of materials choice on occupational radiation exposure in ITER

    International Nuclear Information System (INIS)

    Forty, C.B.A.; Firth, J.D.; Butterworth, G.J.

    1998-01-01

    In fission reactor plant, the radiation doses associated with inspection and maintenance of the primary cooling circuit account for a substantial fraction of the collective occupational radiation exposure (ORE). Similarly, it is anticipated that much of the ORE occurring during normal operation of ITER will arise from active deposits in the cooling loop. Using a number of calculation steps ranging from neutron activation analysis, mobilisation and transport modelling and Monte Carlo simulation, estimates for the gamma photon flux and radiation dose fields around a typical 'hot-leg' cooling pipe have been made taking SS316, OPTSTAB, MANET-II and F-82H steels as alternative candidate loop materials. (orig.)

  11. Updated candidate list for engineered barrier materials

    International Nuclear Information System (INIS)

    McCright, R.D.

    1995-10-01

    This report describes candidate materials to be evaluated over the next several years during advanced design phases for the waste package to be used for the underground disposal of high-level radioactive wastes at the Yucca Mountain facility

  12. Characterization of nanoparticles as candidate reference materials

    International Nuclear Information System (INIS)

    Martins Ferreira, E.H.; Robertis, E. de; Landi, S.M.; Gouvea, C.P.; Archanjo, B.S.; Almeida, C.A.; Araujo, J.R. de; Kuznetsov, O.; Achete, C.A.

    2013-01-01

    We report the characterization of three different nanoparticles (silica, silver and multi-walled carbon nanotubes) as candidate reference material. We focus our analysis on the size distribution of those particles as measured by different microscopy techniques. (author)

  13. The materials selection in ITER and the first materials workshop

    International Nuclear Information System (INIS)

    Matera, R.; Barabash, V.; Kalinin, G.; Tanaka, S.

    1998-01-01

    The selection of materials and joining technologies to be used in ITER is a trade-off between multiple and often conflicting requirements derived from the unique features of the fusion environment. Materials selection must encompass a total engineering approach, by considering not only physical and mechanical properties, but also the components' manufacturing, their maintainability and reliability, and, finally, how they can be recycled or disposed of at the end of machine operation

  14. Materials challenges for ITER - Current status and future activities

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V. [ITER International Team, Boltsmannstrasse 2, 85748 Garching (Germany)]. E-mail: valdimir.barabash@iter.org; Peacock, A. [EFDA Close Support Unit, 85748 Garching (Germany); Fabritsiev, S. [D.V. Efremov Scientific Research Institute, 196641 St. Petersburg (Russian Federation); Kalinin, G. [ENES, P.O. Box 788, 101000 Moscow (Russian Federation); Zinkle, S. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rowcliffe, A. [Metals and Ceramics Division, ORNL, P.O. Box 2008, Oak Ridge, TN 37831-6138 (United States); Rensman, J.-W. [NRG, P.O. Box 25, 1755 ZG Petten (Netherlands); Tavassoli, A.A. [Commissariat a l' Energie Atomique, CEA/Saclay, 91191 Gif sur Yvette cedex (France); Marmy, P. [CRPP, EPFL, Association EURATOM-Confederation Suisse, 5232, Villigen PSI (Switzerland); Karditsas, P.J. [EURATOM/UKAEA Fusion Association, Abingdon, Oxon OX14 3DB (United Kingdom); Gillemot, F. [AEKI Atomic Research Institute, 1121 Budapest, (Hungary); Akiba, M. [JAEA, Naka-machi, Naka-gun, Ibaraki-ken 311-0193 (Japan)

    2007-08-01

    ITER will be the first experimental fusion facility, which brings together the key physical, material and technological issues related to development of fusion reactors. The design of ITER is complete and the construction will start soon. This paper discusses the main directions of the project oriented materials activity and main challenges related to selection of materials for the ITER components. For each application in ITER the main materials issues were identified and these issues were addressed in the dedicated ITER R and D program. The justification of materials performance was fully documented, which allows traceability and reliability of design data. Several examples are given to illustrate the main achievements and recommendations from the recently updated ITER Materials Properties Handbook. The main ongoing and future materials activities are described.

  15. ITER at the international conference on fusion reactor materials

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Matera, R.

    1998-01-01

    The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

  16. Eu contributions to the ITER materials properties data assessment

    Energy Technology Data Exchange (ETDEWEB)

    Peacock, A.T. [EFDA CSU, Boltzmannstrasse 2, D-85748 Garching (Germany)]. E-mail: alan.peacock@tech.efda.org; Barabash, V. [IT, ITER Joint Work Site, Boltzmannstrasse 2, D-85748 Garching (Germany)]. E-mail: barabav@itereu.de; Gillemot, F. [ASI Consulting, Budafoki ut 21, H 2040 Budaors (Hungary)]. E-mail: gillemot@sunserv.kfki.hu; Karditsas, P. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)]. E-mail: Panos.Karditsas@ukaea.org.uk; Lloyd, G. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom); Rensman, J.-W. [NRG Petten, Westerduinweg 3, P.O. Box 25, 1755 ZG Petten (Netherlands)]. E-mail: rensman@nrg-nl.com; Tavassoli, A.-A.F. [DMN/Dir, CEA/Saclay, CEA, 91191 Gif sur Yvette Cedex (France)]. E-mail: tavassoli@cea.fr; Walters, M. [EURATOM/UKAEA Fusion Association, Culham Science Centre, Abingdon (United Kingdom)

    2005-11-15

    In order to fully organise the materials property data from the European next Fusion programme, a database of materials properties has been established. With the help of the database application and resulting data organisation, European materials experts have supported the recent activities within ITER aimed at updating and re-organising the ITER materials documentation. A European web based database application is described and its main features are detailed. In addition, we report on the details and the status of the work aimed at updating the ITER materials documentation. An outline of the future planned activities in the development of the European database and in the revision of the ITER materials documentation is also given.

  17. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  18. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  19. Cryogenic Properties of Inorganic Insulation Materials for ITER Magnets: A Review

    International Nuclear Information System (INIS)

    Simon, N.J.

    1994-01-01

    Results of a literature search on the cryogenic properties of candidate inorganic insulators for the ITER TF magnets are reported. The materials investigated include: Al 2 O 3 , AlN, MgO, porcelain, SiO 2 , MgAl 2 O 4 , ZrO 2 , and mica. A graphical presentation is given of mechanical, elastic, electrical, and thermal properties between 4 and 300 K. A companion report reviews the low temperature irradiation resistance of these materials

  20. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  1. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  2. Selection, development and characterisation of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Akiba, M.; Ulrickson, M.; Vieider, G.

    1996-01-01

    The current status of the selection of the armour materials for first wall, limiters and divertor are presented. The candidate armour materials are beryllium, tungsten and carbon base materials (mainly carbon fiber composites). The selection of the references grades from these material classes is discussed and the candidate grades are described. The main reasons for the selection of the reference grades are also discussed. The urgent materials R and D needs for the development of the design are described briefly. (orig.)

  3. Erosion products of ITER divertor materials under plasma disruption simulation

    Energy Technology Data Exchange (ETDEWEB)

    Guseva, M.I.; Gureev, V.M.; Kolbasov, B.N.; Korshunov, S.N.; Martynenko, Yu.V. E-mail: martyn@nfi.kiae.ru; Stolyarova, V.G.; Strunnikov, V.M.; Vasiliev, V.I

    2003-09-01

    Candidate ITER divertor armor materials: carbon-fiber-composite and four tungsten grades/alloys as well as mixed re-deposited W+Be and W+C layers were exposed in electrodynamic plasma accelerator MKT which provided a pulsed deuterium plasma flux simulating plasma disruptions with maximum ion energy of 1-2 keV, an energy density of 300 kJ/m{sup 2} per shot and a pulse duration of {approx}60 {mu}s. The number of pulses was from 2 to 10. The resultant erosion products were collected on a basalt filter and Si-collectors and studied in terms of morphology and size distribution using both scanning and transmission electron microscopy. Metal erosion products usually occurred in the form of spherical droplets, sometimes flakes. Their size distribution depended on the positioning of the collector. Simultaneously irradiated W, CFC and mixed W+Be targets appeared to have undergone a greater erosion than the same targets irradiated individually. Particles sized from 0.01 to 30 {mu}m were found on collectors and on a molten W-surface. A model of droplet emission and behavior in shielding plasma is provided.

  4. 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)

  5. ITER solid breeder blanket materials database

    International Nuclear Information System (INIS)

    Billone, M.C.; Dienst, W.; Noda, K.; Roux, N.

    1993-11-01

    The databases for solid breeder ceramics (Li 2 ,O, Li 4 SiO 4 , Li 2 ZrO 3 and LiAlO 2 ) and beryllium multiplier material are critically reviewed and evaluated. Emphasis is placed on physical, thermal, mechanical, chemical stability/compatibility, tritium, and radiation stability properties which are needed to assess the performance of these materials in a fusion reactor environment. Correlations are selected for design analysis and compared to the database. Areas for future research and development in blanket materials technology are highlighted and prioritized

  6. Cryogenic Properties of Inorganic Insulation Materials for ITER Magnets: A Review

    Energy Technology Data Exchange (ETDEWEB)

    Simon, N.J.

    1994-12-01

    Results of a literature search on the cryogenic properties of candidate inorganic insulators for the ITER TF magnets are reported. The materials investigated include: Al{sub 2}O{sub 3}, AlN, MgO, porcelain, SiO{sub 2}, MgAl{sub 2}O{sub 4}, ZrO{sub 2}, and mica. A graphical presentation is given of mechanical, elastic, electrical, and thermal properties between 4 and 300 K. A companion report reviews the low temperature irradiation resistance of these materials.

  7. An annotated history of container candidate material selection

    International Nuclear Information System (INIS)

    McCright, R.D.

    1988-07-01

    This paper documents events in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project that have influenced the selection of metals and alloys proposed for fabrication of waste package containers for permanent disposal of high-level nuclear waste in a repository at Yucca Mountain, Nevada. The time period from 1981 to 1988 is covered in this annotated history. The history traces the candidate materials that have been considered at different stages of site characterization planning activities. At present, six candidate materials are considered and described in the 1988 Consultation Draft of the NNWSI Site Characterization Plan (SCP). The six materials are grouped into two alloy families, copper-base materials and iron to nickel-base materials with an austenitic structure. The three austenitic candidates resulted from a 1983 survey of a longer list of candidate materials; the other three candidates resulted from a special request from DOE in 1984 to evaluate copper and copper-base alloys. 24 refs., 2 tabs

  8. Significance of ITER IWS Material Selection and Qualification

    Science.gov (United States)

    Mehta, Bhoomi K.; Raval, Jigar; Maheshwari, Abha; Laad, Rahul; Singh, Gurlovleen; Pathak, Haresh

    2017-04-01

    In-Wall Shielding (IWS) is one of the important components of ITER Vacuum Vessel (VV) which fills the space between double walls of VV with cooling water. Procurement Arrangement (PA) for IWS has been signed with Indian Domestic Agency (INDA). Procurement of IWS materials, fabrication of IWS blocks and its delivery to respective Domestic Agency (DA) and ITER Organization (IO) are the main scope of this PA. Hence, INDIA is the only country which is contributing to VV IWS among all seven ITER partners. The main functions of the IWS are to provide Neutron Shielding with blanket, VV shells and water during plasma operations and to reduce ripple of the Toroidal Magnetic Field. To meet these functional requirements IWS blocks are made up of special materials (Borated Steels SS304 B4 & SS304 B7, Ferritic Steels SS 430, Austenitic Steel SS 316 L (N)-IG, XM-19 and Inconel-625) which are qualified, reliable and traceable for the design assessment. The choice of these materials has a significant influence on performance, maintainability, licensing, detailed design parameters and waste disposal. The main reasons for the materials selected for IWS are its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics and low cost relative to other similar materials. All the materials are qualified with respect to their respective codes (ASTM/EN standards with additional requirements as described in RCC-MR code 2007) and ITER requirements. Agreed Notified Body (ANB) has control conformity of materials certificates with approved material specification and traceability procedure for Safety Important Component (SIC). The procurement strategy for all the IWS materials has been developed in close collaboration with IO, ANB and Industries as per Product Procurement Specification (PPS). The R&D for sample, bulk material production, testing, inspection and handling as required are carried out by IN DA and IO. At present almost all

  9. Materials development for ITER shielding and test blanket in China

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J.M., E-mail: Chenjm@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wu, J.H.; Liu, X.; Wang, P.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China); Wang, Z.H.; Li, Z.N. [Ningxia Orient Non-ferrous Metals Group Co. Ltd., P.O. Box 105, Shizuishan (China); Wang, X.S.; Zhang, P.C. [China Academy of Engineering Physics, P.O. Box 919-71, Mianyang 621900 (China); Zhang, N.M.; Fu, H.Y.; Liu, D.H. [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041 (China)

    2011-10-01

    China is a member of the ITER program and is developing her own materials for its shielding and test blanket modules. The materials include vacuum-hot-pressing (VHP) Be, CuCrZr alloy, 316L(N) and China low activation ferritic/martensitic (CLF-1) steels. Joining technologies including Be/Cu hot isostatic pressing (HIP) and electron beam (EB) weldability of 316L(N) were investigated. Chinese VHP-Be showed good properties, with BeO content and ductility that satisfy the ITER requirements. Be/Cu mock-ups were fabricated for Be qualification tests at simulated ITER vertical displacement event (VDE) and heat flux cycling conditions. Fine microstructure and good mechanical strength of the CuCrZr alloy were achieved by a pre-forging treatment, while the weldability of 316L(N) by EB was demonstrated for welding depths varying from 5 to 80 mm. Fine microstructure, high strength, and good ductility were achieved in CLF-1 steel by an optimized normalizing, tempering and aging procedure.

  10. Armour Materials for the ITER Plasma Facing Components

    Science.gov (United States)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

    The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

  11. Armour materials for the ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

    1999-01-01

    The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

  12. Status of international collaborative efforts on selected ITER materials

    International Nuclear Information System (INIS)

    Belyakov, V.A.; Fabritsiev, S.A.; Mazul, I.V.; Rowcliffe, A.F.

    2000-01-01

    The paper presents an overview of the performance of refractory metals, beryllium, and copper alloys, for the international thermonuclear experimental reactor (ITER) high heat flux structures. High temperature brazing, hot isostatic pressing (HIP), friction welding, explosive bonding, and other methods were explored to join copper alloys to 316 stainless steel for first wall and limiter applications. It is concluded that the main material problems for the ITER high heat flux components are: (a) degradation of properties after the manufacturing cycle (especially for Be/Cu and Cu/stainless steel (SS) joints); (b) helium embrittlement of Be, and Cu, and; (c) radiation-induced loss of fracture toughness for Be, W, and Cu alloys

  13. The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

    International Nuclear Information System (INIS)

    Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

    1992-10-01

    The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

  14. Candidate materials to prevent brittle fracture - (186)

    International Nuclear Information System (INIS)

    Chanzy, Y.; Roland, V.

    2004-01-01

    For heavy transport or dual purpose casks, selecting the appropriate materials for the body is a key decision. To get a Type B(U) approval, it is necessary to demonstrate that the mechanical strength of the material is good enough at temperature as low as -40 C so as to prevent the cask from any risk of brittle fracture in regulatory accident conditions. Different methods are available to provide such a demonstration and can lead to different choices. It should be noted also that the material compositions given by national or international standards display relatively wide tolerances and therefore are not necessarily sufficient to guarantee a required toughness. It is therefore necessary to specify to the fabricator the minimum value for toughness, and to verify it. This paper gives an overview of the different methods and materials that are used in several countries. Although the safety is strongly linked to the choice of the material, it is shown that many other parameters are important, such as the design, the fabrication process (multi layer, cast or forged body), the welding material and process, the ability to detect flaws, and the measured and/or calculated stress level, including stress concentration, in particular when bolts are used. The paper will show that relying exclusively on high toughness at low temperature does not necessarily deliver the maximum safety as compared with other choices. It follows that differences in approaches to licensing by different competent authorities may bias the choice of material depending on the country of application, even though B(U) licenses are meant to guarantee unilaterally a uniform minimum level of safety

  15. Diagnostic mirrors for ITER: A material choice and the impact of erosion and deposition on their performance

    International Nuclear Information System (INIS)

    Litnovsky, A.; Wienhold, P.; Philipps, V.; Sergienko, G.; Schmitz, O.; Kirschner, A.; Kreter, A.; Droste, S.; Samm, U.; Mertens, Ph.; Donne, A.H.; Rudakov, D.; Allen, S.; Boivin, R.; McLean, A.; Stangeby, P.; West, W.; Wong, C.; Lipa, M.; Schunke, B.; De Temmerman, G.; Pitts, R.; Costley, A.; Voitsenya, V.; Vukolov, K.; Oelhafen, P.; Rubel, M.; Romanyuk, A.

    2007-01-01

    Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper

  16. Immune reactivity of candidate reference materials

    NARCIS (Netherlands)

    Fernandez-Rivas, Montserrat; Aalbers, Marja; Fötisch, Kay; de Heer, Pleuni; Notten, Silla; Vieths, Stefan; van Ree, Ronald

    2006-01-01

    Immune reactivity is a key issue in the evaluation of the quality of recombinant allergens as potential reference materials. Within the frame of the CREATE project, the immune reactivity of the natural and recombinant versions of the major allergens of birch pollen (Bet v 1), grass pollen (Phl p 1

  17. Effects of plasma disruption events on ITER first wall materials

    International Nuclear Information System (INIS)

    Cardella, A.; Gorenflo, H.; Lodato, A.; Ioki, K.; Raffray, R.

    2000-01-01

    In ITER, plasma disruption events may occur producing large fast thermal transients on plasma facing materials. Particularly important for the integrity of the first wall (FW) are relatively 'long' duration off-normal events such as plasma vertical displacement events (VDE) and runaway electrons (RE). An analytical methodology has been developed to specifically assess the effect of these events on FW plasma facing materials. For the typical energy densities and event duration expected for the primary and baffle FW, some melting and evaporation of the FW armor will occur without the beneficial effect of vapor shielding, and the metallic heat sink may also be damaged due to over-heating. The method is able to calculate the amount of melted and evaporated material, taking into account the evolution of the evaporated and melted layer and to evaluate possible effects of local temporary loss of cooling. The method has been used to analyze the effects of VDE and RE events for ITER, to study recent disruption simulation experiments and to benchmark experimental and analytical results

  18. Candidate container materials for Yucca Mountain waste package designs

    International Nuclear Information System (INIS)

    McCright, R.D.; Halsey, W.G.; Gdowski, G.E.; Clarke, W.L.

    1991-09-01

    Materials considered as candidates for fabricating nuclear waste containers are reviewed in the context of the Conceptual Design phase of a potential repository located at Yucca Mountain. A selection criteria has been written for evaluation of candidate materials for the next phase -- Advanced Conceptual Design. The selection criteria is based on the conceptual design of a thin-walled container fabricated from a single metal or alloy; the criteria consider the performance requirements on the container and the service environment in which the containers will be emplaced. A long list of candidate materials is evaluated against the criteria, and a short list of materials is proposed for advanced characterization in the next design phase

  19. 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.).

  20. 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

  1. Certification of biological candidates reference materials by neutron activation analysis

    Science.gov (United States)

    Kabanov, Denis V.; Nesterova, Yulia V.; Merkulov, Viktor G.

    2018-03-01

    The paper gives the results of interlaboratory certification of new biological candidate reference materials by neutron activation analysis recommended by the Institute of Nuclear Chemistry and Technology (Warsaw, Poland). The correctness and accuracy of the applied method was statistically estimated for the determination of trace elements in candidate reference materials. The procedure of irradiation in the reactor thermal fuel assembly without formation of fast neutrons was carried out. It excluded formation of interfering isotopes leading to false results. The concentration of more than 20 elements (e.g., Ba, Br, Ca, Co, Ce, Cr, Cs, Eu, Fe, Hf, La, Lu, Rb, Sb, Sc, Ta, Th, Tb, Yb, U, Zn) in candidate references of tobacco leaves and bottom sediment compared to certified reference materials were determined. It was shown that the average error of the applied method did not exceed 10%.

  2. ITER JCT presentation at the International Conference on Fusion Reactor Materials (ICFRM-9)

    International Nuclear Information System (INIS)

    Kalinin, G.; Barabash, V.; Ioki, K.

    1999-01-01

    During this conference four invited papers and one poster paper were presented on behalf of the ITER Joint Central Team with the review of latest achievements. The results of the comprehensive materials R and D program in support of the ITER design were extensively reported the ITER Home Teams

  3. Low activation structural material candidates for fusion power plants

    International Nuclear Information System (INIS)

    Forty, C.B.A.; Cook, I.

    1997-06-01

    Under the SEAL Programme of the European Long-Term Fusion Safety Programme, an assessment was performed of a number of possible blanket structural materials. These included the steels then under consideration in the European Blanket Programme, as well as materials being considered for investigation in the Advanced Materials Programme. Calculations were performed, using SEAFP methods, of the activation properties of the materials, and these were related, based on the SEAFP experience, to assessments of S and E performance. The materials investigated were the SEAFP low-activation martensitic steel (LA12TaLC); a Japanese low-activation martensitic steel (F-82H), a range of compositional variants about this steel; the vanadium-titanium-chromium alloy which was the original proposal of the ITER JCT for the ITER in-vessel components; a titanium-aluminium intermetallic (Ti-Al) which is under investigation in Japan; and silicon carbide composite (SiC). Assessed impurities were included in the compositions of these materials, and they have very important impacts on the activation properties. Lack of sufficiently detailed data on the composition of chromium alloys precluded their inclusion in the study. (UK)

  4. Tests of candidate materials for particle bed reactors

    International Nuclear Information System (INIS)

    Horn, F.L.; Powell, J.R.; Wales, D.

    1987-01-01

    Rhenium metal hot frits and zirconium carbide-coated fuel particles appear suitable for use in flowing hydrogen to at least 2000 K, based on previous tests. Recent tests on alternate candidate cooled particle and frit materials are described. Silicon carbide-coated particles began to react with rhenium frit material at 1600 K, forming a molten silicide at 2000 K. Silicon carbide was extensively attacked by hydrogen at 2066 K for 30 minutes, losing 3.25% of its weight. Vitrous carbon was also rapidly attacked by hydrogen at 2123 K, losing 10% of its weight in two minutes. Long term material tests on candidate materials for closed cycle helium cooled particle bed fuel elements are also described. Surface imperfections were found on the surface of pyrocarbon-coated fuel particles after ninety days exposure to flowing (∼500 ppM) impure helium at 1143 K. The imperfections were superficial and did not affect particle strength

  5. Survey of Swedish buffer material candidates and methods for characterization

    International Nuclear Information System (INIS)

    Erlstroem, M.; Pusch, R.

    1987-12-01

    The study has given a good overview of potential clay buffer candidates in the part of Sweden that offers the best possibilities to find large accessible quantities of smectitic materials. The most promising Scanian materials are those in the Kaageroed and Vallaakra (Margreteberg) areas since they represent the most smectitic ones, which may serve as raw material for the production of canister embedment. The moraine clays in the Lund-Landskrona region seem to be useful for backfilling purposes. A refined version of Reynolds technique is suggested as an SKB standard for prospecting and characterization of buffer materials. (orig./DG)

  6. Oxidation of carbon based first wall materials of ITER

    International Nuclear Information System (INIS)

    Moormann, R.R.M.; Hinssen, H.K.; Wu, C.H.

    2001-01-01

    The safety relevance of oxidation reactions on carbon materials in fusion reactors is discussed. Because tritium codeposited in ITER will probably exceed tolerable limits, countermeasures have to be developed: In this paper ozone is tested as oxidising agent for removal of codeposited layers on thick a-C:D-flakes from TEXTOR. In preceeding experiments the advantageous features of using ozonised air instead of ozonised oxygen, reported in literature for reactions with graphite, is not found for nuclear grade graphite. At 185 deg. C = 458 K ozone (0.8-3.4 vol-% in oxygen) is able to gasify the carbon content of these flakes with initial rates, comparable to initial rates in oxygen (21 kPa) for the same material at >200K higher temperatures. The layer reduction rate in ozone drops with increasing burn-off rapidly from about 0.9-2.0 μm/h to 0.20-0.25 μm/h, but in oxygen it drops to zero for all temperatures ≤ 450 deg. C = 723 K, before carbon is completely gasified. Altogether, ozone seems to be a promising oxidising agent for removal of codeposited layers, but further studies are necessary with respect to rate dependence on temperature and ozone concentration even on other kinds of codeposited layers. Further on, the optimum reaction temperature considering the limited thermal stability of ozone has to be found out and studies on the general reaction mechanism have to be done. Besides these examinations on codeposited layers, a short overview on the status of our oxidation studies on different types of fusion relevant C-based materials is given; open problems in this field are outlined. (author)

  7. 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

  8. Cryogenic Thermal Conductivity Measurements on Candidate Materials for Space Missions

    Science.gov (United States)

    Tuttle, JIm; Canavan, Ed; Jahromi, Amir

    2017-01-01

    Spacecraft and instruments on space missions are built using a wide variety of carefully-chosen materials. In addition to having mechanical properties appropriate for surviving the launch environment, these materials generally must have thermal conductivity values which meet specific requirements in their operating temperature ranges. Space missions commonly propose to include materials for which the thermal conductivity is not well known at cryogenic temperatures. We developed a test facility in 2004 at NASAs Goddard Space Flight Center to measure material thermal conductivity at temperatures between 4 and 300 Kelvin, and we have characterized many candidate materials since then. The measurement technique is not extremely complex, but proper care to details of the setup, data acquisition and data reduction is necessary for high precision and accuracy. We describe the thermal conductivity measurement process and present results for several materials.

  9. 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

  10. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Alexandrov, Evgeny [Institution “Project Center ITER”, 1, Akademika Kurchatova sq., 123182 Moscow (Russian Federation); Batyunin, Alexander; Kashchuk, Yuri [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Korban, Svetlana; Lyublin, Boris [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Obudovsky, Sergey [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Senik, Konstantin [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity.

  11. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Alexandrov, Evgeny; Batyunin, Alexander; Kashchuk, Yuri; Korban, Svetlana; Lyublin, Boris; Obudovsky, Sergey; Senik, Konstantin

    2013-01-01

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity

  12. Development of advanced tritium breeding material with added lithium for ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Hoshino, Tsuyoshi, E-mail: hoshino.tsuyoshi@jaea.go.jp [Blanket Irradiation and Analysis Group, Fusion Research and Development Directorate, Japan Atomic Energy Agency, 4002, Narita-cho, Oarai-machi, Higashi Ibaraki-gun, Ibaraki 311-1393 (Japan); Kato, Kenichi; Natori, Yuri; Oikawa, Fumiaki; Nakano, Natsuko; Nakamura, Mutsumi [Kaken, Co. Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan); Sasaki, Kazuya [Institute of Engineering Innovation and Department of Nuclear Engineering and Management School of Engineering, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-8656 (Japan); Suzuki, Akihiro [Nuclear Professional School, School of Engineering, University of Tokyo, 2-22 Shirakata-Shirane, Ibaraki 319-1188 (Japan); Terai, Takayuki [Institute of Engineering Innovation and Department of Nuclear Engineering and Management School of Engineering, University of Tokyo, 2-11-16 Yayoi, Bunkyo-ku, Tokyo 113-8656 (Japan); Tatenuma, Katsuyoshi [Kaken, Co. Ltd., 1044, Hori, Mito-city, Ibaraki 310-0903 (Japan)

    2011-10-01

    Lithium titanate (Li{sub 2}TiO{sub 3}) is one of the most promising candidates among tritium breeding materials because of its good tritium release characteristics. However, the mass of Li{sub 2}TiO{sub 3} decreased with time in a hydrogen atmosphere by the reduction of Ti and Li evaporation. In order to prevent the mass decrease at high temperatures, advanced tritium breeding material with added Li (Li{sub 2+x}TiO{sub 3+y}) should be developed. For this purpose, an advanced Li{sub 2}TiO{sub 3} with added Li was synthesized from proportionally mixed LiOH.H{sub 2}O and H{sub 2}TiO{sub 3} with a Li/Ti ratio of 2.2. The results of X-ray diffraction measurement showed that this advanced tritium breeding material existed as the non-stoichiometric compound Li{sub 2+x}TiO{sub 3+y}. The desired molar ratio of Li/Ti was achieved by appropriate mixing of LiOH.H{sub 2}O and H{sub 2}TiO{sub 3}. Therefore, synthesis by mixing LiOH.H{sub 2}O and H{sub 2}TiO{sub 3} is a promising mass production method for the advanced tritium breeding material with added Li for the test blanket module of ITER.

  13. Space Environmental Effects on Candidate Solar Sail Materials

    Science.gov (United States)

    Edwards, David L.; Nehls, Mary; Semmel, Charles; Hovater, Mary; Gray, Perry; Hubbs, Whitney; Wertz, George

    2004-01-01

    The National Aeronautics and Space Administration's (NASA) Marshall Space Flight Center (MSFC) continues research into the utilization of photonic materials for spacecraft propulsion. Spacecraft propulsion, using photonic materials, will be achieved using a solar sail. A solar sail operates on the principle that photons, originating from the sun, impart pressure to the sail and therefore provide a source for spacecraft propulsion. The pressure imparted ot a solar sail can be increased, up to a factor of two, if the sun-facing surface is perfectly reflective. Therefore, these solar sails are generally composed of a highly reflective metallic sun-facing layer, a thin polymeric substrate and occasionally a highly emissive back surface. Near term solar sail propelled science missions are targeting the Lagrange point 1 (L1) as well as locations sunward of L1 as destinations. These near term missions include the Solar Polar Imager and the L1 Diamond. The Environmental Effects Group at NASA's Marshall Space Flight Center (MSFC) continues to actively characterize solar sail material in preparation for these near term solar sail missions. Previous investigations indicated that space environmental effects on sail material thermo-optical properties were minimal and would not significantly affect the propulsion efficiency of the sail. These investigations also indicated that the sail material mechanical stability degrades with increasing radiation exposure. This paper will further quantify the effect of space environmental exposure on the mechanical properties of candidate sail materials. Candidate sail materials for these missions include Aluminum coated Mylar, Teonex, and CP1 (Colorless Polyimide). These materials were subjected to uniform radiation doses of electrons and protons in individual exposures sequences. Dose values ranged from 100 Mrads to over 5 Grads. The engineering performance property responses of thermo-optical and mechanical properties were characterized

  14. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    Allen, T.R.; Was, G.S.

    2008-01-01

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept. The objective of the proposed research was to investigate degradation of materials in the supercritical water environment (SCW). First, representative alloys from the important classes of candidate materials were studied for their corrosion and stress-corrosion cracking (SCC) resistance in supercritical water. These included ferritic/martensitic (F/M) steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests were conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies were applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and SCC of alloys in the as-received and modified/engineered conditions were examined by irradiating samples using high-energy protons and then exposing them to SCW

  15. Behavior of candidate canister materials in deep ocean environments

    International Nuclear Information System (INIS)

    Smyrl, W.H.; Stephenson, L.L.; Braithwaite, J.W.

    1977-04-01

    Corrosion tests have been conducted under simulated deep ocean conditions for nine months. The materials tested were base alloys of titanium, zirconium, and nickel. All materials tested showed corrosion rates that were very low even at the highest test temperature. None showed susceptibility to either stress corrosion cracking or differential aeration corrosion. Ambient electrochemical tests confirmed the findings that none should be sensitive to differential oxygen effects. The zirconium alloys may be more susceptible to pitting corrosion than the others, although the pitting conditions are unlikely to be found in service, unless higher temperatures are encountered. All the alloys tested could give long life under deep ocean conditions and are candidates for more detailed corrosion studies

  16. ITER transient consequences for material damage: modelling versus experiments

    International Nuclear Information System (INIS)

    Bazylev, B; Janeschitz, G; Landman, I; Pestchanyi, S; Loarte, A; Federici, G; Merola, M; Linke, J; Zhitlukhin, A; Podkovyrov, V; Klimov, N; Safronov, V

    2007-01-01

    Carbon-fibre composite (CFC) and tungsten macrobrush armours are foreseen as PFC for the ITER divertor. In ITER the main mechanisms of metallic armour damage remain surface melting and melt motion erosion. In the case of CFC armour, due to rather different heat conductivities of CFC fibres a noticeable erosion of the PAN bundles may occur at rather small heat loads. Experiments carried out in the plasma gun facilities QSPA-T for the ITER like edge localized mode (ELM) heat load also demonstrated significant erosion of the frontal and lateral brush edges. Numerical simulations of the CFC and tungsten (W) macrobrush target damage accounting for the heat loads at the face and lateral brush edges were carried out for QSPA-T conditions using the three-dimensional (3D) code PHEMOBRID. The modelling results of CFC damage are in a good qualitative and quantitative agreement with the experiments. Estimation of the droplet splashing caused by the Kelvin-Helmholtz (KH) instability was performed

  17. 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)

  18. Phosphate bonded ceramics as candidate final-waste-form materials

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Cunnane, J.; Sutaria, M.; Kurokawa, S.; Mayberry, J.

    1994-04-01

    Room-temperature setting phosphate-bonded ceramics were studied as candidate materials for stabilization of DOE low-level problem mixed wastes which cannot be treated by other established stabilization techniques. Phosphates of Mg, Mg-Na, Al and Zr were studied to stabilize ash surrogate waste containing RCRA metals as nitrates and RCRA organics. We show that for a typical loading of 35 wt.% of the ash waste, the phosphate ceramics pass the TCLP test. The waste forms have high compression strength exceeding ASTM recommendations for final waste forms. Detailed X-ray diffraction studies and differential thermal analyses of the waste forms show evidence of chemical reaction of the waste with phosphoric acid and the host matrix. The SEM studies show evidence of physical bonding. The excellent performance in the leaching tests is attributed to a chemical solidification and physical as well as chemical bonding of ash wastes in these phosphate ceramics

  19. Compatibility of candidate structural materials with static gallium

    International Nuclear Information System (INIS)

    Luebbers, P.R.; Michaud, W.F.; Chopra, O.K.

    1993-01-01

    Scoping tests were conducted on compatibility of gallium with candidate structural materials, e.g., Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, as well as Armco iron, Nickel 270, and pure chronimum. Type 316 stainless steel is least resistant and Nb-5 Mo-1 Zr alloy is most resistant to corrosion in static gallium. At 400 degrees C, corrosion rates are ∼4.0, 0.5, and 0.03 mm/y for Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, respectively. The pure metals react rapidly with gallium. In contrast to findings in earlier studies, pure iron shows greater corrosion than does nickel. The corrosion rates at 400 degrees C are ≥90 and 17 mm/y, respectively, for Armco iron and Nickel 270. The results indicate that at temperatures up to 400 degrees C, corrosion occurs primarily by dissolution accompanied by formation of metal/gallium intermetallic compounds

  20. Surface segregation in binary alloy first wall candidate materials

    International Nuclear Information System (INIS)

    Gruen, D.M.; Krauss, A.R.; Mendelsohn, M.H.; Susman, S.; Argonne National Lab., IL

    1982-01-01

    We have been studying the conditions necessary to produce a self-sustaining stable lithium monolayer on a metal substrate as a means of creating a low-Z film which sputters primarily as secondary ions. It is expected that because of the toroidal field, secondary ions originating at the first wall will be returned and contribute little to the plasma impurity influx. Aluminum and copper have, because of their high thermal conductivity and low induced radioactivity, been proposed as first wall candidate materials. The mechanical properties of the pure metals are very poorly suited to structural applications and an alloy must be used to obtain adequate hardness and tensile strength. In the case of aluminum, mechanical properties suitable for aircraft manufacture are obtained by the addition of a few at% Li. In order to investigate alloys of a similar nature as candidate structural materials for fusion machines we have prepared samples of Li-doped aluminum using both a pyro-metallurgical and a vapor-diffusion technique. The sputtering properties and surface composition have been studied as a function of sample temperature and heating time, and ion beam mass. The erosion rate and secondary ion yield of both the sputtered Al and Li have been monitored by secondary ion mass spectroscopy and Auger analysis providing information on surface segregation, depth composition profiles, and diffusion rates. The surface composition ahd lithium depth profiles are compared with previously obtained computational results based on a regular solution model of segregation, while the partial sputtering yields of Al and Li are compared with results obtained with a modified version of the TRIM computer program. (orig.)

  1. Corrosion of candidate container materials by Yucca Mountain bacteria

    International Nuclear Information System (INIS)

    Horn, J; Jones, D; Lian, T; Martin, S; Rivera, A

    1999-01-01

    Several candidate container materials have been studied in modified Yucca Mountain (YM) ground water in the presence or absence of YM bacteria. YM bacteria increased corrosion rates by 5-6 fold in UNS G10200 carbon steel, and nearly 100-fold in UNS NO4400 Ni-Cu alloy. YM bacteria caused microbiologically influenced corrosion (MIC) through de-alloying or Ni-depletion of Ni-Cu alloy as evidenced by scanning electronic microscopy (SEM) and inductively coupled plasma spectroscopy (ICP) analysis. MIC rates of more corrosion-resistant alloys such as UNS NO6022 Ni-Cr- MO-W alloy, UN's NO6625 Ni-Cr-Mo alloy, and UNS S30400 stainless steel were measured below 0.05 umyr, however YM bacteria affected depletion of Cr and Fe relative to Ni in these materials. The chemical change on the metal surface caused by depletion was characterized in anodic polarization behavior. The anodic polarization behavior of depleted Ni-based alloys was similar to that of pure Ni. Key words: MIC, container materials, YM bacteria, de-alloying, Ni-depletion, Cr-depletion, polarization resistance, anodic polarization,

  2. Effects of ELMs on ITER divertor armour materials

    Energy Technology Data Exchange (ETDEWEB)

    Zhitlukhin, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)]. E-mail: zhitlukh@triniti.ru; Klimov, N. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Landman, I. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany)]. E-mail: j.linke@fz-juelich.de; Loarte, A. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Merola, M. [EFDA, Boltzmannstr. 2, 85748 Garching (Germany); Podkovyrov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Federici, G. [ITER JWS Garching, Boltzmannstr. 2, 85748 Garching (Germany); Bazylev, B. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Pestchanyi, S. [Forschungszentrum Karlsruhe, P.O. Box 3640, 76021 Karlsruhe (Germany); Safronov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Hirai, T. [Forschungszentrum Juelich, EURATOM-Association, Juelich (Germany); Maynashev, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Levashov, V. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation); Muzichenko, A. [SRC RF TRINITI, Troitsk, 142190, Moscow Region (Russian Federation)

    2007-06-15

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m{sup 2} and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 {mu}m per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m{sup 2} and was increased to the average value 0.3 {mu}m per pulse at 1.5 MJ/m{sup 2}. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m{sup 2}. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m{sup 2}.

  3. Effects of ELMs on ITER divertor armour materials

    Science.gov (United States)

    Zhitlukhin, A.; Klimov, N.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Federici, G.; Bazylev, B.; Pestchanyi, S.; Safronov, V.; Hirai, T.; Maynashev, V.; Levashov, V.; Muzichenko, A.

    2007-06-01

    This paper is concerned with investigation of an erosion of the ITER-like divertor plasma facing components under plasma heat loads expected during the Type I ELMs in ITER. These experiments were carried out on plasma accelerator QSPA at the SRC RF TRINITI under EU/RF collaboration. Targets were exposed by series repeated plasma pulses with heat loads in a range of 0.5-1.5 MJ/m2 and pulse duration 0.5 ms. Erosion of CFC macrobrushes was determined mainly by sublimation of PAN-fibres that was less than 2.5 μm per pulse. The CFC erosion was negligible at the energy density less than 0.5 MJ/m2 and was increased to the average value 0.3 μm per pulse at 1.5 MJ/m2. The pure tungsten macrobrushes erosion was small in the energy range of 0.5-1.3 MJ/m2. The sharp growth of tungsten erosion and the intense droplet ejection were observed at the energy density of 1.5 MJ/m2.

  4. Review of candidate welding processes of RAFM steels for ITER test blanket modules and DEMO

    Energy Technology Data Exchange (ETDEWEB)

    Aubert, P., E-mail: philippe.aubert@cea.fr [CEA, DEN, DM2S, F-91191 Gif-sur-Yvette (France); Tavassoli, F. [CEA, DEN, DMN, F-91191 Gif-sur-Yvette (France); Rieth, M. [FZK, Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Diegele, E.; Poitevin, Y. [Fusion for Energy (F4E), C/Josep Pla 2 - Ed. B3, 08019 Barcelona (Spain)

    2011-10-01

    EUROFER weldability is investigated in support of the European TBM manufacturing. Electron beam, hybrid, laser and NGTIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel. It is shown that the most promising processes are laser, electron beam and hybrid welding, depending on the section size and accessibility. They produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The FZ are typically composed of martensite laths, with small grain sizes. In the HAZ, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. A one step PWHT (750 deg. C/3 h) is successfully applied to joints restoring good material performance. Distortion levels, with and without PWHT, are controlled through adaptation of manufacturing steps and clamping devices, obtaining levels not exceeding 120 {mu}m (+/-60 {mu}m) on a full 'one cell mock-up'.

  5. Stress corrosion cracking of candidate materials for nuclear waste containers

    International Nuclear Information System (INIS)

    Maiya, P.S.; Shack, W.J.; Kassner, T.F.

    1989-09-01

    Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

  6. Effects of ELMs and disruptions on ITER divertor armour materials

    International Nuclear Information System (INIS)

    Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

    2005-01-01

    This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ∼1.5 MJ/m 2 , consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ∼1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized

  7. Effects of ELMs and disruptions on ITER divertor armour materials

    Science.gov (United States)

    Federici, G.; Zhitlukhin, A.; Arkhipov, N.; Giniyatulin, R.; Klimov, N.; Landman, I.; Podkovyrov, V.; Safronov, V.; Loarte, A.; Merola, M.

    2005-03-01

    This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ˜1.5 MJ/m 2, consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ˜1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

  8. Laboratory Reference Spectroscopy of Icy Satellite Candidate Surface Materials (Invited)

    Science.gov (United States)

    Dalton, J. B.; Jamieson, C. S.; Shirley, J. H.; Pitman, K. M.; Kariya, M.; Crandall, P.

    2013-12-01

    The bulk of our knowledge of icy satellite composition continues to be derived from ultraviolet, visible and infrared remote sensing observations. Interpretation of remote sensing observations relies on availability of laboratory reference spectra of candidate surface materials. These are compared directly to observations, or incorporated into models to generate synthetic spectra representing mixtures of the candidate materials. Spectral measurements for the study of icy satellites must be taken under appropriate conditions (cf. Dalton, 2010; also http://mos.seti.org/icyworldspectra.html for a database of compounds) of temperature (typically 50 to 150 K), pressure (from 10-9 to 10-3 Torr), viewing geometry, (i.e., reflectance), and optical depth (must manifest near infrared bands but avoid saturation in the mid-infrared fundamentals). The Planetary Ice Characterization Laboratory (PICL) is being developed at JPL to provide robust reference spectra for icy satellite surface materials. These include sulfate hydrates, hydrated and hydroxylated minerals, and both organic and inorganic volatile ices. Spectral measurements are performed using an Analytical Spectral Devices FR3 portable grating spectrometer from .35 to 2.5 microns, and a Thermo-Nicolet 6500 Fourier-Transform InfraRed (FTIR) spectrometer from 1.25 to 20 microns. These are interfaced with the Basic Extraterrestrial Environment Simulation Testbed (BEEST), a vacuum chamber capable of pressures below 10-9 Torr with a closed loop liquid helium cryostat with custom heating element capable of temperatures from 30-800 Kelvins. To generate optical constants (real and imaginary index of refraction) for use in nonlinear mixing models (i.e., Hapke, 1981 and Shkuratov, 1999), samples are ground and sieved to six different size fractions or deposited at varying rates to provide a range of grain sizes for optical constants calculations based on subtractive Kramers-Kronig combined with Hapke forward modeling (Dalton and

  9. Status of experimental data related to Be in ITER materials R and D data bank

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, Shigeru [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    To keep traceability of many valuable raw data that were experimentally obtained in the ITER Technology R and D Tasks related to materials for In-Vessel components (divertor, first wall, blanket, vacuum vessel, etc.) and to easily make the best use of these data in the ITER design activities, the `ITER Materials R and D Data Bank` has been built up, with the use of Excel{sup TM} spread sheets. The paper describes status of experimental data collected in this data bank on thermo-mechanical properties of unirradiated and neutron irradiated Be, on plasma-material interactions of Be, on mechanical properties of various kinds of Be/Cu joints (including plasma sprayed Be), and on thermal fatigue tests of Be/Cu mock-ups. (author)

  10. 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.

  11. 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.

  12. Neutron activation behavior of NET/ITER divertor structural materials

    International Nuclear Information System (INIS)

    Smid, I.; Weimann, G.; Kny, E.; Kneringer, G.; Reheis, N.

    1995-01-01

    The post-activation behavior of the materials carbon, TZM (99.3 % Mo) and Mo.41Re, as well as of high temperature brazes suitable for their joining after irradiation with 14 MeV neutrons has been evaluated. The activity, dose rate and energy generation after exposure to an ignited fusion plasma is presented for various time steps after shutdown. The impact of the activity and the afterheat production on the handling and storage conditions of retired divertor components is simulated, the required protection for maintenance is discussed. Further the temperature of stored divertor elements after a full time operation in NET was calculated. No major afterheat production will occur and thus no special cooling is to be provided after approximately one month. Taking into account convection and radiation the equilibrium temperature of vertically stored environment/aircooled divertor elements is predicted to be approximately 100 degree C. (author)

  13. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

    Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

  14. 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

  15. Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

    Science.gov (United States)

    H, L. SWAMI; C, DANANI; A, K. SHAW

    2018-06-01

    Activation analyses play a vital role in nuclear reactor design. Activation analyses, along with nuclear analyses, provide important information for nuclear safety and maintenance strategies. Activation analyses also help in the selection of materials for a nuclear reactor, by providing the radioactivity and dose rate levels after irradiation. This information is important to help define maintenance activity for different parts of the reactor, and to plan decommissioning and radioactive waste disposal strategies. The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential, due to the presence of a high-energy neutron environment which makes decisive demands on material selection. This study comprises two parts; in the first part the activation characteristics, in a fusion radiation environment, of several elements which are widely present in structural materials, are studied. It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment. The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions. The structural materials selected for this study, i.e. India-specific Reduced Activation Ferritic‑Martensitic steel (IN-RAFMS), P91-grade steel, stainless steel 316LN ITER-grade (SS-316LN-IG), stainless steel 316L and stainless steel 304, are candidates for use in ITER either in vessel components or test blanket systems. Tungsten is also included in this study because of its use for ITER plasma-facing components. The study is carried out using the reference parameters of the ITER fusion reactor. The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port. The presence of elements like Nb, Mo, Co and Ta in a structural material enhance the activity level as well

  16. 77 FR 20886 - Proposed Information Collection (Advertising, Sales, and Enrollment Materials, and Candidate...

    Science.gov (United States)

    2012-04-06

    ... (Advertising, Sales, and Enrollment Materials, and Candidate Handbooks) Activity: Comment Request AGENCY... the Office of Management and Budget (OMB) for each collection of information they conduct or sponsor... information technology. Title: Advertising, Sales, and Enrollment Materials, and Candidate Handbooks, 38 CFR...

  17. Assessment of database for interaction of tritium with ITER plasma facing materials

    International Nuclear Information System (INIS)

    Dolan, T.J.; Anderl, R.A.

    1994-09-01

    The present work surveys recent literature on hydrogen isotope interactions with Be, SS and Inconels, Cu, C, and V, and alloys of Cu and V. The goals are (1) to provide input to the International Thermonuclear Experimental Reactor (ITER) team to help with tritium source term estimates for the Early Safety and Environmental Characterization Study and (2) to provide guidance for planning additional research that will be needed to fill gaps in the present materials database. Properties of diffusivity, solubility, permeability, chemical reactions, Soret effect, recombination coefficient, surface effects, trapping, porosity, layered structures, interfaces, and oxides are considered. Various materials data are tabulated, and a matrix display shows an assessment of the quality of the data available for each main property of each material. Recommendations are made for interim values of diffusivity and solubility to be used, pending further discussion by the ITER community

  18. 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.).

  19. European TBM for ITER: Structural material assessment and breeding capability - Comparative analysis

    International Nuclear Information System (INIS)

    Herreras, Y.; Perlado, J.M.; Ibarra, A.

    2007-01-01

    Full text of publication follows: The ITER European Party is currently developing for DEMO reactor specifications two breeding blanket concepts: the Helium-Cooled Lithium-Lead blanket (HCLL), using a liquid breeder; and the Helium-Cooled Pebble-Bed blanket (HCPB), using a lithiated solid breeder. These two research lines are expected to be tested in ITER as Test Blanket Modules (TBM), in order to demonstrate their safety, economical and environmental suitability. In this sense, structural material activation and breeding blanket capability represent two major challenges. This paper presents new calculations regarding neutronic irradiation inside the ITER Vacuum Vessel. In particular, results are focused on the irradiation affecting the equatorial ports, where the TBM will be located for testing. The methodology employed mainly consists in calculating the neutronic irradiation levels at the required locations with the transport code MCNP, where the input geometry has been previously designed with the program CATIA V5. The main structural materials proposed for the European Test blanket Modules are selected in order to carry out a comparative analysis in safety terms: material activation and basic parameters for damage analysis are evaluated with the code ACAB, based on the neutronic irradiation results mentioned above. Finally, the breeding blanket capability is assessed for both breeding blanket concepts; the results are compared considering the choice of the structural material. (authors)

  20. Research status and issues of tungsten plasma facing materials for ITER and beyond

    International Nuclear Information System (INIS)

    Ueda, Y.; Coenen, J.W.; De Temmerman, G.; Doerner, R.P.; Linke, J.; Philipps, V.; Tsitrone, E.

    2014-01-01

    This review summarizes surface morphology changes of tungsten caused by heat and particle loadings from edge plasmas, and their effects on enhanced erosion and material lifetime in ITER and beyond. Pulsed heat loadings by transients (disruption and ELM) are the largest concerns due to surface melting, cracking, and dust formation. Hydrogen induced blistering is unlikely to be an issue of ITER. Helium bombardment would cause surface morphology changes such as W fuzz, He holes, and nanometric bubble layers, which could lead to enhanced erosion (e.g. unipolar arcing of W fuzz). Particle loadings could enhance pulsed heat effects (cracking and erosion) due to surface layer embrittlement by nanometric bubbles and solute atoms. But pulsed heat loadings alleviate surfaces morphology changes in some cases (He holes by ELM-like heat pulses). Effects of extremely high fluence (∼10 30 m −2 ), mixed materials, and neutron irradiation are important issues to be pursued for ITER and beyond. In addition, surface refurbishment to prolong material lifetime is also an important issue

  1. Status of development of functional materials with perspective on beyond ITER

    International Nuclear Information System (INIS)

    Shikama, T.; Knitter, R.; Moeslang, A.; Konys, J.; Deli, L.; Muroga, T.; Kawamura, H.; Kohyama, A.

    2007-01-01

    Any engineering system is composed of functional materials as well as of structural materials, and more advanced systems tend to demand a more important and versatile role to functional materials. In nuclear fusion systems, examples of principle functional materials will be breeders and neutron multipliers for tritium production, coatings on structural materials for corrosion-resistance, MHD-loss-reduction and control of tritium permeation, thermal insertions for heat transport control, and optical and electrical materials for plasma and environmental diagnostics. For incarnation of a nuclear fusion power plant, namely DEMO, development of the functional materials with appropriate properties is essential. A role of functional materials depends strongly on a specific design of DEMO, namely designs of systems for tritium-breeding, system-cooling and heat-transfer. In the framework of ITER project, development of tritium blanket modules (TBM) is underway. Also, in parallel with the ITER project, a complemental program called the BA (Broader Approach) is launched for realization of a DEMO nuclear fusion reactor in an appropriate time schedule, where key issues of the nuclear fusion engineering needed for the DEMO will be studied under EU/Japan collaboration. In the meantime, technologies and materials needed for diagnostics and control of burning plasma are extensively discussed under the framework of International Tokamak Physics Activity (ITPA). The present paper reviews a present status of development of functional materials from views of internationally coordinated activities based on fundamental aspects of the DEMO demands as well as from views of activities based on specific but currently dominant DEMO designs. Examples of functional materials reviewed here are solid breeders, beryllium and beryllium alloys, coating layers on structural materials, thermal inserts, and some electrical and optical materials. (orig.)

  2. Dust Erosion Performance of Candidate Motorcase Thermal Protection Materials.

    Science.gov (United States)

    1980-03-10

    REFERENCE DESCRIPTION SOURCE NUMBER 4.01 NBR B. F. Goodrich Aerospace and Defense Products (Nitrile butadiene 500 South Main Street rubber ) Akron, Ohio...material degradation occurs. 5.3 BALLISTIC RANGES Ballistic ranges are widely used for reentry erosion testing for two reasons: 1) no other type of facility...DET REFERENCE OTHER COMMENTS NUMBER DESIGNATION 2002 KEVLAR-EPOXY STAGE 3 MOTORCASE MATERIAL MOTORCAS E 2402 NBR 68 2403 NBR 69 2404 NBR -19709-6A (60

  3. 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.

  4. Corrosion of candidate materials in Lake Rotokawa geothermal exposure

    Energy Technology Data Exchange (ETDEWEB)

    Estill, J.C.; McCright, R.D.

    1995-05-01

    Corrosion rates were determined for CDA 613, CDA 715, A-36 carbon steel, 1020 carbon steel, and Alloy 825 flat coupons which were exposed to geothermal spring water at Paraiki site number 9 near Lake Rotokawa, New Zealand. Qualitative observations of the corrosion performance of Type 304L stainless steel and CDA 102 exposed to the same environment were noted. CDA 715, Alloy 825, 1020 carbon steel, and other alloys are being considered for the materials of construction for high-level radioactive waste containers for the United States civilian radioactive waste disposal program. Alloys CDA 613 and CDA 102 were tested to provide copper-based materials for corrosion performance comparison purposes. A36 was tested to provide a carbon steel baseline material for comparison purposes, and alloy 304L stainless steel was tested to provide an austenitic stainless steel baseline material for comparison purposes. In an effort to gather corrosion data from an environment that is rooted in natural sources of water and rock, samples of some of the proposed container materials were exposed to a geothermal spring environment. At the proposed site at Yucca Mountain, Nevada, currently under consideration for high-level nuclear waste disposal, transient groundwater may come in contact with waste containers over the course of a 10,000-year disposal period. The geothermal springs environment, while extremely more aggressive than the anticipated general environment at Yucca Mountain, Nevada, could have similarities to the environment that arises at selected local sites on a container as a result of crevice corrosion, pitting corrosion, microbiologically influenced corrosion (MIC), or the concentration of the ionic species due to repetitive evaporation or boiling of the groundwater near the containers. The corrosion rates were based on weight loss data obtained after six weeks exposure in a 90{degrees}C, low-pH spring with relatively high concentrations of SO{sub 4}{sup 2-} and Cl{sup -}.

  5. Remote-LIBS characterization of ITER-like plasma facing materials

    International Nuclear Information System (INIS)

    Almaviva, S.; Caneve, L.; Colao, F.; Fantoni, R.; Maddaluno, G.

    2012-01-01

    Graphical abstract: Display Omitted Highlights: ► Description of a LIBS set-up as remote diagnostics in new generation fusion machines. ► Identification of the atomic composition of samples simulating plasma facing components. ► Submicrometric resolution in depth profiling the elemental composition of the samples. ► Identification of elements present in traces or as impurities on the sample surface. ► Discussion on the applicability of the Calibration Free method for quantitative analysis. - Abstract: The occurrence of several plasma-wall interaction processes, eventually affecting the overall system performances, is expected in a working fusion device chamber. Monitoring the changes in the composition of the plasma facing component (PFC) surface layer, as a result of erosion and redeposition mechanisms, can provide useful information on the possible plasma pollution and fuel retention. To this aim, suitable diagnostic techniques able to perform depth profiling analysis of the superficial layers on the PFCs have been developed. Due to the constraints commonly found in fusion devices, the measuring apparatus must be non invasive, remote and sensitive to light elements. These requirements make LIBS (Laser Induced Breakdown Spectroscopy) an ideal candidate for on-line monitoring the walls of current and of next generation (as ITER) fusion devices. LIBS is a well established tool for qualitative, semi-quantitative and quantitative analysis of surfaces, with micro-destructive characteristics and some capabilities for stratigraphy. In this work, LIBS depth profiling capability has been verified for the determination of the composition of multilayer structures simulating plasma facing components covered with deposited impurity layers. A new experimental setup has been designed and realized in order to optimize the characteristics of a LIBS system working in vacuum conditions and remotely, two noticeable properties for an ITER-relevant diagnostics. A quantitative

  6. Sound absorption of low-temperature reusable surface insulation candidate materials

    Science.gov (United States)

    Johnston, J. D.

    1974-01-01

    Sound absorption data from tests of four candidate low-temperature reusable surface insulation materials are presented. Limitations on the use of the data are discussed, conclusions concerning the effective absorption of the materials are drawn, and the relative significance to Vibration and Acoustic Test Facility test planning of the absorption of each material is assessed.

  7. Irradiation creep experiments on fusion reactor candidate structural materials

    International Nuclear Information System (INIS)

    Hausen, H.; Cundy, M.R.; Schuele, W.

    1991-01-01

    Irradiation creep rates were determined for annealed and cold-worked AMCR- and 316-type steel alloys in the high flux reactor at Petten, for various irradiation temperatures, stresses and for neutron doses up to 4 dpa. Primary creep elongations were found in all annealed materials. A negative creep elongation was found in cold-worked materials for stresses equal to or below about 100 MPa. An increase of the negative creep elongation is found for decreasing irradiation temperatures and decreasing applied stresses. The stress exponent of the irradiation creep rate in annealed and cold-worked AMCR alloys is n = 1.85 and n = 1.1, respectively. The creep rates of cold-worked AMCR alloys are almost temperature independent over the range investigated (573-693 K). The results obtained in the HFR at Petten are compared with those obtained in ORR and EBR II. The smallest creep rates are found for cold-worked materials of AMCR- and US-PCA-type at Petten which are about a factor two smaller than the creep rates obtained of US-316 at Petten or for US-PCA at ORR or for 316L at EBR II. The scatter band factor for US-PCA, 316L, US-316 irradiated in ORR and EBR II is about 1.5 after a temperature and damage rate normalization

  8. VUV photoemission studies of candidate LHC vacuum chamber materials

    CERN Document Server

    Baglin, V; Collins, I R

    1998-01-01

    In the context of future accelerators and, in particular, the beam vacuum of the LargeHadron Collider (LHC), a 27 km circumference proton collider to be built at CERN, VUVsynchrotron radiation (SR) has been used to study both qualitatively and quantitatively candidatevacuum chamber materials. Emphasis is given to show that angle and energy resolvedphotoemission is an extremely powerful tool to address important issues relevant to the LHC, suchas the emission of electrons that contribute to the creation of an electron cloud which may causeserious beam instabilities. Here we present not only the measured photoelectron yields (PY)from the proposed materials, prepared on an industrial scale, but also the energy and, in some cases,the angular dependence of the emitted electrons when excited with either a white light (WL)spectrum, simulating that in the arcs of the LHC or monochromatic light in the photon energy rangeof interest. The effects on the materials examined of WL irradiation and/or ion sputtering,simulati...

  9. Corrosion susceptibility study of candidate pin materials for ALTC (Active Lithium/Thionyl Chloride) batteries

    Science.gov (United States)

    Bovard, Francine S.; Cieslak, Wendy R.

    1987-09-01

    The corrosion susceptibilities of eight alternate battery pin material candidates for ALTC (Active Lithium/Thionyl Chloride) batteries in 1.5M LiAlCl4/SOCl2 electrolyte have been investigated using ampule exposure and electrochemical tests. The thermal expansion coefficients of these candidate materials are expected to match Sandia-developed Li-corrosion resistant glasses. The corrosion resistances of the candidate materials, which included three stainless steels (15-5 PH, 17-4 PH, and 446), three Fe-Ni glass sealing alloys (Kovar, Alloy 52, and Niromet 426), a Ni-based alloy (Hastelloy B-2) and a zirconium-based alloy (Zircaloy), were compared to the reference materials Ni and 316L SS. All of the candidate materials showed some evidence of corrosion and, therefore, did not perform as well as the reference materials. The Hastelloy B-2 and Zircaloy are clearly unacceptable materials for this application. Of the remaining alternate materials, the 446 SS and Alloy 52 are the most promising candidates.

  10. Simulated Space Environment Effects on a Candidate Solar Sail Material

    Science.gov (United States)

    Kang, Jin Ho; Bryant, Robert G.; Wilkie, W. Keats; Wadsworth, Heather M.; Craven, Paul D.; Nehls, Mary K.; Vaughn, Jason A.

    2017-01-01

    For long duration missions of solar sails, the sail material needs to survive harsh space environments and the degradation of the sail material controls operational lifetime. Therefore, understanding the effects of the space environment on the sail membrane is essential for mission success. In this study, we investigated the effect of simulated space environment effects of ionizing radiation, thermal aging and simulated potential damage on mechanical, thermal and optical properties of a commercial off the shelf (COTS) polyester solar sail membrane to assess the degradation mechanisms on a feasible solar sail. The solar sail membrane was exposed to high energy electrons (about 70 keV and 10 nA/cm2), and the physical properties were characterized. After about 8.3 Grad dose, the tensile modulus, tensile strength and failure strain of the sail membrane decreased by about 20 95%. The aluminum reflective layer was damaged and partially delaminated but it did not show any significant change in solar absorbance or thermal emittance. The effect on mechanical properties of a pre-cracked sample, simulating potential impact damage of the sail membrane, as well as thermal aging effects on metallized PEN (polyethylene naphthalate) film will be discussed.

  11. Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W

    International Nuclear Information System (INIS)

    Skinner, C.H.; Haasz, A.A.; Alimov, V.Kh.; Bekris, N.; Causey, R.A.; Clark, R.E.H.; Coad, J.P.; Davis, J.W.; Doerner, R.P.; Mayer, M.; Pisarev, A.; Roth, J.; Tanabe, T.

    2008-01-01

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described

  12. Recent Advances on Hydrogenic Retention in ITER's Plasma-Facing Materials: BE, C, W.

    Energy Technology Data Exchange (ETDEWEB)

    Skinner, C H; Alimov, Kh; Bekris, N; Causey, R A; Clark, R.E.H.; Coad, J P; Davis, J W; Doerner, R P; Mayer, M; Pisarev, A; Roth, J

    2008-03-29

    Management of tritium inventory remains one of the grand challenges in the development of fusion energy and the choice of plasma-facing materials is a key factor for in-vessel tritium retention. The Atomic and Molecular Data Unit of the International Atomic Energy Agency organized a Coordinated Research Project (CRP) on the overall topic of tritium inventory in fusion reactors during the period 2001-2006. This dealt with hydrogenic retention in ITER's plasma-facing materials, Be, C, W, and in compounds (mixed materials) of these elements as well as tritium removal techniques. The results of the CRP are summarized in this article together with recommendations for ITER. Basic parameters of diffusivity, solubility and trapping in Be, C and W are reviewed. For Be, the development of open porosity can account for transient hydrogenic pumping but long term retention will be dominated by codeposition. Codeposition is also the dominant retention mechanism for carbon and remains a serious concern for both Be and C containing layers. Hydrogenic trapping in unirradiated tungsten is low but will increase with ion and neutron damage. Mixed materials will be formed in a tokamak and these can also retain significant amounts of hydrogen isotopes. Oxidative and photon-based techniques for detritiation of plasma-facing components are described.

  13. 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.

  14. Graphene oxide as an optimal candidate material for methane storage.

    Science.gov (United States)

    Chouhan, Rajiv K; Ulman, Kanchan; Narasimhan, Shobhana

    2015-07-28

    Methane, the primary constituent of natural gas, binds too weakly to nanostructured carbons to meet the targets set for on-board vehicular storage to be viable. We show, using density functional theory calculations, that replacing graphene by graphene oxide increases the adsorption energy of methane by 50%. This enhancement is sufficient to achieve the optimal binding strength. In order to gain insight into the sources of this increased binding, that could also be used to formulate design principles for novel storage materials, we consider a sequence of model systems that progressively take us from graphene to graphene oxide. A careful analysis of the various contributions to the weak binding between the methane molecule and the graphene oxide shows that the enhancement has important contributions from London dispersion interactions as well as electrostatic interactions such as Debye interactions, aided by geometric curvature induced primarily by the presence of epoxy groups.

  15. Irradiation creep of candidate materials for advanced nuclear plants

    Energy Technology Data Exchange (ETDEWEB)

    Chen, J., E-mail: jiachao.chen@psi.ch; Jung, P.; Hoffelner, W.

    2013-10-15

    In the present paper, irradiation creep results of an intermetallic TiAl alloy and two ferritic oxide dispersion strengthened (ODS) steels are summarized. In situ irradiation creep measurements were performed using homogeneous implantation with α- and p-particles to maximum doses of 0.8 dpa at displacement damage rates of 2–8 × 10{sup −6} dpa/s. The strains of miniaturized flat dog-bone specimens were monitored under uniaxial tensile stresses ranging from 20 to 400 MPa at temperatures of 573, 673 and 773 K, respectively. The effects of material composition, ODS particle size, and bombarding particle on the irradiation creep compliance was studied and results are compared to literature data. Evolution of microstructure during helium implantation was investigated in detail by TEM and is discussed with respect to irradiation creep models.

  16. Breeding blanket design for ITER and prototype (DEMO) fusion reactors and breeding materials issues

    Energy Technology Data Exchange (ETDEWEB)

    Takatsu, H; Enoeda, M [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    1998-03-01

    Current status of the designs of the ITER breeding blanket and DEMO blankets is introduced placing emphasis on the breeding materials selection and related issues. The former design is based on the up-to-date design activities, as of October 1997, being performed jointly by Joint Central Team (JCT) and Home Teams (HT`s), while the latter is based on the DEMO blanket test module designs being proposed by each Party at the TBWG (Test Blanket Working Group) meetings. (J.P.N.)

  17. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Science.gov (United States)

    Klimov, N. S.; Putrik, A. B.; Linke, J.; Pitts, R. A.; Zhitlukhin, A. M.; Kuprianov, I. B.; Spitsyn, A. V.; Ogorodnikova, O. V.; Podkovyrov, V. L.; Muzichenko, A. D.; Ivanov, B. V.; Sergeecheva, Ya. V.; Lesina, I. G.; Kovalenko, D. V.; Barsuk, V. A.; Danilina, N. A.; Bazylev, B. N.; Giniyatulin, R. N.

    2015-08-01

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic-martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, "corrugated" surface, with hills and valleys spaced by 0.2-2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

  18. Plasma facing materials performance under ITER-relevant mitigated disruption photonic heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Klimov, N.S., E-mail: klimov@triniti.ru [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Putrik, A.B. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Linke, J. [Forschungszentrum Jülich GmbH, EURATOM Association, Jülich D-52425 (Germany); Pitts, R.A. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Zhitlukhin, A.M. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Kuprianov, I.B. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Spitsyn, A.V. [NRC «Kurchatov Institute», Akademika Kurchatova pl., 1, Moscow 123182 (Russian Federation); Ogorodnikova, O.V. [Max-Planck-Institut für Plasmaphysik, Garching (Germany); National Research Nuclear University MEPhI (Moscow Engineering Physics Institute), Kashirskoye shosse 31, Moscow 115409 (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Ivanov, B.V.; Sergeecheva, Ya.V.; Lesina, I.G. [Bochvar Institute, ul. Rogova, 5a, Moscow 123098 (Russian Federation); Kovalenko, D.V.; Barsuk, V.A.; Danilina, N.A. [SRC RF TRINITI, ul. Pushkovykh, vladenie 12, Troitsk, Moscow 142190 (Russian Federation); Bazylev, B.N. [Karlsruhe Institute of Technology, P.O. Box 3640, Karlsruhe 76021 (Germany); Giniyatulin, R.N. [Efremov Institute, Doroga na Metallostroy, 3 bld., Metallostroy, Saint-Petersburg 196641 (Russian Federation)

    2015-08-15

    PFMs (Plasma-facing materials: ITER grade stainless steel, beryllium, and ferritic–martensitic steels) as well as deposited erosion products of PFCs (Be-like, tungsten, and carbon based) were tested in QSPA under photonic heat loads relevant to those expected from photon radiation during disruptions mitigated by massive gas injection in ITER. Repeated pulses slightly above the melting threshold on the bulk materials eventually lead to a regular, “corrugated” surface, with hills and valleys spaced by 0.2–2 mm. The results indicate that hill growth (growth rate of ∼1 μm per pulse) and sample thinning in the valleys is a result of melt-layer redistribution. The measurements on the 316L(N)-IG indicate that the amount of tritium absorbed by the sample from the gas phase significantly increases with pulse number as well as the modified layer thickness. Repeated pulses significantly below the melting threshold on the deposited erosion products lead to a decrease of hydrogen isotopes trapped during the deposition of the eroded material.

  19. Materials effects and design implications of disruptions and off-normal events in ITER

    International Nuclear Information System (INIS)

    Hassanein, A.; Federici, G.; Konkashbaev, I.; Zhitlukhin, A.; Litunovsky, V.

    1997-01-01

    Damage to plasma-facing components (PFCs) and structural materials during abnormal plasma behavior such as hard disruptions, edge-localized modes (ELMs), and vertical displacement events (VDEs) is considered a serious life-limiting concern for these components. The PFCs in the International Thermonuclear Experimental Reactor (ITER), such as the divertor, limiter, and parts of the first wall, will be subjected to high energy deposition during these plasma instabilities. High erosion losses on material surfaces, high temperature rise in structural materials (particularly at the bonding interface), and high heat flux levels and possible burnout of the coolant tubes are critical constraints that severely limit component lifetime and therefore degrade reactor performance, safety, and economics. Recently developed computer models and simulation experiments are being used to evaluate various damage to PFCs during the abnormal events. The design implications of plasma-facing and nearby components are discussed, and recommendations are made to mitigate the effects of these events

  20. Element concentrations in candidate biological and environmental reference materials by k0-standardized INAA

    International Nuclear Information System (INIS)

    Freitas, M.C.

    1993-01-01

    K 0 -Based Neutron Activation Analysis (k 0 INAA) was used to analyze the candidate reference materials Apple Leaves and Peach Leaves, and Oriental Tobacco Leaves and Virginia Tobacco Leaves. Concentration values for 27 elements were measured. The accuracy was ascertained by analysis of two certified reference materials. NIST 1572 Citrus Leaves and 1573 Tomato Leaves. The homogeneity test of the IAEA Evernia prunastri candidate reference material in aliquots ≥ 100 mg is extended to the elements Sc, Cr, Fe, Co, Zn, Rb, Sb, Cs, Ba, Ce and Th. (orig.)

  1. Reliability of Scores Obtained from Self-, Peer-, and Teacher-Assessments on Teaching Materials Prepared by Teacher Candidates

    Science.gov (United States)

    Nalbantoglu Yilmaz, Funda

    2017-01-01

    This study aims to determine the reliability of scores obtained from self-, peer-, and teacher-assessments in terms of teaching materials prepared by teacher candidates. The study group of this research constitutes 56 teacher candidates. In the scope of research, teacher candidates were asked to develop teaching material related to their study.…

  2. Material migration patterns and overview of first surface analysis of the JET ITER-like wall

    International Nuclear Information System (INIS)

    Widdowson, A; Ayres, C F; Baron-Wiechec, A; Matthews, G F; Alves, E; Catarino, N; Brezinsek, S; Coad, J P; Likonen, J; Heinola, K; Mayer, M; Rubel, M

    2014-01-01

    Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition. (paper)

  3. Preliminary cleaning tests on candidate materials for APS beamline and front end UHV components

    International Nuclear Information System (INIS)

    Nielsen, R.; Kuzay, T.M.

    1992-01-01

    Comparative cleaning tests have been done on four candidate materials for use in APS beamline and front-end vacuum components. These materials are 304 SS, 304L SS, OFHC copper, and Glidcop* (Cu-Al 2 O 3 )- Samples of each material were prepared and cleaned using two different methods. After cleaning, the sample surfaces were analyzed using ESCA (Electron Spectography for Chemical Analysis). Uncleaned samples were used as a reference. The cleaning methods and surface analysis results are further discussed

  4. Helium embrittlement model and program plan for weldability of ITER materials

    International Nuclear Information System (INIS)

    Louthan, M.R. Jr.; Kanne, W.R. Jr.; Tosten, M.H.; Rankin, D.T.; Cross, B.J.

    1997-02-01

    This report presents a refined model of how helium embrittles irradiated stainless steel during welding. The model was developed based on experimental observations drawn from experience at the Savannah River Site and from an extensive literature search. The model shows how helium content, stress, and temperature interact to produce embrittlement. The model takes into account defect structure, time, and gradients in stress, temperature and composition. The report also proposes an experimental program based on the refined helium embrittlement model. A parametric study of the effect of initial defect density on the resulting helium bubble distribution and weldability of tritium aged material is proposed to demonstrate the roll that defects play in embrittlement. This study should include samples charged using vastly different aging times to obtain equivalent helium contents. Additionally, studies to establish the minimal sample thickness and size are needed for extrapolation to real structural materials. The results of these studies should provide a technical basis for the use of tritium aged materials to predict the weldability of irradiated structures. Use of tritium charged and aged material would provide a cost effective approach to developing weld repair techniques for ITER components

  5. Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

    International Nuclear Information System (INIS)

    Pan, Z.L.

    2010-01-01

    Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

  6. Composite Materials and Meta Materials for a New Approach to ITER ICRH Loads

    International Nuclear Information System (INIS)

    Bottollier-Curtet, H.; Argouarch, A.; Vulliez, K.; Becoulet, A.; Litaudon, X.; Magne, R.; Champeaux, S.; Gouard, Ph.; Primout, M.; Le Gallou, J.-H.

    2009-01-01

    Preliminary laboratory testing of ICRH antennas is a very useful step before their commissioning. Traditionally, pure water, salt water or baking soda water loads are used. These 'water' loads are convenient but strongly limited in terms of performance testing. We have started two feasibility studies for advanced ICRH loads made of ferroelectric ceramics (passive loads) and meta materials (active loads). Preliminary results are very encouraging.

  7. Composite Materials and Meta Materials for a New Approach to ITER ICRH Loads

    Science.gov (United States)

    Bottollier-Curtet, H.; Argouarch, A.; Champeaux, S.; Gouard, Ph.; Le Gallou, J.-H.; Primout, M.; Vulliez, K.; Bécoulet, A.; Litaudon, X.; Magne, R.

    2009-11-01

    Preliminary laboratory testing of ICRH antennas is a very useful step before their commissioning. Traditionally, pure water, salt water or baking soda water loads are used. These "water" loads are convenient but strongly limited in terms of performance testing. We have started two feasibility studies for advanced ICRH loads made of ferroelectric ceramics (passive loads) and meta materials (active loads) [1]. Preliminary results are very encouraging.

  8. Mechanical Properties of High Manganese Austenitic Stainless Steel JK2LB for ITER Central Solenoid Jacket Material

    Science.gov (United States)

    Saito, Toru; Kawano, Katsumi; Yamazaki, Toru; Ozeki, Hidemasa; Isono, Takaaki; Hamada, Kazuya; Devred, Arnaud; Vostner, Alexander

    A suite of advanced austenitic stainless steels are used for the ITER TF, CS and PF coil systems.These materials will be exposed to cyclic-stress at cryogenic temperature. Therefore, high manganese austenitic stainless steel JK2LB, which has high tensile strength, high ductility and high resistance to fatigue at 4 K has been chosen for the CS conductor. The cryogenic temperature mechanical property data of this material are very important for the ITER magnet design. This study is focused on mechanical characteristics of JK2LB and its weld joint.

  9. The opinions of primary school teachers’ candidates towards material preparation and usage

    Directory of Open Access Journals (Sweden)

    Zeynep Genc

    2017-04-01

    Full Text Available Abstract Instruction materials help students to acquire more memorable information. Instruction materials have an important effect on providing more permanent and simple way of learning in every step of education. Instruction materials are the most frequently used by primary school teachers. Primary school teachers should support their lectures with instruction materials in order to provide permanent learning. The Teaching Technologies and Material Designing (TTMD course which is one of the compulsory courses that students must take aims to acquire students the information and skills related with the preparation and use of materials. Evaluation of TTMD course is important in terms of the effectiveness of the course which provides the opportunity of motivating the students to learn by attracting their attention, keeping their attentions alive, making abstract concepts more concrete, facilitating the acquisition of knowledge in an organized way in the process of learning and teaching. In this context, it was aimed to determine the opinions of students in the department of primary school teaching about preparation and use of materials through teaching practice which is done within TTMD course in this study. This study is a descriptive study based on qualitative data. The sample of this research included 37 students from the department of primary school teaching who took TTMD course in the second semester in 2014-2015 academic year at Ataturk Education Faculty of Near East University or students who took this course in previous academic years. The data of this research were collected with structured interview form. According to the results, it was revealed that primary school teachers’ candidates attach importance to prepare and use materials based on their answers about the use and preparation of materials in instruction. When the opinions of primary school teachers candidates about the criteria that they give value in preparing and using

  10. Material activation assessment for waste analysis of the EU design of RC/RTO ITER

    International Nuclear Information System (INIS)

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

    2001-01-01

    This paper presents the results of Sn radiation transport and activation calculations related to the ITER RC/RTO EU-I design, performed in support of safety and waste management analyses. The activation characteristics (included the clearance levels) have been estimated for the different materials/zones of the equatorial plane up to 10 5 years after plasma operations. The Bonami-XSDNRPM sequence of the Scale 4.4 code system (using Vitamin-ENEA library, based on ENDF/B-VI data) has been used for radiation transport analyses. The ANITA-4M activation code (with FENDL/A-2 and FENDL/D-2 activation and decay data libraries) is used for the activation calculation. The unconditional clearance level data library is based on IAEA-TECDOC-855. First, a sensitivity analysis to optimise the radial spatial meshing for the neutron flux distribution evaluation and, accordingly, for the activation calculation, has been performed. Then, the clearance indexes of vessel and ex-vessel zones/materials have been calculated. The results are presented and discussed. A design option that considers copper instead of superconductor material for TFC winding pack has also been considered and assessed

  11. Investigation of cryogenic irradiation influence on mechanical and physical properties of ITER magnetic system insulation materials

    International Nuclear Information System (INIS)

    Kozlov, A.V.; Scherbacov, E.N.; Dudchenko, N.A.; Shihalev, V.S.; Bedin, V.V.; Paltusov, N.A.; Korsunskiy, V.E.

    1998-01-01

    A set of methods of cryogenic irradiation influence test on mechanical and physical properties of insulation of ITER magnetic system are presented in this paper. Investigations are carried out without intermediate warming up of samples. A Russian insulating composite material was irradiated in the IVV-2M reactor. The ratio of energy absorbed by insulation materials from neutron irradiation to that from gamma irradiation can be varied from ∝(25:75)% to ∝(50:50)% in the reactor. The test results on the thermal expansion, thermal conductivity and gas evolution of the above material are presented. It was shown, that cryogenic irradiation up to the fluence ∝2 x 10 22 n/m 2 (E ≥ 0.1 MeV) leads to 0.27% linear size changes along layers of fiber-glass, the thermal conductivity coefficient is decreased on 15% at 100 k in perpendicular direction to fiber-glass plane, and thermal coefficient of linear expansion (TCLE) has anomalous temperature dependence. (orig.)

  12. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  13. Thermal properties variations in unconsolidated material for very shallow geothermal application (ITER project)

    Science.gov (United States)

    Sipio, Eloisa Di; Bertermann, David

    2018-04-01

    In engineering, agricultural and meteorological project design, sediment thermal properties are highly important parameters, and thermal conductivity plays a fundamental role when dimensioning ground heat exchangers, especially in very shallow geothermal systems. Herein, the first 2 m of depth from surface is of critical importance. However, the heat transfer determination in unconsolidated material is difficult to estimate, as it depends on several factors, including particle size, bulk density, water content, mineralogy composition and ground temperature. The performance of a very shallow geothermal system, as a horizontal collector or heat basket, is strongly correlated to the type of sediment at disposal and rapidly decreases in the case of dry-unsaturated conditions. The available experimental data are often scattered, incomplete and do not fully support thermo-active ground structure modeling. The ITER project, funded by the European Union, contributes to a better knowledge of the relationship between thermal conductivity and water content, required for understanding the very shallow geothermal systems behaviour in saturated and unsaturated conditions. So as to enhance the performance of horizontal geothermal heat exchangers, thermally enhanced backfilling material were tested in the laboratory, and an overview of physical-thermal properties variations under several moisture and load conditions for different mixtures of natural material was here presented.

  14. Characteristics study of bentonite as candidate of buffer materials for radioactive waste disposal system

    International Nuclear Information System (INIS)

    Suryantoro; Arimuladi, S.P.; Sastrowardoyo, P.B.

    1998-01-01

    Literature studies on bentonite characteristic of, as candidate for radioactive waste disposal system, have been conducted. Several information have been obtained from references, which would be contributed on performance assessment of engineered barrier. The functions bentonite includes the buffering of chemical and physical behavior, i.e. swelling property, self sealing, hydraulic conductivities and gas permeability. This paper also presented long-term stability of bentonite in natural condition related to the illitisazation, which could change its buffering capacities. These information, showed that bentonite was satisfied to be used for candidate of buffer materials in radioactive waste disposal system. (author)

  15. Characterisation of candidate reference materials by PIXE analysis and nuclear microprobe PIXE imaging

    International Nuclear Information System (INIS)

    Jaksic, M.; Pastuovic, Z.; Bogdanovic, I.; Tadic, T.

    2002-01-01

    In order to test whether some candidate reference materials show homogeneity that can satisfy quality control of the PIXE technique, six bottles of each of the two Candidate RM's - Lichen (IAEA 338) and Algae (IAEA 413) were tested. Four different tests were performed. First, two pellets from each bottle were prepared and analysed using broad beam (φ = 5 mm) PIXE. Second and third was analysis of homogeneity using scanning focussed beam at the nuclear microprobe. Scans of 50x50 μm 2 and 240x260 μm 2 were performed. Finally, individual grains with composition differing from the rest of the sample, were analysed using PIXE and RBS. (author)

  16. River bottom sediment from the Vistula as matrix of candidate for a new reference material.

    Science.gov (United States)

    Kiełbasa, Anna; Buszewski, Bogusław

    2017-08-01

    Bottom sediments are very important in aquatic ecosystems. The sediments accumulate heavy metals and compounds belonging to the group of persistent organic pollutants. The accelerated solvent extraction (ASE) was used for extraction of 16 compounds from PAH group from bottom sediment of Vistula. For the matrix of candidate of a new reference material, moisture content, particle size, loss on ignition, pH, and total organic carbon were determined. A gas chromatograph with a selective mass detector (GC/MS) was used for the final analysis. The obtained recoveries were from 86% (SD=6.9) for anthracene to 119% (SD=5.4) for dibenzo(ah)anthracene. For the candidate for a new reference material, homogeneity and analytes content were determined using a validated method. The results are a very important part of the development and certification of a new reference materials. Copyright © 2017 Elsevier Inc. All rights reserved.

  17. Material Surface Damage under High Pulse Loads Typical for ELM Bursts and Disruptions in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Landman, I.S.; Pestchanyi, S.E.; Bazylev, B.N [Forschungszentrum Karlsruhe (Germany). Inst. for Pulsed Power and Microwave Technology; Safronov, V.M. [Troitsk Inst. for Innovation and Fusion Research (TRINITI) (Russian Federation); Garkusha, I.E. [Kharkov Inst. of Physics and Technology (KIPT) (Ukraine). Inst. of Plasma Physics

    2004-08-01

    The divertor armour material for the tokamak ITER will probably be carbon manufactured as fibre composites (CFC) and tungsten as either brush-like structures or thin plates. Disruptive pulse loads where the heat deposition Q may reach 10{sup 2} MJ/m{sup 2} on a time scale {tau} of 3 ms, or operation in the ELMy H-mode at repetitive loads with Q {approx} 3MJ/m{sup 2} and {tau}{approx}0.3 ms; deteriorate armour performance. This work surveys recent numerical and experimental investigations of erosion mechanisms at these off-normal regimes carried out at FZK, TRINITI, and IPP-Kharkov. The modelling uses the anisotropic thermomechanics code PEGASUS-3D for the simulation of CFC brittle destruction, the surface melt motion code MEMOS-1.5D for tungsten targets, and the radiation-magnetohydrodynamics code FOREV-2D for calculating the plasma impact and simulating the heat loads for the ITER regime. Experiments aimed at validating these codes are being carried out at the plasma gun facilities MK-200UG, QSPA-T, and QSPA-Kh50 which produce powerful streams of hydrogen plasma with Q=10-30MJ/m{sup 2} and {tau} = 0.03-0.5 ms. Essential results are, for CFC targets, the experiments at high heat loads and the development of a local overheating model incorporated in PEGASUS-3D, and for the tungsten targets the analysis of evaporation- and melt motion erosion on the base of MEMOS-1.5D calculations for repetitive ELMs.

  18. Material Surface Damage under High Pulse Loads Typical for ELM Bursts and Disruptions in ITER

    Science.gov (United States)

    Landman, I. S.; Pestchanyi, S. E.; Safronov, V. M.; Bazylev, B. N.; Garkusha, I. E.

    The divertor armour material for the tokamak ITER will probably be carbon manufactured as fibre composites (CFC) and tungsten as either brush-like structures or thin plates. Disruptive pulse loads where the heat deposition Q may reach 102 MJ/m 2 on a time scale Ïä of 3 ms, or operation in the ELMy H-mode at repetitive loads with Q âe 1/4 3 MJ/m2 and Ïä âe 1/4 0.3 ms, deteriorate armour performance. This work surveys recent numerical and experimental investigations of erosion mechanisms at these off-normal regimes carried out at FZK, TRINITI, and IPP-Kharkov. The modelling uses the anisotropic thermomechanics code PEGASUS-3D for the simulation of CFC brittle destruction, the surface melt motion code MEMOS-1.5D for tungsten targets, and the radiation-magnetohydrodynamics code FOREV-2D for calculating the plasma impact and simulating the heat loads for the ITER regime. Experiments aimed at validating these codes are being carried out at the plasma gun facilities MK-200UG, QSPA-T, and QSPA-Kh50 which produce powerful streams of hydrogen plasma with Q = 10–30 MJ/m2 and Ïä = 0.03–0.5 ms. Essential results are, for CFC targets, the experiments at high heat loads and the development of a local overheating model incorporated in PEGASUS-3D, and for the tungsten targets the analysis of evaporation- and melt motion erosion on the base of MEMOS-1.5D calculations for repetitive ELMs.

  19. Element content and particle size characterization of a mussel candidate reference material

    International Nuclear Information System (INIS)

    Moreira, Edson G.; Vasconcellos, Marina B.A.; Santos, Rafaela G. dos; Martinelli, Jose R.

    2011-01-01

    The use of certified reference materials is an important tool in the quality assurance of analytical measurements. To assure reliability on recently prepared powder reference materials, not only the characterization of the property values of interest and their corresponding uncertainties, but also physical properties such as the particle size distribution must be well evaluated. Narrow particle size distributions are preferable than larger ones; as different size particles may have different analyte content. Due to this fact, the segregation of the coarse and the fine particles in a bottle may lead to inhomogeneity of the reference material, which should be avoided. In this study the element content as well as the particle size distribution of a mussel candidate reference material produced at IPEN-CNEN/SP was investigated. Instrumental Neutron Activation Analysis was applied to the determination of 15 elements in seven fractions of the material with different particle size distributions. Subsamples of the materials were irradiated simultaneously with elemental standards at the IEA-R1 research nuclear reactor and the induced gamma ray energies were measured in a hyperpure germanium detector. Three vials of the candidate reference material and three coarser fractions, collected during the preparation, were analyzed by Laser Diffraction Particle Analysis to determine the particle size distribution. Differences on element content were detected for fractions with different particle size distribution, indicating the importance of particle size control for biological reference materials. From the particle size analysis, Gaussian particle size distribution was observed for the candidate reference material with mean particle size μ = 94.6 ± 0.8 μm. (author)

  20. Basic study on weldability and machinability of structural materials for ITER toroidal field coils

    International Nuclear Information System (INIS)

    Onozuka, M.; Shimizu, K.; Urata, K.; Kimura, M.; Kadowaki, H.; Okamoto, M.; Nakajima, H.; Hamada, K.; Okuno, K.

    2006-01-01

    The toroidal field (TF) coils for ITER are very large components. The main structural component of the coil is the coil case, which requires a massive complex geometry with high fabrication accuracy to attain the required magnetic performance for plasma operations. To provide high mechanical strength and toughness at cryogenic temperature, the structural components employ high-strength austenite stainless steels that have been specially developed for ITER. However, one of the main drawbacks of using those materials is the difficulty of manufacturing capabilities. A manufacturing study has been conducted to examine welding and machining capabilities for JJ1 and ST-SS316LN, to be employed for TF coil structural components. Both materials include a high nitrogen content up to around 0.2%, which makes welding and machining difficult compared with conventional stainless steels. Electron beam welding conditions were studied for the JJ1 material. The applicable welding condition was found for a bead length of up to about 300 mm in the case of 40 mm thick plates. No optimal condition was found for plates thicker than 40 mm. An additional experimental study was also conducted to explore suitable welding conditions for different welding positions and directions. It was found that the appearance of defects depends on the welding positions and directions. A wider range of welding conditions was found for cases in the vertical upward direction, as opposed to those in the vertical downward and horizontal directions. Based on those results, a verification test up to 900 mm in length was conducted. The test results showed that vertical upward EB welding should be used for the coil case wherever possible. With respect to TIG welding, an average deposition rate as high as 26 g/min (i.e. the filler wire supplying speed of 3,000 mm/min) was achieved. A series of tests have been conducted to examine machinability of JJ1 and ST-SS316LN. Various types of milling tools, including face

  1. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Farmer, J.C.; McCright, R.D.; Kass, J.N.

    1988-06-01

    Three iron- to nickel-based austenitic alloys and three copper-based alloys are being considered as candidate materials for the fabrication of high-level radioactive-waste disposal containers. The austenitic alloys are Types 304L and 316L stainless steels and the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). Waste in the forms of both spent fuel assemblies from reactors and borosilicate glass will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking; and transgranular stress corrosion cracking. Problems specific to welds, such as hot cracking, may also occur. A survey of the literature has been prepared as part of the process of selecting, from among the candidates, a material that is adequate for repository conditions. The modes of degradation are discussed in detail in the survey to determine which apply to the candidate alloys and the extent to which they may actually occur. The eight volumes of the survey are summarized in Sections 1 through 8 of this overview. The conclusions drawn from the survey are also given in this overview

  2. The function of packing materials in a high-level nuclear waste repository and some candidate materials: Salt Repository Project

    International Nuclear Information System (INIS)

    Bunnell, L.R.; Shade, J.W.

    1987-03-01

    Packing materials should be included in waste package design for a high-level nuclear waste repository in salt. A packing material barrier would increase confidence in the waste package by alleviating possible shortcomings in the present design and prolonging confinement capabilities. Packing materials have been studied for uses in other geologic repositories; appropriately chosen, they would enhance the confinement capabilities of salt repository waste packages in several ways. Benefits of packing materials include retarding or chemically modifying brines to reduce corrosion of the waste package, providing good thermal conductivity between the waste package and host rock, retarding or absorbing radionuclides, and reducing the massiveness of the waste package. These benefits are available at low percentage of total repository cost, if the packing material is properly chosen and used. Several candidate materials are being considered, including oxides, hydroxides, silicates, cement-based mixtures, and clay mixtures. 18 refs

  3. Mechanical and thermal resistance of multi-material components for ITER

    International Nuclear Information System (INIS)

    Burlet, H.

    2013-01-01

    The First Wall panels for ITER are complex parts composed of stainless steel, copper and beryllium [1]. These materials are joined using diffusion bonding technique. The stainless steel is a commonly used in nuclear reactors 316LN material and acts as a structural material. The copper alloy is a CuCrZr material which acts as a heat sink. The beryllium consisting in tiles and layer is used as the protective plasma facing material. The fabrication of these panels is performed through 2 main steps. The first step consists in welding all together a bi-metallic support structure made from a thick CuCrZr plate embedded with 316LN tubes and bonded to a thick 316LN backing plate with cooling channels. The bonding is performed in a HIP (Hot Isostatic Pressure) facility. The second step is performed at a lower temperature and aims at simultaneously welding by HIP Be onto CuCrZr and ageing the CuCrZr heat sink to obtain the correct mechanical resistance of this alloy reinforced by precipitates. The various joints 316LN/316LN, 316LN/CuCrZr, and CuCrZr/Be are then characterized [2] from a microstructural point of view and by mechanical tests. It is quite hard to characterize the strength of a diffusion bonded joints. Standard tests may be used for homogeneous joints whereas specific tests have been developed to characterize the heterogeneous bonds. To optimize the bond, we performed mainly impact and tensile bi-material tests (Fig 1). Once the manufacture parameters have been optimized, advanced mechanical tests are performed based on Bimetallic CT specimens, axisymmetric notched specimens, 4P bending specimens. Numerical simulations are required to analyse the mechanical response. In order to characterize the fatigue resistance of the joints, specific mock-ups have been designed by the European Fusion Development Agreement EFDA team (Fig 2). Results of heat flux testing will be reviewed for the various joints. The assembly of heterogeneous materials by Hipping is very complex

  4. Application of a passive electrochemical noise technique to localized corrosion of candidate radioactive waste container materials

    International Nuclear Information System (INIS)

    Korzan, M.A.

    1994-05-01

    One of the key engineered barriers in the design of the proposed Yucca Mountain repository is the waste canister that encapsulates the spent fuel elements. Current candidate metals for the canisters to be emplaced at Yucca Mountain include cast iron, carbon steel, Incoloy 825 and titanium code-12. This project was designed to evaluate passive electrochemical noise techniques for measuring pitting and corrosion characteristics of candidate materials under prototypical repository conditions. Experimental techniques were also developed and optimized for measurements in a radiation environment. These techniques provide a new method for understanding material response to environmental effects (i.e., gamma radiation, temperature, solution chemistry) through the measurement of electrochemical noise generated during the corrosion of the metal surface. In addition, because of the passive nature of the measurement the technique could offer a means of in-situ monitoring of barrier performance

  5. Stability of aflatoxin B1 in animal feed candidate reference materials

    NARCIS (Netherlands)

    Roos, A.H.; Mazijk, van R.J.; Tuinstra, L.G.M.T.; Huf, F.A.

    1991-01-01

    Two candidate reference materials animal feed were stored at a temperature of -18°C, 4 C, 20°C and 37°C. The stability of aflatoxin B1 was studied duringa period of two years. A significant decrease in the aflatoxin B1 content was measured in the samples stared at 20°C and 37°C. In the samples

  6. Static and Dynamic Friction Behavior of Candidate High Temperature Airframe Seal Materials

    Science.gov (United States)

    Dellacorte, C.; Lukaszewicz, V.; Morris, D. E.; Steinetz, B. M.

    1994-01-01

    The following report describes a series of research tests to evaluate candidate high temperature materials for static to moderately dynamic hypersonic airframe seals. Pin-on-disk reciprocating sliding tests were conducted from 25 to 843 C in air and hydrogen containing inert atmospheres. Friction, both dynamic and static, was monitored and serves as the primary test measurement. In general, soft coatings lead to excessive static friction and temperature affected friction in air environments only.

  7. Data for the sorption of actinides on candidate materials for use in repositories

    International Nuclear Information System (INIS)

    Morgan, R.D.; Pryke, D.C.; Rees, J.H.

    1988-02-01

    The sorptive behaviour of the actinides uranium, neptunium, plutonium and americium has been investigated under air-saturated conditions on a number of candidate near-field materials by batch sorption experiments. Distribution ratios were measured with respect to initial actinide concentration, the solid:liquid ratio and contact time. Desorption experiments were carried out to help elucidate the mechanism of sorption. The fit of the data to the Freundlich isotherm was assessed. This work contains the data obtained in the investigation. (author)

  8. Characterization of Candidate Solar Sail Material Exposed to Space Environmental Effects

    Science.gov (United States)

    Edwards, David; Hovater, Mary; Hubbs, Whitney; Wertz, George; Hollerman, William; Gray, Perry

    2003-01-01

    Solar sailing is a unique form of propulsion where a spacecraft gains momentum from incident photons. Solar sails are not limited by reaction mass and provide continual acceleration, reduced only by the lifetime of the lightweight film in the space environment and the distance to the Sun. Once thought to be difficult or impossible, solar sailing has come out of science fiction and into the realm of possibility. Any spacecraft using this method would need to deploy a thin sail that could be as large as many kilometers in extent. The availability of strong, ultra lightweight, and radiation resistant materials will determine the future of solar sailing. The National Aeronautics and Space Administration's Marshall Space Flight Center (MSFC) is concentrating research into the utilization of ultra lightweight materials for spacecraft propulsion. The Space Environmental Effects Team at MSFC is actively characterizing candidate solar sail material to evaluate the thermo-optical and mechanical properties after exposure to space environmental effects. This paper will describe the exposure of candidate solar sail materials to emulated space environmental effects including energetic electrons, combined electrons and Ultraviolet radiation, and hypervelocity impact of irradiated solar sail material. This paper will describe the testing procedure and the material characterization results of this investigation.

  9. Fracture toughness properties of candidate canister materials for spent fuel storage by concrete cask

    International Nuclear Information System (INIS)

    Arai, Taku; Mayuzumi, Masami; Libin, Niu; Takaku, Hiroshi

    2005-01-01

    It is very significant to clarify the fracture toughness properties of candidate canister materials to ensure the structural integrity against the accidents during handling in the storage facility. Fracture toughness tests on the CT specimens cut from base metal, heat affected zone (HAZ) and weld metal in the 2 types of weld joints made by candidate canister materials (SUS329J4L duplex stainless steel and YUS270 super stainless steel) were conducted under various test temperature between 233K and 473K. Stable ductile crack extensions were observed in all of the specimens. The fracture toughness J Q of the base metal and the HAZ of SUS329L4L showed the smallest value at 233K, and increased with temperature, then reached to the largest value at 298K. At the higher temperature, the value of J Q decreased slightly with temperature. While, the value of J Q in the weld metal increased with temperature. The value of J Q of YUS270 increased with temperature. The values of J Q for weld metal in both of the materials were not greater than those in base metal and HAZ at each test temperature. The values of J Q in weld metal of both materials at 213K and 473K were greater than applied J derived from postulated semi-elliptical surface flaw and maximum allowable stress in JSME design coed. This result suggested that these materials have enough toughness for use as the canister material. (author)

  10. Corrosion of candidate materials for canister: applications in rock salt formations

    International Nuclear Information System (INIS)

    Azkarate, I.; Madina, V.; Barrio, A. del; Macarro, J.M.

    1994-01-01

    Previous corrosion studies carried out on various metallic materials in typical salt rock environments show that carbon steel and titanium alloys are the most promising candidates for canister applications in this geological formation. Although carbon steels have a low corrosion resistance, they are considered acceptable as corrosion-allowance materials for a thick walled container due to their practical immunity to the localized corrosion phenomena such as stress corrosion cracking, pitting or crevice corrosion. Aiming to improve the performances of these materials, studies on the effect of small additions of Ni and V on the general corrosion are in process. The improvement in the resistance to general corrosion should not be accompanied by a sensitivity to stress corrosion cracking. On the contrary, alfa titanium alloys are considered the most resistant materials to general corrosion in salt brines. However, pitting, are potential deficiencies of this corrosion-resistant materials for a thin walled container. (Author)

  11. Development of Candidate Reference Materials of Endosulfan Sulfate and Bifenthrin in Black Tea

    Directory of Open Access Journals (Sweden)

    Nurhani Aryana

    2016-03-01

    Full Text Available The candidate reference materials of endosulfan sulfate and bifenthrin in black tea have been developed according to the requirements of ISO Guide 34 and 35. Preparation of candidate material includes grinding and sieving of the black tea leaves, spiking the black tea powder by both analytes, homogenization, and bottling. Homogeneity and short-term stability test were performed using a GC-µECD instrument. Meanwhile, the characterization was carried out by a collaborative study using both of GC-µECD and GC-MS instruments. The uncertainty budget was evaluated from sample inhomogeneity, short-term instability and variability in the characterization procedure. In a dry mass fraction, endosulfan sulfate was assigned to be 491 µg kg-1 with a relative expanded uncertainty of ± 33.2%, and bifenthrin was assigned to be 937 µg kg-1 with a relative expanded uncertainty of ± 18.5%. The candidate reference materials are aimed to support the need of matrix CRM especially for the measurement of pesticide residue for quality assurance work done by laboratories in Indonesia.

  12. Experimental assessment of the effects of ELMs and disruptions on ITER divertor armour materials

    International Nuclear Information System (INIS)

    Zhitlukhin, A.; Federici, G.; Giniyatulin, R.; Landman, I.; Linke, J.; Loarte, A.; Merola, M.; Podkovyrov, V.; Safronov, V.

    2005-01-01

    The response of plasma protection materials to thermal energy deposited during simulated Type I Edge Localised Modes (ELMs) and disruptions was studied. The paper describes the design and manufacture of special CFC and tungsten macrobrush targets, the experimental conditions achievable at simulating facilities and results of selected experiments. Experiments are conducted primarily under an EU/RF research collaboration in two plasma guns (QSPA and MK-200UG) located in TRINITI, Troitsk, Russia. The targets were exposed to a large number of repetitive pulses in QSPA plasma gun with heat loads varying in a range of 1-2 MJ/m 2 lasting 0.1-0.5 ms, with the purpose to determine the total expected erosion rate in ITER. MK-200UG experiments were focused on studying mainly vapour plasma production and impurity transport during ELMs. Moderate tungsten erosion less than 0.3 microns per shot was demonstrated for 1.5 MJ/m 2 energy densities. Energy density increasing up to 1.8 MJ/m 2 resulted in sharp growth of tungsten erosion, caused by intensive droplet ejection from irradiated tungsten surface. The program of further experiments is discussed. (author)

  13. Space Environmental Effects Testing and Characterization of the Candidate Solar Sail Material Aluminized Mylar

    Science.gov (United States)

    Edwards, D. L.; Hubbs, W. S.; Wertz, G. E.; Alstatt, R.; Munafo, Paul (Technical Monitor)

    2001-01-01

    The usage of solar sails as a propellantless propulsion system has been proposed for many years. The technical challenges associated with solar sails are fabrication of ultralightweight films, deploying the sails and controlling the spacecraft. Integral to all these challenges is the mechanical property integrity of the sail while exposed to the harsh environment of space. This paper describes testing and characterization of a candidate solar sail material, Aluminized Mylar. This material was exposed to a simulated Geosynchronous Transfer Orbit (GTO) and evaluated by measuring thermooptical and mechanical property changes. Testing procedures and results are presented.

  14. In-situ hot corrosion testing of candidate materials for exhaust valve spindles

    DEFF Research Database (Denmark)

    Bihlet, Uffe; Hoeg, Harro A.; Dahl, Kristian Vinter

    2011-01-01

    The two stroke diesel engine has been continually optimized since its invention more than a century ago. One of the ways to increase fuel efficiency further is to increase the compression ratio, and thereby the temperature in the combustion chamber. Because of this, and the composition of the fuel...... used, exhaust valve spindles in marine diesel engines are subjected to high temperatures and stresses as well as molten salt induced corrosion. To investigate candidate materials for future designs which will involve the HIP process, a spindle with Ni superalloy material samples inserted in a HIPd Ni49...

  15. Value determination of ZrO2 in-house reference material (RM) candidate

    International Nuclear Information System (INIS)

    Susanna Tuning Sunanti; Samin; Supriyanto C

    2013-01-01

    The value determination of zirconium oxide in-house reference materials (RM) candidate has been done by referring to ISO:35-2006 standard. The raw material of RM was 4 kg of ZrO 2 , Merck, that was dried at 90°C for 2×6 hours in a closed room. The samples were crushed with stainless steel (SS) pestle to pass ≤ 200 mesh sieve, homogenized in a homogenizer for 3×6 hours to obtain the powdered, dried and homogenous samples. The gravimetric method was performed to test the moisture content, while XRF and AAS methods were used to test the homogeneity and stability of samples candidates. Reference material (RM) candidates of ZrO 2 powder were put into polyethylene bottles, each weighing 100 g. Samples were distributed to 10 testing laboratories that have been accredited for testing the composition of the oxide contents and loss of ignition (LOI) using variety of analytical methods that have been validated such as AAS, XRF, NAA, and UV-Vis. The testing results of oxide content and loss of ignition parameters from various laboratories were analyzed using statistical methods. The testing data of oxide concentration in zirconium oxide RM candidates obtained from various laboratories were ZrO 2 : 97.7334 ± 0.0016%, HfO 2 : 1.7329 ± 0.0024%, SiO 2 : 30.1224 ± 0.0053%, Al 2 O 3 : 0.0245 ± 0.0015%, TiO 2 : 0.0153 ± 0.0006%, Fe 2 O 3 : 0.0068 ± 0.0005%, CdO: 3.1798 ± 0.00006 ppm, and the LOI results was = 0.0217 ± 0.00022%. (author)

  16. A feasibility study for producing an egg matrix candidate reference material for the polyether ionophore salinomycin.

    Science.gov (United States)

    Ferreira, Rosana Gomes; Monteiro, Mychelle Alves; Pereira, Mararlene Ulberg; da Costa, Rafaela Pinto; Spisso, Bernardete Ferraz; Calado, Veronica

    2016-08-01

    The aim of this work was to study the feasibility of producing an egg matrix candidate reference material for salinomycin. Preservation techniques investigated were freeze-drying and spray drying dehydration. Homogeneity and stability studies of the produced batches were conducted according to ISO Guides 34 and 35. The results showed that all produced batches were homogeneous and both freeze-drying and spray drying techniques were suitable for matrix dehydrating, ensuring the material stability. In order to preserve the material integrity, it must be transported within the temperature range of -20 up to 25°C. The results constitute an important step towards the development of an egg matrix reference material for salinomycin is possible. Copyright © 2016 Elsevier B.V. All rights reserved.

  17. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Bullen, D.B.; Gdowski, G.E.; Weiss, H.

    1988-06-01

    Three copper-based alloys, CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni), are being considered along with three austenitic candidates as possible materials for fabrication of containers for disposal of high-level radioactive waste. The waste will include spent fuel assemblies from reactors as well as high-level reprocessing wastes in borosilicate glass and will be sent to the prospective repository at Yucca Mountain, Nevada, for disposal. The containers must maintain mechanical integrity for 50 yr after emplacement to allow for retrieval of waste during the preclosure phase of repository operation. Containment is required to be substantially complete for up to 300 to 1000 yr. During the early period, the containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. The final closure joint will be critical to the integrity of the containers. This volume surveys the available data on the metallurgy of the copper-based candidate alloys and the welding techniques employed to join these materials. The focus of this volume is on the methods applicable to remote-handling procedures in a hot-cell environment with limited possibility of postweld heat treatment. The three copper-based candidates are ranked on the basis of the various closure techniques. On the basis of considerations regarding welding, the following ranking is proposed for the copper-based alloys: CDA 715 (best) > CDA 102 > CDA 613 (worst). 49 refs., 15 figs., 1 tab

  18. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    Energy Technology Data Exchange (ETDEWEB)

    Ritz, Guillaume Henri

    2011-07-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m{sup -2} as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m{sup -2}. The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform

  19. Performance of tungsten-based materials and components under ITER and DEMO relevant steady-state thermal loads

    International Nuclear Information System (INIS)

    Ritz, Guillaume Henri

    2011-01-01

    In nuclear fusion devices the surfaces directly facing the plasma are irradiated with high energy fluxes. The most intense loads are deposited on the divertor located at the bottom of the plasma chamber, which has to withstand continuous heat loads with a power density of several MW . m -2 as well as transient events. These are much shorter (in the millisecond and sub-millisecond regime) but deposit a higher power densities of a few GW . m -2 . The search for materials that can survive to those severe loading conditions led to the choice of tungsten which possesses advantageous attributes such as a high melting point, high thermal conductivity, low thermal expansion and an acceptable activation rate. These properties made it an attractive and promising candidate as armor material for divertors of future fusion devices such as ITER and DEMO. For the DEMO divertor, conceptual studies on helium-cooled tungsten plasma-facing components were performed. The concept was realized and tested under DEMO specific cyclic thermal loads. The examination of the plasma-facing components by microstructural analyses before and after thermal loading enabled to determine the mechanisms for components failure. Among others, it clearly showed the impact of the tungsten grade and the thermal stress induced crack formation on the performance of the armor material and in general of the plasma-facing component under high heat loads. A tungsten qualification program was launched to study the behaviour of various tungsten grades, in particular the crack formation, under fusion relevant steady-state thermal loads. In total, seven commercially available materials from two industrial suppliers were investigated. As the material's thermal response is strongly related to its microstructure, this program comprised different material geometries and manufacturing technologies. It also included the utilization of an actively cooled specimen holder which has been designed to perform sophisticated

  20. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-04-01

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

  1. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Bullen, D.B.

    1988-04-01

    Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs

  2. Fissile material disposition program: Screening of alternate immobilization candidates for disposition of surplus fissile materials

    International Nuclear Information System (INIS)

    Gray, L.W.

    1996-01-01

    With the end of the Cold War, the world faces for the first time the need to dismantle vast numbers of ''excess'' nuclear weapons and dispose of the fissile materials they contain, together with fissile residues in the weapons production complex left over from the production of these weapons. If recently agreed US and Russian reductions are fully implemented, tens of thousands of nuclear weapons, containing a hundred tons or more of plutonium and hundreds of tonnes* of highly enriched uranium (HEU), will no longer be needed worldwide for military purposes. These two materials are the essential ingredients of nuclear weapons, and limits on access to them are the primary technical barrier to prospective proliferants who might desire to acquire a nuclear weapons capability. Theoretically, several kilograms of plutonium, or several times that amount of HEU, is sufficient to make a nuclear explosive device. Therefore, these materials will continue to be a potential threat to humanity for as long as they exist

  3. A study on homogeneity of the IAEA candidate reference materials for microanalysis and analytical support in the certification of these materials

    International Nuclear Information System (INIS)

    Dybczynski, R.; Danko, B.; Polkowska-Motrenko, H.

    2002-01-01

    In this paper a study on homogeneity of new IAEA candidate reference materials: IAEA 338 Lichen and IAEA 413 Algae in small (ca.10 mg) samples as well as some data contributing to certification of these materials are presented. (author)

  4. Micro-homogeneity evaluation of a bovine kidney candidate reference material

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Liliana; Moreira, Edson G.; Vasconcellos, Marina B.A., E-mail: lcastroesnal@usp.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The minimum sample intake for which a reference material remains homogeneous is one of the parameters that must be estimated in the homogeneity assessment study of reference materials. In this work, Instrumental Neutron Activation Analysis was used to evaluate this quantity in a bovine kidney candidate reference material. The mass fractions of 9 inorganic constituents were determined in subsamples between 1 and 2 mg in order to estimate the relative homogeneity factor (HE) and the minimum sample mass to achieve 5% and 10% precision on a 95% confidence level. Results obtained for H{sub E} in all the analyzed elements were satisfactory. The estimated minimum sample intake was between 2 mg and 40 mg, depending on the element. (author)

  5. Post-Irradiation Properties of Candidate Materials for High-Power Targets

    International Nuclear Information System (INIS)

    Kirk, H.G.; Ludewig, H.; Mausner, L.F.; Simos, N.; Thieberger, P.; Brookhaven; Hayato, Y.; Yoshimura, K.; McDonald, K.T.; Sheppard, J.; Trung, L.P.

    2006-01-01

    The desire of the high-energy-physics community for more intense secondary particle beams motivates the development of multi-megawatt, pulsed proton sources. The targets needed to produce these secondary particle beams must be sufficiently robust to withstand the intense pressure waves arising from the high peak-energy deposition which an intense pulsed beam will deliver. In addition, the materials used for the targets must continue to perform in a severe radiation environment. The effect of the beam-induced pressure waves can be mitigated by use of target materials with high-yield strength and/or low coefficient of thermal expansion (CTE) [1, 2, 3]. We report here first results of an expanded study of the effects of irradiation on several additional candidate materials with high strength (AlBeMet, beryllium, Ti-V6-Al4) or low CTE (a carbon-carbon composite, a new Toyota ''gum'' metal alloy [4], Super-Invar)

  6. Scoping corrosion tests on candidate waste package basket materials for the Yucca Mountain project

    International Nuclear Information System (INIS)

    Konynenburg, R.A. van; Curtis, P.G.; Summers, T.S.E.

    1998-03-01

    A scoping corrosion test was performed on candidate waste package basket materials. The corrosion medium was a pH-buffered solution of chemical species expected to be produced by radiolysis. The test was conducted at 90 C for 96 hours. Samples included aluminum-, copper-, stainless steel- and zirconium-based metallic materials and several ceramics, incorporating neutron-absorbing elements. Sample weight losses and solution chemical changes were measured. Both corrosion of the host materials and dissolution of the neutron-absorbing elements were studied. The ceramics and the zirconium-based materials underwent only minor corrosion. The stainless steel-based materials performed well except for a welded sample. The aluminum- and copper-based materials exhibited the highest corrosion rates. Boron dissolution depends on its chemical form. Boron oxide and many metal borides dissolve readily in acidic solutions while high-chromium borides and boron carbide, though thermodynamically unstable, exhibit little dissolution in short times. The results of solution chemical analyses were consistent with this. Gadolinium did not dissolve significantly from monazite, and hafnium showed little dissolution from a variety of host materials, in keeping with its low solubility

  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. Corrosion of candidate iron-base waste package structural barrier materials in moist salt environments

    International Nuclear Information System (INIS)

    Westerman, R.E.; Pitman, S.G.

    1984-11-01

    Mild steels are considered to be strong candidates for waste package structural barrier (e.g., overpack) applications in salt repositories. Corrosion rates of these materials determined in autoclave tests utilizing a simulated intrusion brine based on Permian Basin core samples are low, generally <25 μm (1 mil) per year. When the steels are exposed to moist salts containing simulated inclusion brines, the corrosion rates are found to increase significantly. The magnesium in the inclusion brine component of the environment is believed to be responsible for the increased corrosion rates. 1 reference, 4 figures, 2 tables

  9. Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhtin, V.P.; Konkashbaev, I.; Landman, I.; Safronov, V.M.; Toporkov, D.A.; Zhitlukhin, A.M.

    1995-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. ((orig.))

  10. Evaluation of the cryogenic mechanical properties of the insulation material for ITER Feeder superconducting joint

    Science.gov (United States)

    Wu, Zhixiong; Huang, Rongjin; Huang, ChuanJun; Yang, Yanfang; Huang, Xiongyi; Li, Laifeng

    2017-12-01

    The Glass-fiber reinforced plastic (GFRP) fabricated by the vacuum bag process was selected as the high voltage electrical insulation and mechanical support for the superconducting joints and the current leads for the ITER Feeder system. To evaluate the cryogenic mechanical properties of the GFRP, the mechanical properties such as the short beam strength (SBS), the tensile strength and the fatigue fracture strength after 30,000 cycles, were measured at 77K in this study. The results demonstrated that the GFRP met the design requirements of ITER.

  11. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Gdowski, G.E.

    1988-06-01

    Three copper-based alloys, CDA 102 (oxygen-free, high-purity copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni), are candidates for the fabrication of high-level radioactive-waste disposal containers. Waste will include spent fuel assemblies from reactors as well as borosilicate glass, and will be sent to the prospective repository site at Yucca Mountain in Nye County, Nevada. The decay of radionuclides will result in the generation of substantial heat and in fluxes of gamma radiation outside the containers. In this environment, container materials might degrade by atmospheric oxidation, general aqueous phase corrosion, localized corrosion (LC), and stress corrosion cracking (SCC). This volume is a critical survey of available data on pitting and crevice corrosion of the copper-based candidates. Pitting and crevice corrosion are two of the most common forms of LC of these materials. Data on the SCC of these alloys is surveyed in Volume 4. Pitting usually occurs in water that contains low concentrations of bicarbonate and chloride anions, such as water from Well J-13 at the Nevada Test Site. Consequently, this mode of degradation might occur in the repository environment. Though few quantitative data on LC were found, a tentative ranking based on pitting corrosion, local dealloying, crevice corrosion, and biofouling is presented. CDA 102 performs well in the categories of pitting corrosion, local dealloying, and biofouling, but susceptibility to crevice corrosion diminishes its attractiveness as a candidate. The cupronickel alloy, CDA 715, probably has the best overall resistance to such localized forms of attack. 123 refs., 11 figs., 3 tabs

  12. Dimethyl terephthalate (DMT) as a candidate phase change material for high temperature thermal energy storage

    Energy Technology Data Exchange (ETDEWEB)

    Kuecuekaltun, Engin [Advansa Sasa Polyester San, A.S., Adana (Turkey); Paksoy, Halime; Bilgin, Ramazan; Yuecebilgic, Guezide [Cukurova Univ., Adana (Turkey). Chemistry Dept.; Evliya, Hunay [Cukurova Univ., Adana (Turkey). Center for Environmental Research

    2010-07-01

    Thermal energy storage at elevated temperatures, particularly in the range of 120-250 C is of interest with a significant potential for industrial applications that use process steam at low or intermediate pressures. At given temperature range there are few studies on thermal energy storage materials and most of them are dedicated to sensible heat. In this study, Dimethyl Terephthalate - DMT (CAS No: 120-61-6) is investigated as a candidate phase change material (PCM) for high temperature thermal energy storage. DMT is a monomer commonly used in Polyethylene terephtalate industry and has reasonable cost and availability. The Differential Scanning Calorimetry (DSC) analysis and heating cooling curves show that DMT melts at 140-146 C within a narrow window. Supercooling that was detected in DSC results was not observed in the cooling curve measurements made with a larger sample. With a latent heat of 193 J/g, DMT is a candidate PCM for high temperature storage. Potential limitations such as, low thermal conductivity and sublimation needs further investigation. (orig.)

  13. Global blending optimization of laminated composites with discrete material candidate selection and thickness variation

    DEFF Research Database (Denmark)

    Sørensen, Søren N.; Stolpe, Mathias

    2015-01-01

    rate. The capabilities of the method and the effect of active versus inactive manufacturing constraints are demonstrated on several numerical examples of limited size, involving at most 320 binary variables. Most examples are solved to guaranteed global optimality and may constitute benchmark examples...... but is, however, convex in the original mixed binary nested form. Convexity is the foremost important property of optimization problems, and the proposed method can guarantee the global or near-global optimal solution; unlike most topology optimization methods. The material selection is limited...... for popular topology optimization methods and heuristics based on solving sequences of non-convex problems. The results will among others demonstrate that the difficulty of the posed problem is highly dependent upon the composition of the constitutive properties of the material candidates....

  14. Creep rupture behavior of candidate materials for nuclear process heat applications

    International Nuclear Information System (INIS)

    Schubert, F.; te Heesen, E.; Bruch, U.; Cook, R.; Diehl, H.; Ennis, P.J.; Jakobeit, W.; Penkalla, H.J.; Ullrich, G.

    1984-01-01

    Creep and stress rupture properties are determined for the candidate materials to be used in hightemperature gas-cooled reactor (HTGR) components. The materials and test methods are briefly described based on experimental results of test durations of about20000 h. The medium creep strengths of the alloys Inconel-617, Hastelloy-X, Nimonic-86, Hastelloy-S, Manaurite-36X, IN-519, and Incoloy-800H are compared showing that Inconel-617 has the best creep rupture properties in the temperature range above 800 0 C. The rupture time of welded joints is in the lower range of the scatterband of the parent metal. The properties determined in different simulated HTGR atmospheres are within the scatterband of the properties obtained in air. Extrapolation methods are discussed and a modified minimum commitment method is favored

  15. Study on corrosion behavior of candidate materials in 650℃ supercritical water

    International Nuclear Information System (INIS)

    Ma Shuli; Luo Ying; Zhang Qiang; Wang Hao; Qiu Shaoyu

    2014-01-01

    The general corrosion behavior of three candidate materials (347, HR3C and In-718) was investigated in 650 ℃/25 MPa deionized water. Morphology and composition of the surface oxide film with different exposure time were observed through FEG-SEM and EDS. The phase constitute was analyzed by GIXRD. For all the test materials, the weight loss follows typical parabolic law and the weight loss of 347 shows more than 40 times higher than that of HR3C and In-718. The oxide film of three alloys mainly consists of Ni(Cr, Fe) 2 O 4 . In-718 shows severe pitting and the oxide film of 347 appears significant spalling, while HR3C has compact oxide film. In the high temperature supercritical water, the high Cr content may enhance the general corrosion property of the alloys, while addition of Nb may be detrimental to the pitting resistance of alloys. (authors)

  16. Accelerator-Based PIXE and STIM Analysis of Candidate Solar Sail Materials

    International Nuclear Information System (INIS)

    Hollerman, W.A.; Stanaland, T.L.; Boudreaux, P.; Elberson, L.; Fontenot, J.; Gates, E.; Greco, R.; McBride, M.; Woodward, A.; Edwards, D.

    2003-01-01

    Solar sailing is a unique form of propulsion where a spacecraft gains momentum from incident photons. A totally reflective sail experiences a pressure of 9.1 μPa at a distance of 1 AU from the Sun. Since sails are not limited by reaction mass, they provide continual acceleration, reduced only by the lifetime of the lightweight film in the space environment and the distance to the Sun. Practical solar sails can expand the number of possible missions, enabling new concepts that are difficult by conventional means. One of the current challenges is to develop strong, lightweight, and radiation resistant sail materials. This paper will discuss initial results from a Particle Induced X-Ray Emission (PIXE) and Scanning Transmission Ion Microscopy (STIM) analysis of candidate solar sail materials

  17. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

    1995-06-01

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

  18. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

    International Nuclear Information System (INIS)

    Vinson, D.W.; Nutt, W.M.; Bullen, D.B.

    1995-06-01

    Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27

  19. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    Energy Technology Data Exchange (ETDEWEB)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D. [Laboratoire PROMES-CNRS, Laboratoire Procedes, Materiaux et Energie Solaire, 7 rue du Four Solaire, 66120 Font-Romeu Odeillo (France)

    2007-07-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H{sup +} ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  20. Divertor materials for ITER - Tungsten and carbon/carbon composite behavior under coupled ionic irradiation and high temperature

    International Nuclear Information System (INIS)

    Raunier, S.; Balat-Pichelin, M.; Sans, J.L.; Hernandez, D.

    2007-01-01

    Full text of publication follows: In the frame of the International Thermonuclear Experimental Reactor ITER, the physical-chemical characterization of plasma-facing components (divertor and structural materials) is essential because they are subjected to simultaneous high thermal and ionic fluxes. In this paper, an experimental and theoretical study of the physical-chemical behavior of carbon/carbon composite and tungsten (materials for ITER divertor) under extreme conditions is performed. The simulation of the interaction of hydrogen ions with the material, the theoretical study of physical erosion (TRIM and TRIDYN codes) and the chemical erosion (GEMINI code) are carried out. The conditions of nominal or accidental mode that can occur during the operation of the reactor (high temperature 1300 - 2500 K, high vacuum, H + ionic flux with different energies) are experimentally simulated. In this work, we have studied the material degradation, the mass loss kinetics, the characterization of the emitted neutral and charged species of heated and both heated and irradiated materials, and the determination of the thermo-radiative properties versus time. This study, done in collaboration with CEA Cadarache, is realized using the MEDIASE experimental device (Moyen d'Essai et de Diagnostic en Ambiance Solaire Extreme) located at the focus of the 1000 kW solar furnace of PROMES-CNRS laboratory in Odeillo. Material characterization pre- and post-processing is performed with classical techniques as SEM, XRD and XPS and also by measuring the BRDF (Bidirectional Reflectivity Diffusion Function). (authors)

  1. Selection and evaluation of inner material candidates for Spanish high level radioactive waste canisters

    International Nuclear Information System (INIS)

    Puig, Francesc; Dies, Javier; Sevilla, Manuel; Pablo, Joan de; Pueyo, Juan Jose; Miralles, Lourdes; Martinez-Esparza, Aurora

    2007-01-01

    This paper summarizes the work carried out to analyse different alternatives related to the inner material selection of the Spanish high level waste canister for long term storage. The preliminary repository design considers granitic or clay formations, compacted bentonite sealing, corrosion allowing steel canisters and glass bead filling between the fuel assemblies and canister walls. This filling material will have the primary role of avoiding the possibility of a criticality event, which becomes an issue of major importance once the container is finally breached by corrosion and flooded by groundwater. In the first place, a complete set of requirements have been devised as evaluation criteria for candidate materials examination and selection; resulting in a compilation of demands significantly deeper and more exhaustive than any other similar work found in literature, including over 20 requirements and some other general aspects that could involve improvements in repository performance. Secondly, eight materials or material families (cast iron or steel, borosilicate glass, spinel, depleted uranium, dehydrated zeolites, hematite, phosphates and olivine) have been chosen and examined in detail, extracting some relevant conclusions. Either cast iron, borosilicate glass, spinel or depleted uranium are considered to look quite promising for the mentioned purpose. (authors)

  2. Krylov Iterative Methods and the Degraded Effectiveness of Diffusion Synthetic Acceleration for Multidimensional SN Calculations in Problems with Material Discontinuities

    International Nuclear Information System (INIS)

    Warsa, James S.; Wareing, Todd A.; Morel, Jim E.

    2004-01-01

    A loss in the effectiveness of diffusion synthetic acceleration (DSA) schemes has been observed with certain S N discretizations on two-dimensional Cartesian grids in the presence of material discontinuities. We will present more evidence supporting the conjecture that DSA effectiveness will degrade for multidimensional problems with discontinuous total cross sections, regardless of the particular physical configuration or spatial discretization. Fourier analysis and numerical experiments help us identify a set of representative problems for which established DSA schemes are ineffective, focusing on diffusive problems for which DSA is most needed. We consider a lumped, linear discontinuous spatial discretization of the S N transport equation on three-dimensional, unstructured tetrahedral meshes and look at a fully consistent and a 'partially consistent' DSA method for this discretization. The effectiveness of both methods is shown to degrade significantly. A Fourier analysis of the fully consistent DSA scheme in the limit of decreasing cell optical thickness supports the view that the DSA itself is failing when material discontinuities are present in a problem. We show that a Krylov iterative method, preconditioned with DSA, is an effective remedy that can be used to efficiently compute solutions for this class of problems. We show that as a preconditioner to the Krylov method, a partially consistent DSA method is more than adequate. In fact, it is preferable to a fully consistent method because the partially consistent method is based on a continuous finite element discretization of the diffusion equation that can be solved relatively easily. The Krylov method can be implemented in terms of the original S N source iteration coding with only slight modification. Results from numerical experiments show that replacing source iteration with a preconditioned Krylov method can efficiently solve problems that are virtually intractable with accelerated source iteration

  3. A simplified in vivo approach for evaluating the bioabsorbable behavior of candidate stent materials.

    Science.gov (United States)

    Pierson, Daniel; Edick, Jacob; Tauscher, Aaron; Pokorney, Ellen; Bowen, Patrick; Gelbaugh, Jesse; Stinson, Jon; Getty, Heather; Lee, Chee Huei; Drelich, Jaroslaw; Goldman, Jeremy

    2012-01-01

    Metal stents are commonly used to revascularize occluded arteries. A bioabsorbable metal stent that harmlessly erodes away over time may minimize the normal chronic risks associated with permanent implants. However, there is no simple, low-cost method of introducing candidate materials into the arterial environment. Here, we developed a novel experimental model where a biomaterial wire is implanted into a rat artery lumen (simulating bioabsorbable stent blood contact) or artery wall (simulating bioabsorbable stent matrix contact). We use this model to clarify the corrosion mechanism of iron (≥99.5 wt %), which is a candidate bioabsorbable stent material due to its biocompatibility and mechanical strength. We found that iron wire encapsulation within the arterial wall extracellular matrix resulted in substantial biocorrosion by 22 days, with a voluminous corrosion product retained within the vessel wall at 9 months. In contrast, the blood-contacting luminal implant experienced minimal biocorrosion at 9 months. The importance of arterial blood versus arterial wall contact for regulating biocorrosion was confirmed with magnesium wires. We found that magnesium was highly corroded when placed in the arterial wall but was not corroded when exposed to blood in the arterial lumen for 3 weeks. The results demonstrate the capability of the vascular implantation model to conduct rapid in vivo assessments of vascular biomaterial corrosion behavior and to predict long-term biocorrosion behavior from material analyses. The results also highlight the critical role of the arterial environment (blood vs. matrix contact) in directing the corrosion behavior of biodegradable metals. Copyright © 2011 Wiley Periodicals, Inc.

  4. Application of INAA complementary gamma ray photopeaks to homogeneity study of candidate reference materials

    International Nuclear Information System (INIS)

    Moreira, Edson G.; Vasconcellos, Marina B.A.; Lima, Ana P.S.; Catharino, Marilia G.M.; Maihara, Vera A.; Saiki, Mitiko

    2009-01-01

    Characterization and certification of reference materials, RMs, is a complex task involving many steps. One of them is the homogeneity testing to assure that key property values will not present variation among RM bottles. Good precision is the most important figure of merit of an analytical technique to allow it to be used in the homogeneity testing of candidate RMs. Due to its inherent characteristics, Instrumental Neutron Activation Analysis, INAA, is an analytical technique of choice for homogeneity testing. Problems with sample digestion and contamination from reagents are not an issue in INAA, as solid samples are analyzed directly. For element determination via INAA, the activity of a suitable gamma ray decay photopeak for an element is chosen and it is compared to the activity of a standard of the element. An interesting possibility is the use of complementary gamma ray photopeaks (for the elements that present them) to confirm the homogeneity test results for an element. In this study, an investigation of the use of the complementary gamma ray photopeaks of 110 mAg, 82 Br, 60 Co, 134 Cs, 152 Eu, 59 Fe, 140 La, 233 Pa (for Th determination), 46 Sc and 75 Se radionuclides was undertaken in the between bottle homogeneity study of a mussel candidate RM under preparation at IPEN - CNEN/SP. Although some photopeaks led to biased element content results, the use of complementary gamma ray photopeaks proved to be helpful in supporting homogeneity study conclusions of new RMs. (author)

  5. Tritium absorption and desorption in ITER relevant materials: comparative study of tungsten dust and massive samples

    Energy Technology Data Exchange (ETDEWEB)

    Grisolia, C., E-mail: christian.grisolia@cea.fr [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Hodille, E. [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Chene, J.; Garcia-Argote, S.; Pieters, G.; El-Kharbachi, A. [CEA Saclay, SCBM, iBiTec-S, PC n° 108, 91191 Gifsur-Yvette (France); Marchetti, L.; Martin, F.; Miserque, F. [CEA Saclay, DEN/DPC/SCCME/LECA, F-91191 Gif-sur-Yvette (France); Vrel, D.; Redolfi, M. [LSPM, Université Paris 13, Sorbonne Paris Cité, UPR 3407 CNRS, 93430 Villetaneuse (France); Malard, V. [CEA, DSV, IBEB, Lab Biochim System Perturb, Bagnols-sur-Cèze F-30207 (France); Dinescu, G.; Acsente, T. [NILPRP, 409 Atomistilor Street, 77125 Magurele, Bucharest (Romania); Gensdarmes, F.; Peillon, S. [IRSN, PSN-RES/SCA/LPMA, Saclay, Gif-sur-Yvette, 91192 (France); Pegourié, B. [CEA, IRFM, F-13108 Saint Paul lez Durance (France); Rousseau, B. [CEA Saclay, SCBM, iBiTec-S, PC n° 108, 91191 Gifsur-Yvette (France)

    2015-08-15

    Tritium adsorption and desorption from well characterized tungsten dust are presented. The dust used are of different types prepared by planetary milling and by aggregation technique in plasma. For the milled powder, the surface specific area (SSA) is 15.5 m{sup 2}/g. The particles are poly-disperse with a maximum size of 200 nm for the milled powder and 100 nm for the aggregation one. Prior to tritiation the particles are carefully de-oxidized. Both samples are experiencing a high tritium inventory from 5 GBq/g to 35 GBq/g. From comparison with massive samples and considering that tritium inventory increases with SSA, it is shown that surface effects are predominant in the tritium trapping process. Extrapolation to the ITER environment is undertaken with the help of a Macroscopic Rate Equation model. It is shown that, during the life time of ITER, these particles can exceed rapidly 1 GBq/g.

  6. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Gdowski, G.E.; Bullen, D.B.

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials [CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)], which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs

  7. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

  8. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    Energy Technology Data Exchange (ETDEWEB)

    Strum, M.J.; Weiss, H.; Farmer, J.C. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

    1988-06-01

    This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs.

  9. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Strum, M.J.; Weiss, H.; Farmer, J.C.; Bullen, D.B.

    1988-06-01

    This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs

  10. Overview of material migration and mixing, fuel retention and cleaning of ITER-like castellated structures in TEXTOR

    International Nuclear Information System (INIS)

    Litnovsky, A.; Philipps, V.; Wienhold, P.; Kreter, A.; Kirschner, A.; Matveev, D.; Brezinsek, S.; Sergienko, G.; Pospieszczyk, A.; Schweer, B.; Schulz, C.; Schmitz, O.; Coenen, J.W.; Samm, U.; Krieger, K.; Hirai, T.; Emmoth, B.; Rubel, M.; Bazylev, B.; Breuer, U.

    2011-01-01

    Plasma-facing components (PFCs) in ITER will be castellated by splitting them into small-size blocks to maintain the thermo-mechanical stability. However, there are concerns in particular on retention of codeposited radioactive fuel in the gaps. An R and D program is underway in TEXTOR addressing this acute issue of castellation. Material migration and fuel inventory are investigated using long- and short-term discharge-resolved experiments with castellated structures in TEXTOR. Significant impurity transport to the gaps was detected and results were in part quantitatively reproduced with 3D-GAPS code. Deposits containing up to 70 at.% of tungsten on the gap areas closest to the plasma were detected in recent experiments. Deposition in the gaps accompanied by metal mixing demand for development of effective cleaning techniques. In experiments with ITER-like castellation, the gaps were cleaned from carbonaceous deposits using oxygen plasmas at 350 deg. C. This contribution contains an overview of experimental and modeling results along with recommendations for PFCs in ITER.

  11. Short-term stability test for thorium soil candidate a reference material

    Energy Technology Data Exchange (ETDEWEB)

    Clain, Almir F.; Fonseca, Adelaide M.G.; Dantas, Vanessa V.D.B.; Braganca, Maura J.C.; Souza, Poliana S., E-mail: almir@ird.gov.br, E-mail: adelaide@ird.gov.br, E-mail: vanessa@ird.gov.br, E-mail: maura@ird.gov.br, E-mail: poliana@bolsista.ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    This work describes a methodology to determine the soil short-term stability after the steps of production in laboratory. The short-term stability of the soil is an essential property to be determined in order to producing a reference material. The soil is a candidate of reference material for chemical analysis of thorium with metrological traceability to be used in environmental analysis, equipment calibration, validation methods, and quality control. A material is considered stable in a certain temperature if the property of interest does not change with time, considering the analytical random fluctuations. Due to this, the angular coefficient from the graphic of Th concentration versus elapsed time must be near to zero. The analytical determinations of thorium concentration were performed by Instrumental Neutron activation Analysis. The slopes and their uncertainties were obtained from the regression lines at temperatures of 20 deg C and 60 deg C, with control temperature of -20 deg C. From the obtained data a t-test was applied. In both temperatures the calculated t-value was lower than the critical value, so we can conclude with 95% confidence level that no significant changes happened during the period studied concerning thorium concentration in soil at temperatures of 20 deg C and 60 deg C, showing stability at these temperatures. (author)

  12. Emissivity of Candidate Materials for VHTR Applicationbs: Role of Oxidation and Surface Modification Treatments

    International Nuclear Information System (INIS)

    Sridharan, Kumar; Allen, Todd; Anderson, Mark; Cao, Guoping; Kulcinski, Gerald

    2011-01-01

    The Generation IV (GEN IV) Nuclear Energy Systems Initiative was instituted by the Department of Energy (DOE) with the goal of researching and developing technologies and materials necessary for various types of future reactors. These GEN IV reactors will employ advanced fuel cycles, passive safety systems, and other innovative systems, leading to significant differences between these future reactors and current water-cooled reactors. The leading candidate for the Next Generation Nuclear Plant (NGNP) to be built at Idaho National Lab (INL) in the United States is the Very High Temperature Reactor (VHTR). Due to the high operating temperatures of the VHTR, the Reactor Pressure Vessel (RPV) will partially rely on heat transfer by radiation for cooling. Heat expulsion by radiation will become all the more important during high temperature excursions during off-normal accident scenarios. Radiant power is dictated by emissivity, a material property. The NGNP Materials Research and Development Program Plan (1) has identified emissivity and the effects of high temperature oxide formation on emissivity as an area of research towards the development of the VHTR.

  13. VUV photoemission studies of candidate Large Hadron Collider vacuum chamber materials

    CERN Document Server

    Cimino, R; Baglin, V

    1999-01-01

    In the context of future accelerators and, in particular, the beam vacuum of the Large Hadron Collider (LHC), a 27 km circumference proton collider to be built at CERN, VUV synchrotron radiation (SR) has been used to study both qualitatively and quantitatively candidate vacuum chamber materials. Emphasis is given to show that angle and energy resolved photoemission is an extremely powerful tool to address important issues relevant to the LHC, such as the emission of electrons that contributes to the creation of an electron cloud which may cause serious beam instabilities and unmanageable heat loads on the cryogenic system. Here we present not only the measured photoelectron yields from the proposed materials, prepared on an industrial scale, but also the energy and in some cases the angular dependence of the emitted electrons when excited with either a white light (WL) spectrum, simulating that in the arcs of the LHC, or monochromatic light in the photon energy range of interest. The effects on the materials ...

  14. Short-term stability test for thorium soil candidate a reference material

    International Nuclear Information System (INIS)

    Clain, Almir F.; Fonseca, Adelaide M.G.; Dantas, Vanessa V.D.B.; Braganca, Maura J.C.; Souza, Poliana S.

    2015-01-01

    This work describes a methodology to determine the soil short-term stability after the steps of production in laboratory. The short-term stability of the soil is an essential property to be determined in order to producing a reference material. The soil is a candidate of reference material for chemical analysis of thorium with metrological traceability to be used in environmental analysis, equipment calibration, validation methods, and quality control. A material is considered stable in a certain temperature if the property of interest does not change with time, considering the analytical random fluctuations. Due to this, the angular coefficient from the graphic of Th concentration versus elapsed time must be near to zero. The analytical determinations of thorium concentration were performed by Instrumental Neutron activation Analysis. The slopes and their uncertainties were obtained from the regression lines at temperatures of 20 deg C and 60 deg C, with control temperature of -20 deg C. From the obtained data a t-test was applied. In both temperatures the calculated t-value was lower than the critical value, so we can conclude with 95% confidence level that no significant changes happened during the period studied concerning thorium concentration in soil at temperatures of 20 deg C and 60 deg C, showing stability at these temperatures. (author)

  15. Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

    2017-04-15

    Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

  16. Behaviour of candidate materials for fusion applications under high surface heat loads

    International Nuclear Information System (INIS)

    Bolt, H.; Nickel, H.; Kuroda, T.; Miyahara, A.

    1988-07-01

    High heat fluxes to in-vessel components of nuclear fusion devices (tokamaks) during normal operation and abnormal operation conditions are one of the governing issues in the selection of a plasma facing material and the design of first wall components. Their failure under high heat loads during service can severely influence the further operability of the entire fusion device. In order to determine the response of candidate materials to high heat fluxes an experimental program was carried out using the 10 MW Neutral Beam Injection Test Stand of the Institute for Plasma Physics of Nagoya University. Metal samples, 13 different fine grain graphites, carbon - carbon composites, and pyrolytic carbon samples were subjected to heat loads between 16 and 117 MW/m 2 and pulse durations of 50 to 950 ms. Afterwards the resulting structural changes as well as threshold values for the occurance of material damage were determined. The main damage observed on carbon materials was cracking in the case of graphites and pyrolytic carbon and erosion in the case of graphites and carbon - carbon composites. Processes leading to such damage were discussed and described in form of models. Parallel to these laboratory experiments numerical analyses of the response of graphite materials to high heat fluxes were carried out. The results are in general agreement with the experimentally determined values. In order to verify the results from experiments and numerical analyses, graphite test limiters were exposed to about 900 discharges in the JIPP T-IIU tokamak. These proof tests fully confirmed the results obtained. (orig.) [de

  17. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Bullen, D.B.; Gdowski, G.E.

    1988-08-01

    Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs

  18. Inclusion and difusion studies of D in fusion breeding blanket candidate materials

    Energy Technology Data Exchange (ETDEWEB)

    Fan, L.

    2015-07-01

    Deuterium-Tritium (D-T) reaction is the most practical fusion reaction on the way to harness fusion energy. As tritium presents trace quantities on Earth [1], tritium fuel is essential to be generated simultaneously with the D-T reaction in a commerical fusion power plant. Tritium can be obtained in the lithium contained breeding blanket as a transmutation product of nuclear reaction 6Li (n, a)T. Li2T iO3 is considered to be one promising candidate solid tritium breeder material, due to its high lithium density, low activation, compatiblity with structure materials and high chemical stability. The tritium generated in Li2T iO3 breeding blanket needs to be collected and recycled back to the fusion reaction. Therefore, the study of the diffusion characteristic of breeder material Li2T iO3 is necessary to determine tritium mobility and tritium extraction efficiency. In order to study tritium release mechanism of Li2T iO3 breeding material in a fusion power plant environment, a fusion like neutron spectrum is essential while it is now not availble in any laboratory. One alternative is using ion accelerator or implantor to get energetic hydrogenic (H,D,T) ions impacting on breeding material, to simulate the tritium distribution situation. Because of the radioactive property of tritium which will complicate processing procedure, another isotope of hydrogen Deuterium is actually used to be studied. The defect structure in Li2T iO3, due to reactor exposure to fusion generated particles and ? ray irradiation, is achieved by energetic Ti ions. SRIM program is implemented to simulate the D ion or Ti ion distributions after bombarding, as well as the defects. X-ray diffraction technique helps to identify phase compositions. Transmission electron microscopy technique is used to observe the microstructures (Author)

  19. Corrosion susceptibility study of candidate pin materials for ALTC (active lithium/thionyl chloride) batteries. [Active lithium/thionyl chloride

    Energy Technology Data Exchange (ETDEWEB)

    Bovard, F.S.; Cieslak, W.R.

    1987-09-01

    (ALTC = active lithium/thionyl chloride.) We have investigated the corrosion susceptibilities of eight alternate battery pin materials in 1.5M LiAlCl/sub 4//SOCl/sub 2/ electrolyte using ampule exposure and electrochemical tests. The thermal expansion coefficients of these candidate materials are expected to match Sandia-developed Li-corrosion resistant glasses. The corrosion resistances of the candidate materials, which included three stainless steels (15-5 PH, 17-4 PH, and 446), three Fe-Ni glass sealing alloys (Kovar, Alloy 52, and Niromet 426), a Ni-based alloy (Hastelloy B-2) and a zirconium-based alloy (Zircaloy), were compared to the reference materials Ni and 316L SS. All of the candidate materials showed some evidence of corrosion and, therefore, did not perform as well as the reference materials. The Hastelloy B-2 and Zircaloy are clearly unacceptable materials for this application. Of the remaining alternate materials, the 446 SS and Alloy 52 are the most promising candidates.

  20. Comparison of candidate materials for a synthetic osteo-odonto keratoprosthesis device.

    Science.gov (United States)

    Tan, Xiao Wei; Perera, A Promoda P; Tan, Anna; Tan, Donald; Khor, K A; Beuerman, Roger W; Mehta, Jodhbir S

    2011-01-05

    Osteo-odonto keratoprosthesis is one of the most successful forms of keratoprosthesis surgery for end-stage corneal and ocular surface disease. There is a lack of detailed comparison studies on the biocompatibilities of different materials used in keratoprosthesis. The aim of this investigation was to compare synthetic bioinert materials used for keratoprosthesis surgery with hydroxyapatite (HA) as a reference. Test materials were sintered titanium oxide (TiO(2)), aluminum oxide (Al(2)O(3)), and yttria-stabilized zirconia (YSZ) with density >95%. Bacterial adhesion on the substrates was evaluated using scanning electron microscopy and the spread plate method. Surface properties of the implant discs were scanned using optical microscopy. Human keratocyte attachment and proliferation rates were assessed by cell counting and MTT assay at different time points. Morphologic analysis and immunoblotting were used to evaluate focal adhesion formation, whereas cell adhesion force was measured with a multimode atomic force microscope. The authors found that bacterial adhesion on the TiO(2), Al(2)O(3), and YSZ surfaces were lower than that on HA substrates. TiO(2) significantly promoted keratocyte proliferation and viability compared with HA, Al(2)O(3,) and YSZ. Immunofluorescent imaging analyses, immunoblotting, and atomic force microscope measurement revealed that TiO(2) surfaces enhanced cell spreading and cell adhesion compared with HA and Al(2)O(3). TiO(2) is the most suitable replacement candidate for use as skirt material because it enhanced cell functions and reduced bacterial adhesion. This would, in turn, enhance tissue integration and reduce device failure rates during keratoprosthesis surgery.

  1. Comprehensive analysis of shielding effectiveness for HDPE, BPE and concrete as candidate materials for neutron shielding

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

    In the compact accelerator based DD neutron generator, the deuterium ions generated by the ion source are accelerated after the extraction and bombarded to a deuterated titanium target. The emitted neutrons have typical energy of ∼2.45MeV. Utilization of these compact accelerator based neutron generators of yield up to 10 9 neutron/second (DD) is under active consideration in many research laboratories for conducting active neutron interrogation experiments. Requirement of an adequately shielded laboratory is mandatory for the effective and safe utilization of these generators for intended applications. In this reference, we report the comprehensive analysis of shielding effectiveness for High Density Polyethylene (HDPE), Borated Polyethylene (BPE) and Concrete as candidate materials for neutron shielding. In shielding calculations, neutron induced scattering and absorption gamma dose has also been considered along with neutron dose. Contemporarily any material with higher hydrogenous concentration is best suited for neutron shielding. Choice of shielding material is also dominated by practical issues like economic viability and availability of space. Our computational analysis results reveal that utilization of BPE sheets results in minimum wall thickness requirement for attaining similar range of attenuation in neutron and gamma dose. The added advantage of using borated polyethylene is that it reduces the effect of both neutron and gamma dose by absorbing neutron and producing lithium and alpha particle. It has also been realized that for deciding upon optimum thickness determination of any shielding material, three important factors to be necessarily considered are: use factor, occupancy factor and work load factor. (author)

  2. A Damage Resistance Comparison Between Candidate Polymer Matrix Composite Feedline Materials

    Science.gov (United States)

    Nettles, A. T

    2000-01-01

    As part of NASAs focused technology programs for future reusable launch vehicles, a task is underway to study the feasibility of using the polymer matrix composite feedlines instead of metal ones on propulsion systems. This is desirable to reduce weight and manufacturing costs. The task consists of comparing several prototype composite feedlines made by various methods. These methods are electron-beam curing, standard hand lay-up and autoclave cure, solvent assisted resin transfer molding, and thermoplastic tape laying. One of the critical technology drivers for composite components is resistance to foreign objects damage. This paper presents results of an experimental study of the damage resistance of the candidate materials that the prototype feedlines are manufactured from. The materials examined all have a 5-harness weave of IM7 as the fiber constituent (except for the thermoplastic, which is unidirectional tape laid up in a bidirectional configuration). The resin tested were 977-6, PR 520, SE-SA-1, RS-E3 (e-beam curable), Cycom 823 and PEEK. The results showed that the 977-6 and PEEK were the most damage resistant in all tested cases.

  3. Superconducting Gamma/Neutron Spectrometer Task 1 Completion Report Evaluation of Candidate Neutron-Sensitive Materials

    CERN Document Server

    Bell, Z W

    2002-01-01

    A review of the scientific literature regarding boron- and lithium-containing compounds was completed. Information such as Debye temperature, heat capacity, superconductivity properties, physical and chemical characteristics, commercial availability, and recipes for synthesis was accumulated and evaluated to develop a list of neutron-sensitive materials likely to perform properly in the spectrometer. The best candidate borides appear to be MgB sub 2 (a superconductor with T sub c = 39 K), B sub 6 Si, B sub 4 C, and elemental boron; all are commercially available. Among the lithium compounds are LiH, LiAl, Li sub 1 sub 2 Si sub 7 , and Li sub 7 Sn sub 2. These materials have or are expected to have high Debye temperatures and sufficiently low heat capacities at 100 mK to produce a useful signal. The responses of sup 1 sup 0 B and sup 6 Li to a fission neutron spectrum were also estimated. These demonstrated that the contribution of scattering events is no more than 3% in a boron-based system and 1.5% in a lith...

  4. Performance of candidate gas turbine abradeable seal materials in high temperature combustion atmospheres

    Energy Technology Data Exchange (ETDEWEB)

    Simms, N.J. [Cranfield University, Power Generation Technology Centre, Cranfield, Beds, MK43 0AL (United Kingdom); Norton, J.F. [Cranfield University, Power Generation Technology Centre, Cranfield, Beds, MK43 0AL (United Kingdom); Consultant in Corrosion Science and Technology, Hemel Hempstead, Herts HP1 1SR (United Kingdom); McColvin, G. [Siemens Industrial Turbines Ltd., Lincoln, LN5 7FD (United Kingdom)

    2005-11-01

    The development of abradeable gas turbine seals for higher temperature duties has been the target of an EU-funded R and D project, ADSEALS, with the aim of moving towards seals that can withstand surface temperatures as high as {proportional_to} 1100 C for periods of at least 24,000 h. The ADSEALS project has investigated the manufacturing and performance of a number of alternative materials for the traditional honeycomb seal design and novel alternative designs. This paper reports results from two series of exposure tests carried out to evaluate the oxidation performance of the seal structures in combustion gases and under thermal cycling conditions. These investigations formed one part of the evaluation of seal materials that has been carried out within the ADSEALS project. The first series of three tests, carried out for screening purposes, exposed candidate abradeable seal materials to a simulated natural gas combustion environment at temperatures within the range 1050-1150 C in controlled atmosphere furnaces for periods of up to {proportional_to} 2,500 h with fifteen thermal cycles. The samples were thermally cycled to room temperature on a weekly basis to enable the progress of the degradation to be monitored by mass change and visual observation, as well as allowing samples to be exchanged at planned intervals. The honeycombs were manufactured from PM2000 and Haynes 214. The backing plates for the seal constructions were manufactured from Haynes 214. Some seals contained fillers or had been surface treated (e.g. aluminised). The second series of three tests were carried out in a natural gas fired ribbon furnace facility that allowed up to sixty samples of candidate seal structures (including honeycombs, hollow sphere structures and porous ceramics manufactured from an extended range of materials including Aluchrom YHf, PM2Hf, Haynes 230, IN738LC and MarM247) to be exposed simultaneously to a stream of hot combustion gas. In this case the samples were cooled

  5. Comparison of lithium and the eutectic lead lithium alloy, two candidate liquid metal breeder materials for self-cooled blankets

    International Nuclear Information System (INIS)

    Malang, S.; Mattas, R.

    1994-06-01

    Liquid metals are attractive candidates for both near-term and long-term fusion applications. The subjects of this comparison are the differences between the two candidate liquid metal breeder materials Li and LiPb for use in breeding blankets in the areas of neutronics, magnetohydrodynamics, tritium control, compatibility with structural materials, heat extraction system, safety, and required R ampersand D program. Both candidates appear to be promising for use in self-cooled breeding blankets which have inherent simplicity with the liquid metal serving as both breeders and coolant. The remaining feasibility question for both breeder materials is the electrical insulation between liquid metal and duct walls. Different ceramic coatings are required for the two breeders, and their crucial issues, namely self-healing of insulator cracks and radiation induced electrical degradation are not yet demonstrated. Each liquid metal breeder has advantages and concerns associated with it, and further development is needed to resolve these concerns

  6. Residual stress measurement of the jacket material for ITER coil by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Tsuchiya, Yoshinori [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1998-03-01

    Nickel-Iron based super alloy INCOLOY 908 is used for the jacket of a central solenoid coil (CS coil) of the International Thermonuclear Experimental Reactor (ITER). INCOLOY 908, however, has a possibility of fracture due to Stress Accelerated Grain Boundary Oxidation (SAGBO) under a tensile residual stress beyond 200MPa. Therefore it is necessary to measure the residual stress of the jacket to avoid SAGBO. We performed residual stress measurement of the jacket by neutron diffraction using the neutron diffractometer for residual stress analysis (RESA) installed at JRR-3M in JAERI. A sample depth dependence of internal strain was obtained from the (111) plane spacing. A residual stress distribution was calculated from the strain using Young`s modulus and Poisson`s ratio that were evaluated by a tensile test with neutron diffraction. The result shows that the tensile residual stress exceeds 200MPa of the SAGBO condition in some regions inside the jacket. (author)

  7. Certification of a new biological reference material - Virginia Tobacco Leaves (CTA-VTL-2) and homogeneity study by NAA on this and other candidate reference materials

    International Nuclear Information System (INIS)

    Dybczynski, Rajmund; Polkowska-Motrenko, Halina; Samczynski, Zbigniew; Szopa, Zygmunt; Kulisa, Krzysztof; Wasek, Marek

    2002-01-01

    This report describes the laboratory's participation in the interlaboratory comparison run where the laboratory applied neutron activation analysis aimed at certification of the candidate reference material. Data evaluation and statistical treatment steps are discussed. The report also describes homogeneity study on the reference material and provides details of the analytical procedures

  8. Development of a candidate certified reference material of cypermethrin in green tea

    International Nuclear Information System (INIS)

    Sin, Della W.M.; Chan, Pui-kwan; Cheung, Samuel T.C.; Wong, Yee-Lok; Wong, Siu-kay; Mok, Chuen-shing; Wong Yiuchung

    2012-01-01

    Highlights: ► A cypermethrin CRM in green tea was developed. ► Using two isotope dilution mass spectrometry techniques for characterization. ► Certified value of 148 μg kg −1 with expanded uncertainty of ±9.2%. ► Support quality assurance of pesticide residue analysis in tea to testing. - Abstract: This paper presents the preparation of a candidate certified reference material (CRM) of cypermethrin in green tea, GLHK-11-01a according to the requirements of ISO Guide 34 and 35. Certification of the material was performed using a newly developed isotope dilution mass spectrometry (IDMS) approach, with gas chromatography high resolution mass spectrometry (GC–HRMS) and gas chromatography–tandem mass spectrometry (GC–MS/MS). Statistical analysis (one-way ANOVA) showed excellent agreement of the analytical data sets generated from the two mass spectrometric detections. The characterization methods have also been satisfactorily applied in an Asia-Pacific Metrology Program (APMP) interlaboratory comparison study. Both the GC–HRIDMS and GC–IDMS/MS methods proved to be sufficiently reliable and accurate for certification purpose. The certified value of cypermethrin in dry mass fraction was 148 μg kg −1 and the associated expanded uncertainty was 14 μg kg −1 . The uncertainty budget was evaluated from sample in homogeneity, long-term and short-term stability and variability in the characterization procedure. GLHK-11-01a is primarily developed to support the local and wider testing community on need basis in quality assurance work and in seeking accreditation.

  9. Laboratory corrosion tests on candidate high-level waste container materials: Results from the Belgian programme

    International Nuclear Information System (INIS)

    Druyts, F.; Kursten, B.; Iseghem, P. Van

    2004-01-01

    The Belgian SAFIR-2 concept foresees the geological disposal of conditioned high-level radioactive waste in stainless steel containers and overpacks placed in a concrete gallery backfilled with Boom clay or a bentonite-type backfill. In addition to earlier in situ experiments, we used a laboratory approach to investigate the corrosion properties of selected stainless steels in Boom clay and bentonite environments. In the SAFIR-2 concept, AISI 316L hMo is the main candidate overpack material. As an alternative, we also investigated the higher alloyed stainless steel UHB 904L. Our study focused on localised corrosion and in particular pitting. We used cyclic potentiodynamic polarisation measurements to determine the pit nucleation potential E NP and the protection potential E PP . The evolution of the corrosion potential with time was determined by monitoring the open circuit potential in synthetic clay-water over extended periods. In this paper we present and discuss some results from our laboratory programme, focusing on long-term interactions between the stainless steel overpack and the backfill materials. We describe in particular the influence of chloride and thio-sulphate ions on the pitting corrosion behaviour. The results show that, under geochemical conditions typical for geological disposal, i.e. [Cl-] ∼ 30 mg/L for a Boom clay backfill and [Cl-] ∼ 90 mg/L for a bentonite backfill, neither AISI 316L hMo nor UHB 904L is expected to present pitting problems. An important factor in the long-term prediction of the corrosion behaviour however, is the robustness of the model for the evolution of the geochemistry of the backfill. Indeed, at chloride levels higher than 1000 mg/L, we predict pitting corrosion for AISI 316L hMo. (authors)

  10. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  11. VUV photoemission studies of candidate Large Hadron Collider vacuum chamber materials

    Directory of Open Access Journals (Sweden)

    R. Cimino

    1999-06-01

    Full Text Available In the context of future accelerators and, in particular, the beam vacuum of the Large Hadron Collider (LHC, a 27 km circumference proton collider to be built at CERN, VUV synchrotron radiation (SR has been used to study both qualitatively and quantitatively candidate vacuum chamber materials. Emphasis is given to show that angle and energy resolved photoemission is an extremely powerful tool to address important issues relevant to the LHC, such as the emission of electrons that contributes to the creation of an electron cloud which may cause serious beam instabilities and unmanageable heat loads on the cryogenic system. Here we present not only the measured photoelectron yields from the proposed materials, prepared on an industrial scale, but also the energy and in some cases the angular dependence of the emitted electrons when excited with either a white light (WL spectrum, simulating that in the arcs of the LHC, or monochromatic light in the photon energy range of interest. The effects on the materials examined of WL irradiation and /or ion sputtering, simulating the SR and ion bombardment expected in the LHC, were investigated. The studied samples exhibited significant modifications, in terms of electron emission, when exposed to the WL spectrum from the BESSY Toroidal Grating Monochromator beam line. Moreover, annealing and ion bombardment also induce substantial changes to the surface thereby indicating that such surfaces would not have a constant electron emission during machine operation. Such characteristics may be an important issue to define the surface properties of the LHC vacuum chamber material and are presented in detail for the various samples analyzed. It should be noted that all the measurements presented here were recorded at room temperature, whereas the majority of the LHC vacuum system will be maintained at temperatures below 20 K. The results cannot therefore be directly applied to these sections of the machine until

  12. Tribological Evaluation of Candidate Gear Materials Operating Under Light Loads in Highly Humid Conditions

    Science.gov (United States)

    Dellacorte, Christopher; Thomas, Fransua; Leak, Olivia Ann

    2015-01-01

    A series of pin-on-disk sliding wear tests were undertaken to identify candidate materials for a pair of lightly loaded timing gears operating under highly humid conditions. The target application involves water purification and thus precludes the use of oil, grease and potentially toxic solid lubricants. The baseline sliding pair is austenitic stainless steel operating against a carbon filled polyimide. The test load and sliding speed (4.9 N, 2.7 m/s) were chosen to represent average contact conditions of the meshing gear teeth. In addition to the baseline materials, the hard superelastic NiTiNOL 60 (60NiTi) was slid against itself, against the baseline polyimide, and against 60NiTi onto which a commercially deposited dry film lubricant (DFL) was applied. The alternate materials were evaluated as potential replacements to achieve a longer wear life and improved dimensional stability for the timing gear application. An attempt was also made to provide solid lubrication to self-mated 60NiTi by rubbing the polyimide against the disk wear track outside the primary 60NiTi-60NiTi contact, a method named stick or transfer-film lubrication. The selected test conditions gave repeatable friction and wear data and smooth sliding surfaces for the baseline materials similar to those in the target application. Friction and wear for self-mated stainless steel were high and erratic. Self-mated 60NiTi gave acceptably low friction (approx. 0.2) and modest wear but the sliding surfaces were rough and potentially unsuitable for the gear application. Tests in which 60NiTi pins were slid against DFL coated 60NiTi and DFL coated stainless steel gave low friction and long wear life. The use of stick lubrication via the secondary polyimide pin provided effective transfer film lubrication to self-mated 60NiTi tribological specimens. Using this approach, friction levels were equal or lower than the baseline polyimide-stainless combination and wear was higher but within data scatter observed

  13. Characterization of a backfill candidate material, IBECO-RWC-BF Baclo Project - Phase 3 Laboratory tests

    International Nuclear Information System (INIS)

    Johannesson, Lars-Erik; Sanden, Torbjoern; Dueck, Ann; Ohlsson, Lars

    2010-01-01

    A backfill candidate material, IBECO-RWC-BF, which origin from Milos, Greece, has been investigated. The material was delivered both as granules and as pellets. The investigation described in this report aimed to characterize the material and evaluate if it can be used in a future repository. The following investigations have been done and are presented in this report: 1. Standard laboratory tests. Water content, liquid limit and swelling potential are examples on standard tests that have been performed. 2. Block manufacturing. The block compaction properties of the material have been determined. A first test was performed in laboratory but also tests in large scale have been performed. After finishing the test phase, 60 tons of blocks were manufactured at Hoeganaes Bjuf AB. The blocks will be used in large scale laboratory tests at Aespoe HRL. 3. Mechanical parameters. The compressibility of the material was investigated with oedometer tests (four tests) where the load was applied in steps after saturation. The evaluated oedometer modulus varied between 34.50 MPa. Tests were made to evaluate the elastic parameters of the material (E, ν). Altogether three tests were made on specimens with dry densities of about 1,710 kg/m 3 . The evaluated E-modulus and Poisson's ratio varied between 231-263 MPa and 0.16-0.19 respectively. The strength of the material, both the compressive strength and the tensile strength were measured on specimens compacted to different dry densities. The test results yielded a relation between density and the two types of strength. Furthermore, tests have been made in order to determine the compressibility of the unsaturated filling of pellets. Two tests were made where the pellets were loosely filled in a Proctor cylinder and then compressed at a constant rate of strain during continuously measurement of the applied load. 4. Swelling pressure and hydraulic conductivity. There is, as expected, a very clear influence of the dry density on the

  14. Basic analysis of weldability and machinability of structural materials for ITER Toroidal Field coils

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, Masanori [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan)], E-mail: masanori_onozuka@mnes-us.com; Shimizu, Katsusuke; Urata, Kazuhiro; Kimura, Masahiro; Kadowaki, Hirokazu; Okamoto, Mamoru [Mitsubishi Heavy Industries Ltd., Konan 2-16-5, Minato-ku, Tokyo 108-8215 (Japan); Nakajima, Hideo; Hamada, Kazuya; Okuno, Kiyoshi [Japan Atomic Energy Agency, Mukouyama 801-1, Naka-shi, Ibaraki 311-0193 (Japan)

    2007-10-15

    A manufacturing study has been conducted to examine the welding and machining capabilities for strengthened austenitic stainless steels with a high nitrogen content, JJ1 and ST-SS316LN, to be employed for ITER Toroidal Field (TF) coil structural components. It was found that the applicable EB welding condition for JJ1 was limited to up to 40 mm thick plates. A wider range of welding conditions was found in the vertical upward direction. Based on those results, a verification test up to 900 mm in length was successfully conducted. With respect to TIG welding, an average deposition rate of 26 g/min (i.e. the filler wire supplying speed of 3000 mm/min) was achieved. In addition to the welding tests, a series of machining tests has been conducted to examine the machinability of JJ1 and ST-SS316LN. Various types of machining tools were examined. In practical application, the cutting speed should be low to extend the tool life. At a cutting speed of 40 m/min, a tool life of more than 2 h (at a traveling distance of up to 9 m) was attained. The degree of cutter wear after 30 min of operation, at a cutting speed of 40 m/min, was found to be around 0.1 mm, which is within an acceptable range.

  15. Characterization of a backfill candidate material, IBECO-RWC-BF Baclo Project - Phase 3 Laboratory tests

    Energy Technology Data Exchange (ETDEWEB)

    Johannesson, Lars-Erik; Sanden, Torbjoern; Dueck, Ann; Ohlsson, Lars (Clay Technology AB, Lund (Sweden))

    2010-01-15

    A backfill candidate material, IBECO-RWC-BF, which origin from Milos, Greece, has been investigated. The material was delivered both as granules and as pellets. The investigation described in this report aimed to characterize the material and evaluate if it can be used in a future repository. The following investigations have been done and are presented in this report: 1. Standard laboratory tests. Water content, liquid limit and swelling potential are examples on standard tests that have been performed. 2. Block manufacturing. The block compaction properties of the material have been determined. A first test was performed in laboratory but also tests in large scale have been performed. After finishing the test phase, 60 tons of blocks were manufactured at Hoeganaes Bjuf AB. The blocks will be used in large scale laboratory tests at Aespoe HRL. 3. Mechanical parameters. The compressibility of the material was investigated with oedometer tests (four tests) where the load was applied in steps after saturation. The evaluated oedometer modulus varied between 34.50 MPa. Tests were made to evaluate the elastic parameters of the material (E, nu). Altogether three tests were made on specimens with dry densities of about 1,710 kg/m3. The evaluated E-modulus and Poisson's ratio varied between 231-263 MPa and 0.16-0.19 respectively. The strength of the material, both the compressive strength and the tensile strength were measured on specimens compacted to different dry densities. The test results yielded a relation between density and the two types of strength. Furthermore, tests have been made in order to determine the compressibility of the unsaturated filling of pellets. Two tests were made where the pellets were loosely filled in a Proctor cylinder and then compressed at a constant rate of strain during continuously measurement of the applied load. 4. Swelling pressure and hydraulic conductivity. There is, as expected, a very clear influence of the dry density on the

  16. Preparation of candidate reference materials for the determination of phosphorus containing flame retardants in styrene-based polymers.

    Science.gov (United States)

    Roth, Thomas; Urpi Bertran, Raquel; Latza, Andreas; Andörfer-Lang, Katrin; Hügelschäffer, Claudia; Pöhlein, Manfred; Puchta, Ralph; Placht, Christian; Maid, Harald; Bauer, Walter; van Eldik, Rudi

    2015-04-01

    Candidate reference materials (RM) for the analysis of phosphorus-based flame retardants in styrene-based polymers were prepared using a self-made mini-extruder. Due to legal requirements of the current restriction for the use of certain hazardous substances in electrical and electronic equipment, focus now is placed on phosphorus-based flame retardants instead of the brominated kind. Newly developed analytical methods for the first-mentioned substances also require RMs similar to industrial samples for validation and verification purposes. Hence, the prepared candidate RMs contained resorcinol-bis-(diphenyl phosphate), bisphenol A bis(diphenyl phosphate), triphenyl phosphate and triphenyl phosphine oxide as phosphorus-based flame retardants. Blends of polycarbonate and acrylonitrile-co-butadiene-co-styrene as well as blends of high-impact polystyrene and polyphenylene oxide were chosen as carrier polymers. Homogeneity and thermal stability of the candidate RMs were investigated. Results showed that the candidate RMs were comparable to the available industrial materials. Measurements by ICP/OES, FTIR and NMR confirmed the expected concentrations of the flame retardants and proved that analyte loss and degradation, respectively, was below the uncertainty of measurement during the extrusion process. Thus, the candidate RMs were found to be suitable for laboratory use.

  17. High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Nelson, Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parker, Scott [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Parkison, Adam [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2013-12-23

    The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H2O with the alumina reaction tube at 1700°C resulted in Al(OH)3

  18. Diatomite: A promising natural candidate as carrier material for low, middle and high temperature phase change material

    International Nuclear Information System (INIS)

    Qian, Tingting; Li, Jinhong; Min, Xin; Deng, Yong; Guan, Weimin; Ning, Lei

    2015-01-01

    Graphical abstract: Low-temperature PCMs are always the objects of prime investigations, however, the field of PCMs’ applications is not limited to low temperatures only. In the present study, three kinds of PCMs: polyethylene glycol (PEG), lithium nitrate, and sodium sulfate were respectively employed as the low-, middle- and high-temperature storage medium. A series of novel form-stable phase change materials (fs-PCMs) were tailor-made by blending diatomite and the three kinds of PCMs via a vacuum impregnation method or a facile mixing and sintering method. Various techniques were employed to characterize their structural and thermal properties. - Highlights: • Low-temperature PEG/diatomite was prepared. • Middle-temperature LiNO 3 /diatomite was prepared. • High-temperature Na 2 SO 4 /diatomite was prepared. - Abstract: Low-temperature PCMs are always the objects of prime investigations, however, the field of PCM’s application is not only limited to low temperatures. In this study, polyethylene glycol (PEG), lithium nitrate (LiNO 3 ), and sodium sulfate (Na 2 SO 4 ) were respectively employed as the low-, middle- and high-temperature storage medium. A series of novel form-stable phase change materials (fs-PCMs) were tailor-made by blending diatomite and the three PCMs via a vacuum impregnation method or a facile mixing and sintering method. Various techniques were employed to characterize their structural and thermal properties. The maximum loads of PEG, LiNO 3 , and Na 2 SO 4 in diatomite powder could respectively reach 58%, 60%, and 65%, while PCM melts during the solid–liquid phase transformation. SEM, XRD, and FT-IR results indicated that PCMs were well dispersed into diatomite pores and no chemical changes took place during the heating and cooling process. The prepared fs-PCMs were quite stable in terms of thermal and chemical manner even after a 200-cycle of melting and freezing. The resulting composite fs-PCMs were promising candidates to

  19. Polychromatic Iterative Statistical Material Image Reconstruction for Photon-Counting Computed Tomography

    Directory of Open Access Journals (Sweden)

    Thomas Weidinger

    2016-01-01

    Full Text Available This work proposes a dedicated statistical algorithm to perform a direct reconstruction of material-decomposed images from data acquired with photon-counting detectors (PCDs in computed tomography. It is based on local approximations (surrogates of the negative logarithmic Poisson probability function. Exploiting the convexity of this function allows for parallel updates of all image pixels. Parallel updates can compensate for the rather slow convergence that is intrinsic to statistical algorithms. We investigate the accuracy of the algorithm for ideal photon-counting detectors. Complementarily, we apply the algorithm to simulation data of a realistic PCD with its spectral resolution limited by K-escape, charge sharing, and pulse-pileup. For data from both an ideal and realistic PCD, the proposed algorithm is able to correct beam-hardening artifacts and quantitatively determine the material fractions of the chosen basis materials. Via regularization we were able to achieve a reduction of image noise for the realistic PCD that is up to 90% lower compared to material images form a linear, image-based material decomposition using FBP images. Additionally, we find a dependence of the algorithms convergence speed on the threshold selection within the PCD.

  20. Structural model for the first wall W-based material in ITER project

    Institute of Scientific and Technical Information of China (English)

    Dehua Xu; Xinkui He; Shuiquan Deng; Yong Zhao

    2014-01-01

    The preparation, characterization, and test of the first wall materials designed to be used in the fusion reactor have remained challenging problems in the material science. This work uses the first-principles method as implemented in the CASTEP package to study the influ-ences of the doped titanium carbide on the structural sta-bility of the W–TiC material. The calculated total energy and enthalpy have been used as criteria to judge the structural models built with consideration of symmetry. Our simulation indicates that the doped TiC tends to form its own domain up to the investigated nano-scale, which implies a possible phase separation. This result reveals the intrinsic reason for the composite nature of the W–TiC material and provides an explanation for the experimen-tally observed phase separation at the nano-scale. Our approach also sheds a light on explaining the enhancing effects of doped components on the durability, reliability, corrosion resistance, etc., in many special steels.

  1. Development of geopolymers as candidate materials for low/intermediate level highly alkaline nuclear waste

    International Nuclear Information System (INIS)

    Perera, D.S.; Vance, E.R.; Kiyama, S.; Aly, Z.; Yee, P.

    2006-01-01

    Full text: Geopolymers have been studied for many years as a possible improvement on cement in respect of compressive strength, resistance to fire, heat and acidity, and as a medium for the encapsulation of hazardous or low/ intermediate level radioactive waste. They are made by adding aluminosilicates to concentrated alkali solutions and the application of heat at 0 Cfor subsequent polymerisation. In this work we studied them as suitable candidate materials to incorporate NaOH/NaA10 2 containing waste with low levels of Cs, Sr and Nd. Geopolymers were produced by incorporating the highly alkaline solution as part of the composition with added metakaolinite, fumed silica and extra NaOH, such that the overall geopolymer composition was of molar ratios Si/Al = 2 and Na/Al = 1. The simulated waste contained Na2SO 4 , therefore Ba(OH) 2 was also added to precipitate the SO 4 x 2 as BaSO 4 . Three geoplymers of the same composition containing simulated wastes were leach tested in triplicate after heating at 400 0 Cfor 1 h (to remove -98% of free and interstitial water) under the PCT-B test protocol at 90 0 Cfor 7 days and their results are listed in Table 1. The Cs, Sr and Nd normalised leach rates were low. The Na leach rate was ∼ 4 g/L thus passing the PCT-B test protocol value of 13.5 g/L for EA glass. The X-ray diffraction and scanning electron microscopy showed that BaS04 did precipitate, however all the S did not appear to have precipitated. The ANSI/ANS-16.1-2003 test was carried out on the above geopolymer composition for 5 days. The ANSI Leachability Index D (diffusivity of 10''cm sec'') for the elements released are listed in Table 2. A Portland cement was also tested for comparison and the Leachability index values are 11, 8 and 10 for Al, Na and Ca respectively. Both passed the test protocol insofar as they were > 6. Geopolymers thus passing the tests for high level nuclear waste glass (PCT-B) and for low level nuclear waste (ANSI) show promising potential

  2. Oxidation behavior of austenitic stainless steels as fuel cladding candidate materials for SCWR in superheated steam

    International Nuclear Information System (INIS)

    Abe, Hiroshi; Hong, Seung Mo; Watanabe, Yutaka

    2014-01-01

    Highlights: • Effect of cold work on oxidation kinetics was clearly observed for 15Cr–20Ni SS. • The tube-shaped 15Cr–20Ni SS showed very good oxidation resistance. • The machined layer by cold drawing has a significant role to mitigate oxidation. - Abstract: Oxidation behavior of austenitic stainless steels as fuel cladding candidate materials for supercritical-water-cooled reactor (SCWR), including three types of 15Cr–20Ni stainless steels (1520 SSs), in the temperature range of 700–780 °C superheated steam have been investigated. Effect of temperature, dissolved oxygen (DO), degree of cold work (CW), and machined layer by cold drawing process on the oxidation kinetics assuming power-law kinetics are discussed. Characteristics of oxide layers and its relation to oxidation behaviors are also discussed. The effect of DO on the weight gain behavior in superheated steam at 700 °C was minor for all specimens at least up to 200 ppb DO. The tube-shaped specimens of 1520 SSs showed very good oxidation resistance at 700–780 °C. There was no clear difference in the oxidation kinetics among the three investigated types of 1520 SSs. The machined layer formed at the tube surface has a significant role to mitigate oxidation in superheated steam. A fine-grained microstructure near the surface due to recrystallization by cold drawing process is effective to form the protective Cr 2 O 3 layer. It has been suggested that since Cr diffusion in the outside surface of tubes is accelerated as a result of an increased dislocation density and/or grain refinement by cold drawing, tube specimens show very slow oxidation kinetics. Breakdown of the protective Cr 2 O 3 layer and nodule oxide formation were partly observed on the tube-shaped specimens of 15Cr–20Ni SSs. The reliability of Cr 2 O 3 layer has to be carefully examined to predict the oxidation kinetics after long-term exposure

  3. Oxidation behavior of austenitic stainless steels as fuel cladding candidate materials for SCWR in superheated steam

    Energy Technology Data Exchange (ETDEWEB)

    Abe, Hiroshi, E-mail: hiroshi.abe@qse.tohoku.ac.jp; Hong, Seung Mo; Watanabe, Yutaka

    2014-12-15

    Highlights: • Effect of cold work on oxidation kinetics was clearly observed for 15Cr–20Ni SS. • The tube-shaped 15Cr–20Ni SS showed very good oxidation resistance. • The machined layer by cold drawing has a significant role to mitigate oxidation. - Abstract: Oxidation behavior of austenitic stainless steels as fuel cladding candidate materials for supercritical-water-cooled reactor (SCWR), including three types of 15Cr–20Ni stainless steels (1520 SSs), in the temperature range of 700–780 °C superheated steam have been investigated. Effect of temperature, dissolved oxygen (DO), degree of cold work (CW), and machined layer by cold drawing process on the oxidation kinetics assuming power-law kinetics are discussed. Characteristics of oxide layers and its relation to oxidation behaviors are also discussed. The effect of DO on the weight gain behavior in superheated steam at 700 °C was minor for all specimens at least up to 200 ppb DO. The tube-shaped specimens of 1520 SSs showed very good oxidation resistance at 700–780 °C. There was no clear difference in the oxidation kinetics among the three investigated types of 1520 SSs. The machined layer formed at the tube surface has a significant role to mitigate oxidation in superheated steam. A fine-grained microstructure near the surface due to recrystallization by cold drawing process is effective to form the protective Cr{sub 2}O{sub 3} layer. It has been suggested that since Cr diffusion in the outside surface of tubes is accelerated as a result of an increased dislocation density and/or grain refinement by cold drawing, tube specimens show very slow oxidation kinetics. Breakdown of the protective Cr{sub 2}O{sub 3} layer and nodule oxide formation were partly observed on the tube-shaped specimens of 15Cr–20Ni SSs. The reliability of Cr{sub 2}O{sub 3} layer has to be carefully examined to predict the oxidation kinetics after long-term exposure.

  4. ITER ITA newsletter. No. 8, September 2003

    International Nuclear Information System (INIS)

    2003-10-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities including Robert Aymar's leaving ITER for CERN, ITER related issues at the IAEA General Conference and status and prospects of thermonuclear power and activity during the ITA on materials foe vessel and in-vessel components

  5. Experimental Study of Plasma-Surface Interaction and Material Damage Relevant to ITER Type I Elms

    International Nuclear Information System (INIS)

    Makhlai, V.A.; Bandura, A.N.; Byrka, O.V. and others; Landman, I.; Neklyudov, I.M.

    2006-01-01

    The paper presents experimental investigations of main features of plasma surface interaction and energy transfer to the material surface in dependence on plasma heat loads. The experiments were performed with QSPA repetitive plasma pulses of the duration of 0.25 ms and the energy density up to 2.5 MJ/m 2 . Surface morphology of the targets exposed to QSPA plasma screams is analyzed. Relative contribution of the Lorentz force and plasma pressure gradient to the resulting surface profile is discussed. development of cracking on the tungsten surface and swelling of the surface are found to be in strong dependence on initial temperature of the target

  6. Surface Catalytic Efficiency of Advanced Carbon Carbon Candidate Thermal Protection Materials for SSTO Vehicles

    Science.gov (United States)

    Stewart, David A.

    1996-01-01

    The catalytic efficiency (atom recombination coefficients) for advanced ceramic thermal protection systems was calculated using arc-jet data. Coefficients for both oxygen and nitrogen atom recombination on the surfaces of these systems were obtained to temperatures of 1650 K. Optical and chemical stability of the candidate systems to the high energy hypersonic flow was also demonstrated during these tests.

  7. Plasma-materials interaction issues for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Cohen, S.A.; Werley, K.A.

    1992-02-01

    Analysis of proposed operating scenarios for the International Thermonuclear Experimental Reactor has yielded predictions for the power and particle fluxes onto the material surfaces facing the plasma. The particles, mostly deuterium, tritium, and helium ions, would have energies in the range of 50--2000 eV and fluxes up to 5 x 10 23 /m 2 s. Lower fluxes of multi-MeV electrons and alpha particles may also strike the plasma-facing surfaces, primarily during transient events. The peak power fluxes onto the plasma-facing surfaces during normal operation are expected to be 5--100 MW/m 2 , but much higher during transient events. At the extreme conditions expected for steady-state operation, commonly used heat-removal structures are unable to withstand either the high sputter erosion rates or power loads. To reduce the time-averaged power flux, active control of the plasma position is specified to sweep the plasma heat load across larger areas of plasma-facing components. However, the cyclic heat load creates fatigue lifetime problems. Solutions to these lifetime and reliability problems by (1) changes in machine design and operation, (2) redeposition mechanisms, and (3) changes in materials, will be discussed. A proposed accelerated-life test facility for prototype divertor plate development is described

  8. Investigation of erosion mechanisms and erosion products in divertor armour materials under conditions relevant to elms and mitigated disruptions in ITER

    International Nuclear Information System (INIS)

    Safronov, V.M.; Arkhipov, N.I.; Klimov, N.S.; Kovalenko, D.V.; Moskaleva, A.A.; Podkovyrov, V.L.; Toporkov, D.A.; Zhitlukhin, A.M.; Landman, I.S.; Poznyak, I.M.

    2008-01-01

    Carbon fibre composite (CFC) and tungsten were irradiated by intense plasma streams at plasma gun facilities MK-200UG and QSPA-T. The targets were tested by plasma loads relevant to Edge Localised Modes (ELM) and mitigated disruptions in ITER. Onset condition of material erosion and properties of erosion products have been studied

  9. Carbon fiber composites application in ITER plasma facing components

    Science.gov (United States)

    Barabash, V.; Akiba, M.; Bonal, J. P.; Federici, G.; Matera, R.; Nakamura, K.; Pacher, H. D.; Rödig, M.; Vieider, G.; Wu, C. H.

    1998-10-01

    Carbon Fiber Composites (CFCs) are one of the candidate armour materials for the plasma facing components of the International Thermonuclear Experimental Reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R&D needs are critically discussed.

  10. Carbon fiber composites application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Akiba, M.; Nakamura, K.; Bonal, J.P.; Pacher, H.D.; Roedig, M.; Vieider, G.; Wu, C.H.

    1998-01-01

    Carbon fiber composites (CFCs) are one of the candidate armour materials for the plasma facing components of the international thermonuclear experimental reactor (ITER). For the present reference design, CFC has been selected as armour for the divertor target near the plasma strike point mainly because of unique resistance to high normal and off-normal heat loads. It does not melt under disruptions and might have higher erosion lifetime in comparison with other possible armour materials. Issues related to CFC application in ITER are described in this paper. They include erosion lifetime, tritium codeposition with eroded material and possible methods for the removal of the codeposited layers, neutron irradiation effect, development of joining technologies with heat sink materials, and thermomechanical performance. The status of the development of new advanced CFCs for ITER application is also described. Finally, the remaining R and D needs are critically discussed. (orig.)

  11. Production of candidate natural matrix reference materials for micro-analytical techniques

    International Nuclear Information System (INIS)

    Zeisler, R.; Fajgelj, A.; Zeiller, E.

    2002-01-01

    Homogeneity is considered to be the most vital prerequisite for a certified reference material (CRM); more stringent requirements exist for the analysis of small subsamples. Many of the natural matrix CRMs are prepared from bulk samples by grinding and milling them to a certain particle size, which is expected to provide a more homogenous material; however recommended sample sizes for biological and environmental reference materials are found to be more than 100 mg. Since the milling of materials is costly and has some drawbacks, natural materials that already occur as small particles such as air particulate matter, certain sediments, and cellular biological materials may form the basis of the required reference materials. The nature of these materials, i.e. naturally occurring particles, may provide ideal model reference material. We describe here the production of the materials and preliminary tests, the evaluation for the micro-analytical techniques

  12. Charge, spin and orbital order in the candidate multiferroic material LuFe2O4

    International Nuclear Information System (INIS)

    Groot, Joost de

    2012-01-01

    This thesis is a detailed study of the magnetic, structural and orbital order parameters of the candidate multiferroic material LuFe 2 O 4 . Multiferroic oxides with a strong magnetoelectric coupling are of high interest for potential information technology applications, but they are rare because the traditional mechanism of ferroelectricity is incompatible with magnetism. Consequently, much attention is focused on various unconventional mechanisms of ferroelectricity. Of these, ferroelectricity originating from charge ordering (CO) is particularly intriguing because it potentially combines large electric polarizations with strong magneto-electric coupling. However, examples of oxides where this mechanism occurs are exceedingly rare and none is really well understood. LuFe 2 O 4 is often cited as the prototypical example of CO-based ferroelectricity. In this material, the order of Fe valences has been proposed to render the triangular Fe/O bilayers polar by making one of the two layers rich in Fe 2+ and the other rich in Fe 3+ , allowing for a possible ferroelectric stacking of the individual bilayers. Because of this new mechanism for ferroelectricity, and also because of the high transition temperatures of charge order (T CO ∝320K) and ferro magnetism (T N ∝240 K) LuFe 2 O 4 has recently attracted increasing attention. Although these polar bilayers are generally accepted in the literature for LuFe 2 O 4 , direct proof is lacking. An assumption-free experimental determination of whether or not the CO in the Fe/O bilayers is polar would be crucial, given the dependence of the proposed mechanism of ferroelectricity from CO in LuFe 2 O 4 on polar bilayers. This thesis starts with a detailed characterization of the macroscopic magnetic properties, where growing ferrimagnetic contributions observed in magnetization could be ascribed to increasing oxygen off-stoichiometry. The main focus is on samples exhibiting a sharp magnetic transition to long-range spin order

  13. Biocompatibility and characterisation of a candidate microelectrode material for biosensor applications

    International Nuclear Information System (INIS)

    Cyster, L.A.

    2001-10-01

    Recent advances in microcircuit technology have enabled the fabrication of Multiple Microelectrode Arrays (MEAs) for investigating the characteristics of networks of neuronal cells either in vivo or in vitro. When producing a MEA materials used must be corrosion resistant, have low electrical impedance and the materials must be biocompatible. Existing MEA's have limited life spans, relatively high impedance values and limited uses. Thus creating a requirement for new MEA technology. TiN thin films have become increasingly useful in a wide variety of applications, due to their nature, which includes chemical stability, high hardness, excellent wear and electrical properties and also biocompatibility. The favourable electrical and biocompatibility characteristics of thin films of TiN make them a possible candidate for use in a MEA. TiN thin films can be deposited by a number of methods including evaporation, ion plating and sputtering. The method of deposition, along with process parameters used can have a marked effect on the characteristics of TiN films, including changes in preferred orientation, hardness and wear and also biocompatibility. TiN thin films were deposited onto glass substrates by pulsed DC reactive sputtering of a Ti target, with Argon and nitrogen gas mixtures and labelled Type I TiN films. Also industrial TIN films deposited by Arc Ion plating were carefully selected for comparison and labelled Type II TiN films. The microstructure, composition, surface chemistry, surface topography and roughness were studied using X-Ray diffraction (XRD), X-ray photoelectron spectroscopy (XPS), Atomic Force Microscopy (AFM) and Profilometry. Type I TIN films showed a surface topography similar to Zone I and Type II TiN films showed a surface topography similar to Zone 2 of the Movchan and Dernchishin structure zone model for sputtered films. XPS showed that the surface composition of all TiN films was predominantly TiO 2 , TiN and TiN x O y . Significant

  14. Numerical modeling and experimental simulation of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

    International Nuclear Information System (INIS)

    Wuerz, H.; Arkhipov, N.I.; Bakhin, V.P.; Goel, B.; Hoebel, W.; Konkashbaev, I.; Landman, I.; Piazza, G.; Safronov, V.M.; Sherbakov, A.R.; Toporkov, D.A.; Zhitlukhin, A.M.

    1994-01-01

    The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. In the optical wavelength range C II, C III, C IV emission lines for graphite, Cu I, Cu II lines for copper and continuum radiation for tungsten samples are observed in the target plasma. The plasma expands along the magnetic field lines with velocities of (4±1)x10 6 cm/s for graphite and 10 5 cm/s for copper. Modeling was done with a radiation hydrodynamics code in one-dimensional planar geometry. The multifrequency radiation transport is treated in flux limited diffusion and in forward reverse transport approximation. In these first modeling studies the overall shielding efficiency for carbon and tungsten defined as ratio of the incident energy and the vaporization energy for power densities of 10 MW/cm 2 exceeds a factor of 30. The vapor shield is established within 2 μs, the power fraction to the target after 10 μs is below 3% and reaches in the stationary state after about 20 μs a value of around 1.5%. ((orig.))

  15. Two spruce shoot candidate reference materials from the German environmental specimen bank

    International Nuclear Information System (INIS)

    Backhaus, F.; Bagschik, U.; Burow, M.; Froning, M.; Mohl, C.; Ostapczuk, P.; Rossbach, M.; Schladot, J.D.; Stoeppler, M.; Waidmann, E.; Byrne, A.R.; Zeisler, R.

    1994-01-01

    Two new materials are introduced that might serve as useful aids for the harmonisation of analytical results. Spruce shoots, cryogenically homogenized and characterized for 50 elements from two sampling sites of the German Environmental Specimen Bank (ESB) are presented as possible third generation reference materials that might also act as calibrating materials in speciation analysis. (author)

  16. Evaluation of Landfill Site Candidate for Naturally Occurring Radioactive Materials (Norm) and Hazardous Waste

    International Nuclear Information System (INIS)

    Sucipta; Hadi Suntoko; Bunawas

    2007-01-01

    Refers to co-location concept, Kabil site, where located at the southeast end of low hills in Batam Island, will be sited as an integrated industrial waste management center including landfill. So that, it is necessary an evaluation of the landfill site candidate for NORM and hazardous waste. The evaluation includes geological and non-geological aspects, to determine the suitability or capability in supporting the function as landfill facility. The site candidate was evaluated by serial sreps as follows: 1) criteria formulation; 2) selecting the parameter for evaluation; 3) Positive screening or evaluation of the land having potentiality for landfill site by descriptive method: and 4) determine the land suitability or capability for landfill site. The evaluation of geological and non- geological aspects include topography, litology, seismicity, groundwater and surface water, climate, hydro-oceanography, flora and fauna, spatial pattern and transportation system. The most of the parameters evaluated show the fulfilling to the site criteria, and can be mentioned that the land is suitable for landfill site. Some parameters are not so suitable for that purpose, especially on permeability and homogeneity of the rocks/soils, distance to surface water body, depth of groundwater, the flow rate of groundwater, precipitation, and humidity of the air. The lack of suitability showed by some parameters can be compensated by improving the appropriate engineered barrier in order to fulfill the landfill performance in providing the supporting capacity, long live stability and waste containment. (author)

  17. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...

  18. 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

  19. Wear Test Results of Candidate Materials for the OK-542 Towed Array Handling Machine Level Winder

    Science.gov (United States)

    1994-12-29

    10 6. Wear Testing Photograph B ....................................................... .11 7. Clad Inconel 625 ...interfere with this wear test. Other materials that were tested included Inconel 625 , Titanium, 304 Stainless, 316 Stainless, and Ni-Al-Br. All of these...Stainless Steel, Inconel 625 , Nickel-Aluminum-Bronze, and Titanium. The specialty materials: Inconel 625 , Monel, Stainless and Stellite, were clad-welded

  20. Screening of candidate corrosion resistant materials for coal combustion environments -- Volume 4. Final report, January 31, 1997

    Energy Technology Data Exchange (ETDEWEB)

    Boss, D.E.

    1997-12-31

    The development of a silicon carbide heat exchanger is a critical step in the development of the Externally-Fired Combined Cycle (EFCC) power system. SiC is the only material that provides the necessary combination of resistance to creep, thermal shock, and oxidation. While the SiC structural materials provide the thermomechanical and thermophysical properties needed for an efficient system, the mechanical properties of the SiC tubes are severely degraded through corrosion by the coal combustion products. To obtain the necessary service life of thousands of hours at temperature, a protective coating is needed that is stable with both the SiC tube and the coal combustion products, resists erosion from the particle laden gas stream, is thermal-shock resistant, adheres to SiC during repeated thermal shocks (start-up, process upsets, shut-down), and allows the EFCC system to be cost competitive. The candidate protective materials identified in a previous effort were screened for their stability to the EFCC combustion environment. Bulk samples of each of the eleven candidate materials were prepared, and exposed to coal slag for 100 hours at 1,370 C under flowing air. After exposure the samples were mounted, polished, and examined via x-ray diffraction, energy dispersive spectroscopy, and scanning electron microscopy. In general, the alumina-based materials behaved well, with comparable corrosion depths in all five samples. Magnesium chromite formed a series of reaction products with the slag, which included an alumina-rich region. These reaction products may act as a diffusion barrier to slow further reaction between the magnesium chromite and the slag and prove to be a protective coating. As for the other materials; calcium titanate failed catastrophically, the CS-50 exhibited extension microstructural and compositional changes, and zirconium titanate, barium zironate, and yttrium chromite all showed evidence of dissolution with the slag.

  1. Science-Driven Candidate Search for New Scintillator Materials FY 2013 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Fei; Kerisit, Sebastien N.; Xie, YuLong; Wu, Dangxin; Prange, Micah P.; Van Ginhoven, Renee M.; Campbell, Luke W.; Wang, Zhiguo

    2013-10-01

    This annual report presents work carried out during Fiscal Year (FY) 2013 at Pacific Northwest National Laboratory (PNNL) under the project entitled “Science-Driven Candidate Search for New Scintillator Materials” (Project number: PL13-SciDriScintMat-PD05) and led by Dr. Fei Gao. This project is divided into three tasks, namely (1) Ab initio calculations of electronic properties, electronic response functions and secondary particle spectra; (2) Intrinsic response properties, theoretical light yield, and microscopic description of ionization tracks; and (3) Kinetics and efficiency of scintillation: nonlinearity, intrinsic energy resolution, and pulse shape discrimination. Detailed information on the findings and insights obtained in each of these three tasks are provided in this report. Additionally, papers published this fiscal year or currently in review are included in Appendix together with presentations given this fiscal year.

  2. Thermal characteristics of non-edible oils as phase change materials candidate to application of air conditioning chilled water system

    Science.gov (United States)

    Irsyad, M.; Indartono, Y. S.; Suwono, A.; Pasek, A. D.

    2015-09-01

    The addition of phase change material in the secondary refrigerant has been able to reduce the energy consumption of air conditioning systems in chilled water system. This material has a high thermal density because its energy is stored as latent heat. Based on material melting and freezing point, there are several non-edible oils that can be studied as a phase change material candidate for the application of chilled water systems. Forests and plantations in Indonesia have great potential to produce non-edible oil derived from the seeds of the plant, such as; Calophyllum inophyllum, Jatropha curcas L, and Hevea braziliensis. Based on the melting temperature, these oils can further studied to be used as material mixing in the secondary refrigerant. Thermal characteristics are obtained from the testing of T-history, Differential Scanning Calorimetric (DSC) and thermal conductivity materials. Test results showed an increase in the value of the latent heat when mixed with water with the addition of surfactant. Thermal characteristics of each material of the test results are shown completely in discussion section of this article.

  3. Integrated Corrosion Facility for long-term testing of candidate materials for high-level radioactive waste containment

    International Nuclear Information System (INIS)

    Estill, J.C.; Dalder, E.N.C.; Gdowski, G.E.; McCright, R.D.

    1994-10-01

    A long-term-testing facility, the Integrated Corrosion Facility (I.C.F.), is being developed to investigate the corrosion behavior of candidate construction materials for high-level-radioactive waste packages for the potential repository at Yucca Mountain, Nevada. Corrosion phenomena will be characterized in environments considered possible under various scenarios of water contact with the waste packages. The testing of the materials will be conducted both in the liquid and high humidity vapor phases at 60 and 90 degrees C. Three classes of materials with different degrees of corrosion resistance will be investigated in order to encompass the various design configurations of waste packages. The facility is expected to be in operation for a minimum of five years, and operation could be extended to longer times if warranted. A sufficient number of specimens will be emplaced in the test environments so that some can be removed and characterized periodically. The corrosion phenomena to be characterized are general, localized, galvanic, and stress corrosion cracking. The long-term data obtained from this study will be used in corrosion mechanism modeling, performance assessment, and waste package design. Three classes of materials are under consideration. The corrosion resistant materials are high-nickel alloys and titanium alloys; the corrosion allowance materials are low-alloy and carbon steels; and the intermediate corrosion resistant materials are copper-nickel alloys

  4. Corrosion Behavior of Candidate Materials Used for Urea Hydrolysis Equipment in Coal-Fired Selective Catalytic Reduction Units

    Science.gov (United States)

    Lu, Jintao; Yang, Zhen; Zhang, Bo; Huang, Jinyang; Xu, Hongjie

    2018-05-01

    Corrosion tests were performed in the laboratory in order to assess the corrosion resistance of candidate materials used in urea hydrolysis equipment. The materials to be evaluated were exposed at 145 °C for 1000 h. Alloys 316L, 316L Mod., HR3C, Inconel 718, and TC4 were evaluated. Additionally, aluminide and chromate coatings applied to a 316L substrate were examined. After exposure, the mass changes in the test samples were measured by a discontinuous weighing method, and the morphologies, compositions, and phases of the corrosion products were analyzed using scanning electron microscopy, energy-dispersive spectroscopy, and x-ray diffraction. Results indicated that continuous pitting and dissolution corrosion were the main failure modes for 316L stainless steel. 316L Mod. and HR3C alloy showed better corrosion resistance than 316L due to their relatively high Cr contents, but HR3C exhibited a strong tendency toward intergranular corrosion. Inconel 718, TC4, and aluminide and chromate coating samples showed similar corrosion processes: only depositions formed by hydrothermal reactions were observed. Based on these results, a possible corrosion process in the urea hydrolysis environment was discussed for these candidate materials and questions to be clarified were proposed.

  5. Evaluation of candidate magnetohydrodynamic materials for the U-02 Phase III test

    International Nuclear Information System (INIS)

    Marchant, D.D.; Bates, J.L.

    1978-06-01

    As part of a cooperative U.S.--U.S.S.R. program, electrode and insulator materials tested at the Westinghouse Electrode Systems Test Facility in Pittsburgh, Pennsylvania, were evaluated. From this evaluation materials will be selected for use in the third phase of tests being conducted in the U-02 magnetohydrodynamics test facility in the Soviet Union. Electrode and insulator materials were examined with both an optical microscope and a scanning electron microscope. The cathodes were found to behave differently from the anodes; most notably, the cathodes showed greater potassium interaction. The lanthanum chromite-based electrodes (excluding those fabricated by plasma-spraying) are recommended for testing in the U-02 Phase III test. Hotpressed, fused-grained MgO and sintered MgAl 2 O 4 are recommended as insulator materials. The electrode attachment techniques used in the Westinghouse Tests were inadequate and need to be modified for the U-02 test

  6. Recent results on the neutron irradiation of ITER candidate copper alloys irradiated in DR-3 at 250{degrees}C to 0.3 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, D.J. [Pacific Northwest National Lab., Richland, WA (United States); Singh, B.N.; Toft, P.; Eldrup, M.

    1997-04-01

    Tensile specimens of CuCrZr and CuNiBe alloys were given various heat treatments corresponding to solution anneal, prime-ageing and bonding thermal treatment with additional specimens re-aged and given a reactor bakeout treatment at 350{degrees}C for 100 h. CuAl-25 was also heat treated to simulate the effects of a bonding thermal cycle on the material. A number of heat treated specimens were neutron irradiated at 250{degrees}C to a dose level of {approximately}0.3 dpa in the DR-3 reactor as Riso. The main effect of the bonding thermal cycle heat treatment was a slight decrease in strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post irradiation tests at 250{degrees}C showed a severe loss of ductility in the case of the CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens exhibited a reasonable amount of uniform elongation, with CuCrZr possessing a lower strength.

  7. PROGRAM ASTEC (ADVANCED SOLAR TURBO ELECTRIC CONCEPT). PART 1. CANDIDATE MATERIALS LABORATORY TESTS

    Science.gov (United States)

    A space power system of the type envisioned by the ASTEC program requires the development of a lightweight solar collector of high reflectance...capable of withstanding the space environment for an extended period. A survey of the environment of interest for ASTEC purposes revealed 4 potential...developed by the solar-collector industry for use in the ASTEC program, and to test the effects of space environment on these materials. Of 6 material

  8. Report on Reactor Physics Assessment of Candidate Accident Tolerant Fuel Cladding Materials in LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); George, Nathan [Univ. of Tennessee, Knoxville, TN (United States); Maldonado, G. Ivan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-08-28

    This work focuses on ATF concepts being researched at Oak Ridge National Laboratory (ORNL), expanding on previous studies of using alternate cladding materials in pressurized water reactors (PWRs). The neutronic performance of two leading alternate cladding materials were assessed in boiling water reactors (BWRs): iron-chromium-aluminum (FeCrAl) cladding, and silicon carbide (SiC)-based composite cladding. This report fulfills ORNL Milestone M3FT-15OR0202332 within the fiscal year 2015 (FY15)

  9. Crevice Corrosion Behavior of Candidate Nuclear Waste Container Materials in Repository Environment Paper Number 02529

    International Nuclear Information System (INIS)

    Hua, F.; Sarver, J.; Mohn, W.

    2001-01-01

    Alloy 22 (UNS N06022) and Ti Grade 7 (UNS R52400) have been proposed as the corrosion resistant materials for fabricating the waste package outer barrier and the drip shield, respectively for the proposed nuclear waste repository Yucca Mountain Project. In this work, the susceptibility of welded and annealed Alloy 22 (N06022) and Ti Grade 7 (UNS R52400) to crevice corrosion was studied by the Multiple Crevice Assembly (ASTM G78) method combined with surface morphological observation after four and eight weeks of exposure to the Basic Saturated Water (BSW-12) in a temperature range from 60 to 105 C. The susceptibility of the materials to crevice corrosion was evaluated based on the appearance of crevice attack underneath the crevice formers and the weight loss data. The results showed that, after exposed to BSW-12 for four and eight weeks, no obvious crevice attack was observed on these materials. The descaled weight loss increased with the increase in temperature for all materials. The weight loss, however, is believed to be caused by general corrosion, rather than crevice corrosion. There was no significant difference between the annealed and welded materials either. On the other hand, to conclude that these materials are immune to crevice corrosion in BSW-12 will require longer term testing

  10. SORPTION AND DISPERSION OF STRONTIUM RADIONUCLIDE IN THE BENTONITE-QUARTZ-CLAY AS BACKFILL MATERIAL CANDIDATE ON RADIOACTIVE WASTE REPOSITORY

    Directory of Open Access Journals (Sweden)

    Herry Poernomo

    2010-12-01

    Full Text Available The experiment of sorption and dispersion characteristics of strontium in the mixture of bentonite-quartz, clay-quartz, bentonite-clay-quartz as candidate of raw material for backfill material in the radioactive waste repository has been performed. The objective of this research is to know the grain size effect of bentonite, clay, and quartz on the weight percent ratio of bentonite to quartz, clay to quartz, bentonite to clay to-quartz can be gives physical characteristics of best such as bulk density (rb, effective porosity (e, permeability (K, best sorption characteristic such as distribution coefficient (Kd, and best dispersion characteristics such as dispersivity (a and effective dispersion coefficient (De of strontium in the backfill material candidate. The experiment was carried out in the column filled by the mixture of bentonite-quartz, clay-quartz, bentonite-clay-quartz with the weight percent ratio of bentonite to quartz, clay to quartz, bentonite to clay to quartz of 100/0, 80/20, 60/40, 40/60, 20/80, 0/100 respectively at saturated condition of water, then flowed 0.1 N Sr(NO32 as buffer solution with tracer of 0.05 Ci/cm3 90Sr as strontium radionuclide simulation was leached from immobilized radioactive waste in the radioactive waste repository. The concentration of 90Sr in the effluents represented as Ct were analyzed by Ortec b counter every 30 min, then by using profile concentration of Co and Ct, values of Kd, a and De of 90Sr in the backfill material was determined. The experiment data showed that the best results were -80+120 mesh grain size of bentonite, clay, quartz respectively on the weight percent ratio of bentonite to clay to quartz of 70/10/20 with physical characteristics of rb = 0.658 g/cm3, e = 0.666 cm3/cm3, and K = 1.680x10-2 cm/sec, sorption characteristic of Kd = 46.108 cm3/g, dispersion characteristics of a = 5.443 cm, and De = 1.808x10-03 cm2/sec can be proposed as candidate of raw material of backfill material

  11. Candidate coffee reference material for element content: production and certification schemes adopted at CENA/USP

    Energy Technology Data Exchange (ETDEWEB)

    Tagliaferro, Fabio Sileno; Fernandes, Elisabete A. de Nadai; Bacchi, Marcio Arruda; Franca, Elvis Joacir de [Centro de Energia Nuclear na Agricultura (CENA/USP), Piracicaba, SP (Brazil). Lab. de Radioisotopos], e-mail: fabiotag@cena.usp.br, e-mail: lis@cena.usp.br, e-mail: mabacchi@cena.usp.br, e-mail: ejfranca@cena.usp.br; Bode, Peter; Bacchi, Marcio Arruda; Franca, Elvis Joacir de [Delft University of Technology, Delft (Netherlands). Interfaculty Reactor Inst.], e-mail: P.Bode@iri.tudelft.nl

    2003-07-01

    Certified reference materials (CRMs) play a fundamental role in analytical chemistry establishing the traceability of measurement results and assuring accuracy and reliability. In spite of the huge importance of measurements in the food sector, Brazil does not produce CRMs to supply the demand. Consequently the acquisition of CRMs depends on imports at high costs. The coffee sector needs CRMs, however there is no material that represents the coffee composition. Since 1998, the Laboratorio de Radioisotopos (LRi) of CENA/USP has been involved in analysis of coffee. During this period, knowledge has been accumulated about several aspects of coffee, such as system of cultivation, elemental composition, homogeneity of the material, possible contaminants and physical properties of beans. Concomitantly, LRi has concentrated efforts in the field of metrology in chemistry, and now all this expertise is being used as the basis for the production of a coffee certified reference material (CRM) for inorganic element content. The scheme developed for the preparation and certification of coffee RM relies on the ISO Guides 34 and 35. The approaches for selection, collection and preparation of the material, moisture determination method, homogeneity testing, certification and long-term stability testing are discussed and a time frame for the expected accomplishments is provided. (author)

  12. Spectral emissivity measurements of candidate materials for very high temperature reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cao, G.; Weber, S.J.; Martin, S.O.; Anderson, M.H. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Sridharan, K., E-mail: kumars@cae.wisc.edu [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin, 1500 Engineering Drive, Madison, WI (United States)

    2012-10-15

    Heat dissipation by radiation is an important consideration in VHTR components, particularly the reactor pressure vessel (RPV), because of the fourth power temperature dependence of radiated heat. Since emissivity is the material property that dictates the ability to radiate heat, measurements of emissivities of materials that are being specifically considered for the construction of VHTR become important. Emissivity is a surface phenomenon and therefore compositional, structural, and topographical changes that occur at the surfaces of these materials as a result of their interactions with the environment at high temperatures will alter their emissivities. With this background, an experimental system for the measurement of spectral emissivity has been designed and constructed. The system has been calibrated in conformance with U.S. DoE quality assurance standards using inert ceramic materials, boron nitride, silicon carbide, and aluminum oxide. The results of high temperature emissivity measurements of potential VHTR materials such as ferritic steels SA 508, T22, T91 and austenitic alloys IN 800H, Haynes 230, IN 617, and 316 stainless steel have been presented.

  13. Candidate coffee reference material for element content: production and certification schemes adopted at CENA/USP

    International Nuclear Information System (INIS)

    Tagliaferro, Fabio Sileno; Fernandes, Elisabete A. de Nadai; Bacchi, Marcio Arruda; Franca, Elvis Joacir de; Bode, Peter; Bacchi, Marcio Arruda; Franca, Elvis Joacir de

    2003-01-01

    Certified reference materials (CRMs) play a fundamental role in analytical chemistry establishing the traceability of measurement results and assuring accuracy and reliability. In spite of the huge importance of measurements in the food sector, Brazil does not produce CRMs to supply the demand. Consequently the acquisition of CRMs depends on imports at high costs. The coffee sector needs CRMs, however there is no material that represents the coffee composition. Since 1998, the Laboratorio de Radioisotopos (LRi) of CENA/USP has been involved in analysis of coffee. During this period, knowledge has been accumulated about several aspects of coffee, such as system of cultivation, elemental composition, homogeneity of the material, possible contaminants and physical properties of beans. Concomitantly, LRi has concentrated efforts in the field of metrology in chemistry, and now all this expertise is being used as the basis for the production of a coffee certified reference material (CRM) for inorganic element content. The scheme developed for the preparation and certification of coffee RM relies on the ISO Guides 34 and 35. The approaches for selection, collection and preparation of the material, moisture determination method, homogeneity testing, certification and long-term stability testing are discussed and a time frame for the expected accomplishments is provided. (author)

  14. Science-Driven Candidate Search for New Scintillator Materials: FY 2014 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Kerisit, Sebastien N.; Gao, Fei; Xie, YuLong; Campbell, Luke W.; Wu, Dangxin; Prange, Micah P.

    2014-10-01

    This annual reports presents work carried out during Fiscal Year (FY) 2014 at Pacific Northwest National Laboratory (PNNL) under the project entitled “Science-Driven Candidate Search for New Scintillator Materials” (Project number: PL13-SciDriScintMat-PD05) and led by Drs. Fei Gao and Sebastien N. Kerisit. This project is divided into three tasks: 1) Ab initio calculations of electronic properties, electronic response functions and secondary particle spectra; 2) Intrinsic response properties, theoretical light yield, and microscopic description of ionization tracks; and 3) Kinetics and efficiency of scintillation: nonproportionality, intrinsic energy resolution, and pulse shape discrimination. Detailed information on the results obtained in each of the three tasks is provided in this Annual Report. Furthermore, peer-reviewed articles published this FY or currently under review and presentations given this FY are included in Appendix. This work was supported by the National Nuclear Security Administration, Office of Nuclear Nonproliferation Research and Development (DNN R&D/NA-22), of the U.S. Department of Energy (DOE).

  15. Cobalt and cerium coated Ni powder as a new candidate cathode material for MCFC

    International Nuclear Information System (INIS)

    Kim, Min Hyuk; Hong, Ming Zi; Kim, Young-Suk; Park, Eunjoo; Lee, Hyunsuk; Ha, Hyung-Wook; Kim, Keon

    2006-01-01

    The dissolution of nickel oxide cathode in the electrolyte is one of the major technical obstacles to the commercialization of molten carbonate fuel cell (MCFC). To improve the MCFC cathode stability, the alternative cathode material for MCFC was prepared, which was made of Co/Ce-coated on the surface of Ni powder using a polymeric precursor based on the Pechini method. X-ray diffraction (XRD) and scanning electron microscopy (SEM) with energy dispersive X-ray analysis (EDAX) were employed in characterization of the alternative cathode materials. The Co/Ce-coated Ni cathode prepared by the tape-casting technique. The solubility of the Co/Ce-coated Ni cathode was about 80% lower when compare to that of pure Ni cathode under CO 2 :O 2 (66.7:33.3%) atmosphere at 650 deg. C. Consequently, the fine Co/Ce-coated Ni powder could be confirmed as a new alternative cathode material for MCFC

  16. Quality assessment of organic coffee beans for the preparation of a candidate reference material

    International Nuclear Information System (INIS)

    Tagliaferro, F.S.; Nadai Fernandes de, E.A.; Bacchi, M.A.

    2006-01-01

    A random sampling was carried out in the coffee beans collected for the preparation of the organic green coffee reference material in view of assessing the homogeneity and the presence of soil as impurity. Fifteen samples were taken for the between-sample homogeneity evaluation. One of the samples was selected and 10 test portions withdrawn for the within-sample homogeneity evaluation. Br, Ca, Co, Cs, Fe, K, Na, Rb, Sc and Zn were determined by instrumental neutron activation analysis (INAA). The F-test demonstrated that the material is homogeneous for Ca, Co, Cs, K and Sc, but not homogeneous for Br, Fe, Na, Rb and Zn. Results of terrigenous elements suggested negligible soil contamination in the raw material. (author)

  17. Testing the homogeneity of candidate reference materials by solid sampling - AAS and INAA

    International Nuclear Information System (INIS)

    Rossbach, M.; Grobecker, K.-H.

    2002-01-01

    The necessity to quantify a natural material's homogeneity with respect to its elemental distribution prior to chemical analysis of a given aliquot is emphasised. Available instruments and methods to obtain the relevant information are described. Additionally the calculation of element specific, relative homogeneity factors, H E , and of a minimum sample mass M 5% to achieve 5% precision on a 95% confidence level is given. Especially, in the production and certification of Certified Reference Materials (CRMs) this characteristic information should be determined in order to provide the user with additional inherent properties of the CRM to enable more economical use of the expensive material and to evaluate further systematic bias of the applied analytical technique. (author)

  18. Evolution and characterization of eggshell as a potential candidate of raw material

    Directory of Open Access Journals (Sweden)

    T. Zaman

    Full Text Available Abstract Characterization of both uncalcined and calcined eggshells was done in this work. Raw eggshells turned out as a good source of calcite phase. Calcined eggshells had a mixture of lime and portlandite phase. A significant impact of calcination temperature on the percentage of generated phases was observed. Qualitative as well as semi-quantitative phase analysis, morphological characterization and physical property estimation was done for the produced powder. The influence of synthesized raw material on soil stabilization and biomaterial formation was further assessed. The eggshell turned out as a potential source of raw material for various sectors.

  19. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  20. Application of miniaturized disk bend test technique for selection of optimum composition of candidate materials for fusion reactors

    International Nuclear Information System (INIS)

    Tsepelev, A.B.; Poymenov, I.L.

    1992-01-01

    An analysis of the potential of a miniaturized disk bend test (MDBT) technique for estimation of irradiated steel mechanical properties behaviour indicates promise in selecting candidate materials for nuclear applications. The advantages of the method are most clearly demonstrated when a large series of tests is needed. The tiny specimen size gives an additional advantage from the point of view of radiation material science. As an example of the MDBT potential, preliminary results of electron irradiation effects on Cr-Mn-W austenitic and Cr-W ferrite carbon and nitrogen steels are presented. It is shown that electron irradiation causes changes of the loading MDBT-curve form of the steels that most probably are connected with radiation-induced structure-phase transformations in the steels. (orig.)

  1. Determination of cadmium, lead and zinc in a candidate reference materials using isotope dilution mass spectrometry

    International Nuclear Information System (INIS)

    Munoz, Luis; Gras, Nuri; Quejido, Alberto; Fernandez, Marta

    2001-01-01

    The growing demands placed on analytical laboratories to ensure the reliability of their results, due to the introduction of systems of quality and to the increasing use of metrology in chemical measurements has led most laboratories to validate their methodologies and to control them statistically. One of the techniques used most often for these purposes is based on the use of reference materials. The proper use of these materials means that laboratory results may be traced to the International System of Units, analytical methodologies can be validated, instruments calibrated and chemical measurements harmonized. One of the biggest challenges in developing reference materials is that of certifying their properties, a process that has been defined as assigning a concentration value that is as close as possible to the true value together with its uncertainty. Organizations that produce reference materials use several options for their certification process, and among these is the use of a primary method. Among the primary methods recognized by the International Office of Weights and Measures is the Isotope Dilution Mass Spectrometry technique. The Chilean Nuclear Energy Commission, through its Reference Materials Program, has prepared a reference material of clam tissue, which has been chemically defined by different analytical methodologies applied in different national and international laboratories. This work describes the methodology developed with the CIEMAT for determining the elements lead, cadmium and zinc in the clam tissue reference material using the primary technique of Isotope Dilution Mass Spectrometry. The calculation is described for obtaining the spike amounts to be added to the sample and the procedure is explained for carrying out the isotopic exchange. The isotopic relationships 204 Pb/ 205 Pb, 111 Cd/ 114 Cd and 66 Zn/ 67 Zn were determined in an atomic emission spectrometer with a plasma source with the following characteristics: plasma

  2. Characterization of a new candidate isotopic reference material for natural Pb using primary measurement method.

    Science.gov (United States)

    Nonose, Naoko; Suzuki, Toshihiro; Shin, Ki-Cheol; Miura, Tsutomu; Hioki, Akiharu

    2017-06-29

    A lead isotopic standard solution with natural abundance has been developed by applying a mixture of a solution of enriched 208 Pb and a solution of enriched 204 Pb ( 208 Pb- 204 Pb double spike solution) as bracketing method. The amount-of-substance ratio of 208 Pb: 204 Pb in this solution is accurately measured by applying EDTA titrimetry, which is one of the primary measurement methods, to each enriched Pb isotope solution. Also metal impurities affecting EDTA titration and minor lead isotopes contained in each enriched Pb isotope solution are quantified by ICP-SF-MS. The amount-of-substance ratio of 208 Pb: 204 Pb in the 208 Pb- 204 Pb double spike solution is 0.961959 ± 0.000056 (combined standard uncertainty; k = 1). Both the measurement of lead isotope ratios in a candidate isotopic standard solution and the correction of mass discrimination in MC-ICP-MS are carried out by coupling of a bracketing method with the 208 Pb- 204 Pb double spike solution and a thallium internal addition method, where thallium solution is added to the standard and the sample. The measured lead isotope ratios and their expanded uncertainties (k = 2) in the candidate isotopic standard solution are 18.0900 ± 0.0046 for 206 Pb: 204 Pb, 15.6278 ± 0.0036 for 207 Pb: 204 Pb, 38.0626 ± 0.0089 for 208 Pb: 204 Pb, 2.104406 ± 0.00013 for 208 Pb: 206 Pb, and 0.863888 ± 0.000036 for 207 Pb: 206 Pb. The expanded uncertainties are about one half of the stated uncertainty for NIST SRM 981, for 208 Pb: 204 Pb, 207 Pb: 204 Pb and 206 Pb: 204 Pb, or one eighth, for 208 Pb: 206 Pb and 207 Pb: 206 Pb, The combined uncertainty consists of the uncertainties due to lead isotope ratio measurements and the remaining time-drift effect of mass discrimination in MC-ICP-MS, which is not removed by the coupled correction method. In the measurement of 208 Pb: 204 Pb, 207 Pb: 204 Pb and 206 Pb: 204 Pb, the latter contribution is two or three times larger than the former. When the coupling of

  3. Nb-base FS-85 Alloy as a Candidate Structural Material for Space Reactor Applications: Effects of Thermal Aging

    International Nuclear Information System (INIS)

    Leonard, Keith J.; Busby, Jeremy T.; Hoelzer, David T.; Zinkle, Steven J.

    2009-01-01

    The proposed use of fission reactors for manned or deep space missions have typically relied on the potential use of refractory metal alloys as structural materials. Throughout the history of these programs, the lead candidate has been Nb-1Zr due to its good fabrication and welding characteristics. However, the less than optimal creep resistance of this alloy has encouraged interest in the more complex FS-85 (Nb-28Ta-10W-1Zr) alloy. Despite this interest, a relatively small database exists for the properties of FS-85. These gaps include potential microstructural instabilities that can lead to mechanical property degradation. In this work, changes in microstructure and mechanical properties of FS-85 were investigated following 1100 h of thermal aging at 1098, 1248 and 1398 K. The changes in electrical resistivity, hardness and tensile properties between the as-annealed and aged materials are compared. Evaluation of the microstructural changes was performed through optical, scanning and transmission electron microscopy. The development of intragranular and grain boundary precipitation of Zr-rich compounds as a function of aging temperature was followed. Brittle tensile behavior was measured in the 1248 K aged material, while ductile behavior occurred in material aged above and below this temperature. The effect of temperature on the under and overaging of the grain boundary particles are believed to have contributed to the mechanical property behavior of the aged material

  4. Engineered materials characterization report for the Yucca Mountain Site Characterization Project. Volume 1, Introduction, history, and current candidates

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; McCright, R.D.; Roy, A.K.; Jones, D.A.

    1995-08-01

    The purpose of the Yucca Mountain Site Characterization Project is to evaluate Yucca Mountain for its suitability as a potential site for the nation's first high-level nuclear waste repository. As part of this effort, Lawrence Livermore National Laboratory (LLNL) has been occupied for a number of years with developing and evaluating the performance of waste packages for the potential repository. In recent years this work has been carried out under the guidance of and in collaboration with the Management and Operating contractor for the Civilian Radioactive Waste Management System, TRW Environmental Safety Systems, Inc., which in turn reports to the Office of Civilian Radioactive Waste Management of the US Department of Energy. This report summarizes the history of the selection and characterization of materials to be used in the engineered barrier system for the potential repository at Yucca Mountain, describes the current candidate materials, presents a compilation of their properties, and summarizes available corrosion data and modeling. The term ''engineered materials'' is intended to distinguish those materials that are used as part of the engineered barrier system from the natural, geologic materials of the site

  5. ITER EDA newsletter. V. 7, no. 1

    International Nuclear Information System (INIS)

    1998-01-01

    This issue of the ITER Newsletter contains a summary report on the Thirteenth meeting of the ITER Management Advisory Committee (MAC), a report on ITER at the International Conference on Fusion Reactor Materials and a report of a Russian scientist working at ITER Garching JWS

  6. Glass: a candidate engineered material for management of high level nuclear waste

    International Nuclear Information System (INIS)

    Mishra, R.K.; Kaushik, C.P.

    2011-01-01

    While the commercial importance of glass is generally recognized, a few people are aware of extremely wide range of glass formulations that can be made and of the versatility of this engineered material. Some of the recent developments in the field of glass leading to various technological applications include glass fiber reinforcement of cement to give new building materials, substrates for microelectronics circuitry in form of semiconducting glasses, nuclear waste immobilization and specific medical applications. The present paper covers fundamental understanding of glass structure and its application for immobilization of high level radioactive liquid waste. High level radioactive liquid waste (HLW) arising during reprocessing of spent fuel are immobilized in sodium borosilicate glass matrix developed indigenously. Glass compositions are modified according to the composition of HLW to meet the criteria of desirable properties in terms. These glass matrices have been characterized for different properties like homogeneity, chemical durability, thermal stability and radiation stability. (author)

  7. Interaction of candidate plasma facing materials with tokamak plasma in COMPASS

    Czech Academy of Sciences Publication Activity Database

    Matějíček, Jiří; Weinzettl, Vladimír; Macková, Anna; Malinský, Petr; Havránek, Vladimír; Naydenkova, Diana; Klevarová, Veronika; Petersson, P.; Gasior, P.; Hakola, A.; Rubel, M.; Fortuna, E.; Kolehmainen, J.; Tervakangas, S.

    2017-01-01

    Roč. 493, September (2017), s. 102-119 ISSN 0022-3115. [International Conference on Plasma-Facing Materials and Components for Fusion Applications/15./. Aix-en-Provence, 18.05.2015-22.05.2015] R&D Projects: GA ČR(CZ) GA14-12837S; GA ČR(CZ) GA15-10723S; GA MŠk(CZ) LM2015045; GA MŠk LM2015056 Institutional support: RVO:61389021 ; RVO:61389005 Keywords : erosion * COMPASS tokamak * plasma-material interaction * ion beam analysis Subject RIV: JF - Nuclear Energetics; JF - Nuclear Energetics (UJF-V) OBOR OECD: Nuclear related engineering ; Nuclear related engineering (UJF-V) Impact factor: 2.048, year: 2016 http://www.sciencedirect.com/science/ article /pii/S0022311517301708

  8. Homogeneity study on biological candidate reference materials: the role of neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Daniel P.; Moreira, Edson G., E-mail: dsilva.pereira@usp.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2015-07-01

    Instrumental Neutron activation Analysis (INAA) is a mature nuclear analytical technique able to accurately determine chemical elements without the need of sample digestion and, hence, without the associated problems of analyte loss or contamination. This feature, along with its potentiality use as a primary method of analysis, makes it an important tool for the characterization of new references materials and in the assessment of their homogeneity status. In this study, the ability of the comparative method of INAA for the within-bottle homogeneity of K, Mg, Mn and V in a mussel reference material was investigated. Method parameters, such as irradiation time, sample decay time and distance from sample to the detector were varied in order to allow element determination in subsamples of different sample masses in duplicate. Sample masses were in the range of 1 to 250 mg and the limitations of the detection limit for small sample masses and dead time distortions for large sample masses were investigated. (author)

  9. Assessment of vanadium alloys for ITER application

    International Nuclear Information System (INIS)

    Borgstedt, H.U.; Clemens, H.; Ehrlich, K.; Fromm, E.; Kelzenberg, S.; Moeslang, A.; Pick, M.; Ruehle, M.; Schaaf, B. van der; Schaefer, L.; Schiller, P.; Schirra, M.; Witwer, M.; Witzenburg, W. van; Zolti, E.; Zucchetti, M.

    1993-09-01

    The assessment effort concerned required evaluation of various relevant properties of vanadium alloys. The outcome predictably shows that these properties, as well as timing, funding, manufacturing and licensing aspects, each set their own specific boundary conditions for application of these alloys in ITER. Some of these are not really felt as constraints. Their capacity to accommodate high heat loads, for example, is better than other candidate materials and appears to be the main reason for the present interest in these alloys. Other favourable properties include neutronic properties (low nuclear heating rates, good tritium breeding performance and low helium generation rates), intrinsically low activation, excellent tensile and creep properties up to high temperatures and high strength-to-density ratio. Not all of these properties necessarily are relevant for ITER, but they would be important for longer term application. (orig.)

  10. Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

    Energy Technology Data Exchange (ETDEWEB)

    Seni, C [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Mills, R H [Toronto Univ., ON (Canada)

    1994-12-31

    Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs.

  11. Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

    International Nuclear Information System (INIS)

    Seni, C.; Mills, R.H.

    1994-01-01

    Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs

  12. Polysaccharide Fabrication Platforms and Biocompatibility Assessment as Candidate Wound Dressing Materials

    Directory of Open Access Journals (Sweden)

    Donald C. Aduba

    2017-01-01

    Full Text Available Wound dressings are critical for wound care because they provide a physical barrier between the injury site and outside environment, preventing further damage or infection. Wound dressings also manage and even encourage the wound healing process for proper recovery. Polysaccharide biopolymers are slowly becoming popular as modern wound dressings materials because they are naturally derived, highly abundant, inexpensive, absorbent, non-toxic and non-immunogenic. Polysaccharide biopolymers have also been processed into biomimetic platforms that offer a bioactive component in wound dressings that aid the healing process. This review primarily focuses on the fabrication and biocompatibility assessment of polysaccharide materials. Specifically, fabrication platforms such as electrospun fibers and hydrogels, their fabrication considerations and popular polysaccharides such as chitosan, alginate, and hyaluronic acid among emerging options such as arabinoxylan are discussed. A survey of biocompatibility and bioactive molecule release studies, leveraging polysaccharide’s naturally derived properties, is highlighted in the text, while challenges and future directions for wound dressing development using emerging fabrication techniques such as 3D bioprinting are outlined in the conclusion. This paper aims to encourage further investigation and open up new, disruptive avenues for polysaccharides in wound dressing material development.

  13. Creep resistance and material degradation of a candidate Ni–Mo–Cr corrosion resistant alloy

    Energy Technology Data Exchange (ETDEWEB)

    Shrestha, Sachin L., E-mail: sachin@ansto.gov.au [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia); Bhattacharyya, Dhriti [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia); Yuan, Guangzhou; Li, Zhijun J. [Center of Thorium Molten Salts Reactor System, Shanghai Institute of Applied Physics, Chinese Academy of Sciences (China); Budzakoska-Testone, Elizabeth; De Los Reyes, Massey; Drew, Michael; Edwards, Lyndon [Institute of Materials Engineering, Australian Nuclear Science and Technology Organisation (ANSTO), Locked Bag 2001, Kirrawee DC, NSW 2232 (Australia)

    2016-09-30

    This study investigated the creep deformation properties of GH3535, a Ni–Mo–Cr corrosion resistant structural alloy being considered for use in future Gen IV molten salt nuclear reactors (MSR) operating at around 700 °C. Creep testing of the alloy was conducted at 650–750 °C under applied stresses between 85–380 MPa. From the creep rupture results the long term creep strain and rupture life of the alloy were estimated by applying the Dorn Shepard and Larson Miller time-temperature parameters and the alloy's allowable ASME design stresses at the MSR's operating temperature were evaluated. The material's microstructural degradation at creep rupture was characterised using scanning electron microscopy (SEM), electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). The microstructural study revealed that the material failure was due to wedge cracking at triple grain boundary points and cavitation at coarse secondary grain boundary precipitates, nucleated and grown during high temperature exposure, leading to intergranular crack propagation. EBSD local misorientation maps clearly show that the root cause of cavitation and crack propagation was due to large strain localisation at the grain boundaries and triple points instigated by grain boundary sliding during creep deformation. This caused the grain boundary decohesion and subsequent material failure.

  14. Challenges in the size analysis of a silica nanoparticle mixture as candidate certified reference material

    International Nuclear Information System (INIS)

    Kestens, Vikram; Roebben, Gert; Herrmann, Jan; Jämting, Åsa; Coleman, Victoria; Minelli, Caterina; Clifford, Charles; Temmerman, Pieter-Jan De; Mast, Jan; Junjie, Liu; Babick, Frank; Cölfen, Helmut; Emons, Hendrik

    2016-01-01

    A new certified reference material for quality control of nanoparticle size analysis methods has been developed and produced by the Institute for Reference Materials and Measurements of the European Commission’s Joint Research Centre. The material, ERM-FD102, consists of an aqueous suspension of a mixture of silica nanoparticle populations of distinct particle size and origin. The characterisation relied on an interlaboratory comparison study in which 30 laboratories of demonstrated competence participated with a variety of techniques for particle size analysis. After scrutinising the received datasets, certified and indicative values for different method-defined equivalent diameters that are specific for dynamic light scattering (DLS), centrifugal liquid sedimentation (CLS), scanning and transmission electron microscopy (SEM and TEM), atomic force microscopy (AFM), particle tracking analysis (PTA) and asymmetrical-flow field-flow fractionation (AF4) were assigned. The value assignment was a particular challenge because metrological concepts were not always interpreted uniformly across all participating laboratories. This paper presents the main elements and results of the ERM-FD102 characterisation study and discusses in particular the key issues of measurand definition and the estimation of measurement uncertainty.

  15. Tritium retention in candidate next-step protection materials: engineering key issues and research requirements

    International Nuclear Information System (INIS)

    Federici, G.; Andrew, P.L.; Wu, C.H.

    1995-01-01

    Although a considerable volume of valuable data on the behaviour of tritium in beryllium and carbon-based armours exposed to hydrogenic fusion plasmas has been compiled over the past years both from operation of present-day tokamaks and from laboratory simulations, knowledge is far from complete and tritium inventory predictions for these materials remain highly uncertain. In this paper we elucidate the main mechanisms responsible for tritium trapping and release in next-step D-T tokamaks, as well as the applicability of some of the presently known data bases for design purposes. Owing to their strong anticipated implications on tritium uptake and release, attention is focused mainly on the interaction of tritium with neutron damage induced defects, on tritium codeposition with eroded carbon and on the effects of oxide and surface contaminants. Some preliminary quantitative estimates are presented based on most recent experimental findings and latest modelling developments as well. The influence of important working conditions such as target temperature, loading particle fluxes, erosion and redeposition rates, as well as material characteristics such as the type of morphology of the protection material (i.e. amorphous plasma-sprayed beryllium vs. solid forms), and design dependent parameters are discussed in this paper. Remaining issues which require additional effort are identified. (orig.)

  16. Challenges in the size analysis of a silica nanoparticle mixture as candidate certified reference material

    Energy Technology Data Exchange (ETDEWEB)

    Kestens, Vikram, E-mail: vikram.kestens@ec.europa.eu; Roebben, Gert [Joint Research Centre (JRC), European Commission, Institute for Reference Materials and Measurements (IRMM) (Belgium); Herrmann, Jan; Jämting, Åsa; Coleman, Victoria [National Measurement Institute Australia, Nanometrology Section (Australia); Minelli, Caterina; Clifford, Charles [National Physical Laboratory, Analytical Science Division (United Kingdom); Temmerman, Pieter-Jan De; Mast, Jan [Service Electron Microscopy, Veterinary and Agrochemical Research Centre (CODA-CERVA) (Belgium); Junjie, Liu [National Institute of Metrology, Division of Nanoscale Measurement and Advanced Materials (China); Babick, Frank [Technische Universität Dresden, Institut für Verfahrens- und Umwelttechnik (Germany); Cölfen, Helmut [University of Konstanz, Physical Chemistry, Department of Chemistry (Germany); Emons, Hendrik [Joint Research Centre (JRC), European Commission, Institute for Reference Materials and Measurements (IRMM) (Belgium)

    2016-06-15

    A new certified reference material for quality control of nanoparticle size analysis methods has been developed and produced by the Institute for Reference Materials and Measurements of the European Commission’s Joint Research Centre. The material, ERM-FD102, consists of an aqueous suspension of a mixture of silica nanoparticle populations of distinct particle size and origin. The characterisation relied on an interlaboratory comparison study in which 30 laboratories of demonstrated competence participated with a variety of techniques for particle size analysis. After scrutinising the received datasets, certified and indicative values for different method-defined equivalent diameters that are specific for dynamic light scattering (DLS), centrifugal liquid sedimentation (CLS), scanning and transmission electron microscopy (SEM and TEM), atomic force microscopy (AFM), particle tracking analysis (PTA) and asymmetrical-flow field-flow fractionation (AF4) were assigned. The value assignment was a particular challenge because metrological concepts were not always interpreted uniformly across all participating laboratories. This paper presents the main elements and results of the ERM-FD102 characterisation study and discusses in particular the key issues of measurand definition and the estimation of measurement uncertainty.

  17. Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

    International Nuclear Information System (INIS)

    Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Gdowski, G.E.

    1988-05-01

    Three copper-based alloys --- CDA 102 (OFHC copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni) --- are being considered as possible materials for the fabrication of high-level radioactive-waste disposal containers. Waste will include fuel assemblies from reactors as well as borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada, for emplacement. The three copper-based alloys discussed here are being considered in addition to the iron- to nickel-based austenitic materials discussed in Volume 3. The decay of radionuclides will result in substantial heat generation and in fluxes of gamma radiation. In this environment, container materials may degrade by atmospheric oxidation, uniform aqueous phase corrosion, pitting, crevice corrosion, transgranular stress corrosion cracking (TGSCC) in tarnishing environments, or intergranular stress corrosion cracking (IGSCC) in nontarnishing environments. This report is a critical survey of available data on the stress corrosion cracking (SCC) of the three copper-based alloys. The requisite conditions for TGSCC and IGSCC include combinations of stress, oxygen, ammonia or nitrite, and water. Note that nitrite is generated by gamma radiolysis of moisture films in air but that ammonia is not. TGSCC has been observed in CDA 102 and CDA 613 exposed to moist ammonia-containing environments whereas SCC has not been documented for CDA 715 under similar conditions. SCC is also promoted in copper by nitrite ions. Furthermore, phosphorus-deoxidized copper is unusually susceptible to embrittlement in such environments. The presence of tin in CDA 613 prevents IGSCC. It is believed that tin segregates to grain boundaries, where it oxidizes very slowly, thereby inhibiting the oxidation of aluminum. 117 refs., 27 figs., 9 tabs

  18. Optimization on electrochemical synthesis of HKUST-1 as candidate catalytic material for Green diesel production

    Science.gov (United States)

    Lestari, W. W.; Nugraha, R. E.; Winarni, I. D.; Adreane, M.; Rahmawati, F.

    2016-04-01

    In the effort to support the discovery of new renewable energy sources in Indonesia, biofuel is one of promising options. The conversion of vegetable oil into ready-biofuel, especially green diesel, needs several steps, one of which is a hydrogenation or hydro-deoxygenation reaction. In this case, the catalyst plays a very important role regarding to its activity and selectivity, and Metal-Organic Frameworks (MOFs) becoming a new generation of heterogeneous catalyst in this area. In this research, a preliminary study to optimize electrochemical synthesis of the catalytic material based on MOFs, namely HKUST-1 [Cu3(BTC)2], has been conducted. Some electrochemical reaction parameters were tested, for example by modifying the electrochemical synthetic conditions, i.e. by performing variation of voltages (12, 13, 14, and 15 Volt), temperatures (RT, 40, 60, and 80 °C) and solvents (ethanol, water, methanol and dimethyl-formamide (DMF)). Material characterization was carried out by XRD, SEM, FTIR, DTA/TG and SAA. The results showed that the optimum synthetic conditions of HKUST-1 are performed at room temperature in a solvent combination of water: ethanol (1: 1) and a voltage of 15 Volt for 2 hours. The XRD-analysis revealed that the resulted peaks are identical to the simulated powder pattern generated from single crystal data and comparable to the peaks of solvothermal method. However, the porosity of the resulting material through electrochemical method is still in the range of micro-pore according to IUPAC and 50% smaller than the porosity resulted from solvothermal synthesis. The corresponding compounds are thermally stable until 300 °C according to TG/DTA.

  19. Characterisation of candidate buffer materials. Vol. 2: Thermo-mechanical calculation of buffer in granitic environment

    International Nuclear Information System (INIS)

    Broc, D.

    1987-01-01

    Mechanical stresses of compacted clays between the canister and the host rock are studied in the different cases during evolution of a vitrified waste storage site. Thermal stress variations are studied in function of time and thermal power decrease of stored wastes and of materials characteristics and behavior. Consequences of stresses produced by partial hydratation of clays are evaluated. The study concludes that an argillaceous containment does not present a rupture risk, even during a partial hydratation in addition stresses on stored packaging are obtained

  20. Deuterium implantation in first wall candidate materials by exposure in the Princeton large torus

    Energy Technology Data Exchange (ETDEWEB)

    Chang, J.; Tobin, A. (Grumman Aerospace Corp., Bethpage, NY (USA). Research and Development Center); Manos, D. (Princeton Univ., NJ (USA). Plasma Physics Lab.)

    Titanium alloys are of interest as a first wall material in fusion reactors because of their excellent thermophysical and thermomechanical properties. A major concern with their application to the first wall is associated with the known affinity of titanium for hydrogen and the related consequences for fuel recycling, tritium inventory, and hydrogen embrittlement. Little information exists on trapping and release of hydrogen isotopes implanted at energies below 500 eV. This work was undertaken to measure hydrogen isotope trapping and release at the first wall of the Princeton Large Torus Tokamak (PLT).

  1. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  2. Nb-Base FS-85 Alloy as a Candidate Structural Material for Space Reactor Applications: Effects of Thermal Aging

    Science.gov (United States)

    Leonard, Keith J.; Busby, Jeremy T.; Hoelzer, David T.; Zinkle, Steven J.

    2009-04-01

    The proposed uses of fission reactors for manned or deep space missions have typically relied on the potential use of refractory metal alloys as structural materials. Throughout the history of these programs, a leading candidate has been Nb-1Zr, due to its good fabrication and welding characteristics. However, the less-than-optimal creep resistance of this alloy has encouraged interest in the more complex FS-85 (Nb-28Ta-10W-1Zr) alloy. Despite this interest, only a relatively small database exists for the properties of FS-85. Database gaps include the potential microstructural instabilities that can lead to mechanical property degradation. In this work, changes in the microstructure and mechanical properties of FS-85 were investigated following 1100 hours of thermal aging at 1098, 1248, and 1398 K. The changes in electrical resistivity, hardness, and tensile properties between the as-annealed and aged materials are compared. Evaluation of the microstructural changes was performed through optical microscopy (OM), scanning electron microscopy (SEM), and transmission electron microscopy (TEM). The development of intragranular and grain-boundary precipitation of Zr-rich compounds as a function of aging temperature was followed. Brittle tensile behavior was measured in the material aged at 1248 K, while ductile behavior occurred in samples aged above and below this temperature. The effect of temperature on the under- and overaging of the grain-boundary particles is believed to have contributed to the mechanical property behavior of the aged materials.

  3. Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

    International Nuclear Information System (INIS)

    McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

    1983-11-01

    A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

  4. Oxidation/volatilization rates in air for candidate fusion reactor blanket materials, PCA and HT-9

    International Nuclear Information System (INIS)

    Piet, S.J.; Kraus, H.G.; Neilson, R.M. Jr.; Jones, J.L.

    1986-01-01

    Large uncertainties exist in the quantity of neutron-induced activation products that can be mobilized in potential fusion accidents. The accidental combination of high temperatures and oxidizing conditions might lead to mobilization of a significant amount of activation products from structural materials. Here, the volatilization of constituents of PCA and HT-9 resulting from oxidation in air was investigated. Tests were conducted in flowing air at temperatures from 600 to 1300 0 C for 1, 5, or 20 h. Elemental volatility was calculated in terms of the weight fraction of the element volatilized from the initial alloy. Molybdenum and manganese were the radiologically significant primary constituents most volatilizized, suggesting that molybdenum and manganese should be minimized in fusion steel compositions. Higher chromium content appears beneficial in reducing hazards from mobile activation products. Scanning electron microscopy and energy dispersive spectroscopy were used to study the oxide layer on samples. (orig.)

  5. The stability of candidate buffer materials for a low-level radioactive waste repository

    International Nuclear Information System (INIS)

    Torok, J.; Buckley, L.P.; Burton, G.R.; Tosello, N.B.; Maves, S.R.; Blimkie, M.E.; Donaldson, J.R.

    1989-11-01

    Inorganic ion-exchangers, clinoptilolite and clay, will be placed on the floor of a low-level radioactive waste repository to be built at Chalk River Nuclear Laboratories. The stability of these ion-exchange materials for a range of potential chemical environments in the repository was investigated. The leaching of waste forms and concrete and biological activity may create acidic or basic environment. The dissolution mechanisms of the ion exchangers for both acid and alkali conditions were established. Changes in distribution coefficients occurred shortly after the commencement of the treatment and were due to changes in the counter-ion content of the ion exchangers. No evidence was found to suggest gradual selective destruction of exchange sites responsible for the high distribution coefficients observed

  6. Oxidation/volatilization rates in air for candidate fusion reactor blanket materials, PCA and HT-9

    International Nuclear Information System (INIS)

    Piet, S.J.; Kraus, H.G.; Neilson, R.M. Jr.; Jones, J.L.

    1986-01-01

    Large uncertainties exist in the quantity of neutron-induced activation products that can be mobilized in potential fusion accidents. The accidental combination of high temperatures and oxidizing conditions might lead to mobilization of a significant amount of activation products from structural materials. Here, the volatilization of constituents of PCA and HT-9 resulting from oxidation in air was investigated. Tests were conducted in flowing air at temperatures from 600 to 1300 0 C for 1, 5, or 20 hours. Elemental volatility was calculated in terms of the weight fraction of the element volatilized from the initial alloy. Molybdenum and manganese were the radiologically significant primary constituents most volatilized, suggesting that molybdenum and manganese should be minimized in fusion steel compositions. Higher chromium content appears beneficial in reducing hazards from mobile activation products. Scanning electron microscopy and energy dispersive spectroscopy were used to study the oxide layer on samples

  7. ITER overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.R.

    2001-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  8. ITER Overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.

    1999-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  9. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  10. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  11. Formulation of a candidate glass for use as an acceptance test standard material

    International Nuclear Information System (INIS)

    Ebert, W.L.; Strachan, D.M.; Wolf, S.F.

    1998-04-01

    In this report, the authors discuss the formulation of a glass that will be used in a laboratory testing program designed to measure the precision of test methods identified in the privatization contracts for the immobilization of Hanford low-activity wastes. Tests will be conducted with that glass to measure the reproducibility of tests and analyses that must be performed by glass producers as a part of the product acceptance procedure. Test results will be used to determine if the contractually required tests and analyses are adequate for evaluating the acceptability of likely immobilized low-activity waste (ILAW) products. They will also be used to evaluate if the glass designed for use in these tests can be used as an analytical standard test material for verifying results reported by vendors for tests withg ILAW products. The results of those tests and analyses will be presented in a separate report. The purpose of this report is to document the strategy used to formulate the glass to be used in the testing program. The low-activity waste reference glass LRM that will be used in the testing program was formulated to be compositionally similar to ILAW products to be made with wastes from Hanford. Since the ILAW product compositions have not been disclosed by the vendors participating in the Hanford privatization project, the composition of LRM was formulated based on simulated Hanford waste stream and amounts of added glass forming chemicals typical for vitrified waste forms. The major components are 54 mass % SiO 2 , 20 mass % Na 2 O, 10 mass % Al 2 O 3 , 8 mass % B 2 O 3 , and 1.5 mass % K 2 O. Small amounts of other chemicals not present in Hanford wastes were also included in the glass, since they may be included as chemical additives in ILAW products. This was done so that the use of LRM as a composition standard could be evaluated. Radionuclides were not included in LRM because a nonradioactive material was desired

  12. The future supply of and demand for candidate materials for the fabrication of nuclear fuel waste disposal containers

    International Nuclear Information System (INIS)

    Grover, L.K.

    1990-01-01

    This report summarizes the findings of a literature survey carried out to assess the future world supply of and demand for titanium, copper and lead. These metals are candidate materials for the fabrication of containers for the immobilization and disposal of Canada's nuclear used-fuel waste for a reference Used-fuel Disposal Centre. Such a facility may begin operation by approximately 2020, and continue for about 40 years. The survey shows that the world has abundant supplies of titanium minerals (mostly in the form of ilmenite), which are expected to last up to at least 2110. However, for copper and lead the balance between supply and demand may warrant increased monitoring beyond the year 2000. A number of factors that can influence future supply and demand are discussed in the report

  13. Thermophysical and heat transfer properties of phase change material candidate for waste heat transportation system

    Science.gov (United States)

    Kaizawa, Akihide; Maruoka, Nobuhiro; Kawai, Atsushi; Kamano, Hiroomi; Jozuka, Tetsuji; Senda, Takeshi; Akiyama, Tomohiro

    2008-05-01

    A waste heat transportation system trans-heat (TH) system is quite attractive that uses the latent heat of a phase change material (PCM). The purpose of this paper is to study the thermophysical properties of various sugars and sodium acetate trihydrate (SAT) as PCMs for a practical TH system and the heat transfer property between PCM selected and heat transfer oil, by using differential scanning calorimetry (DSC), thermogravimetry-differential thermal analysis (TG-DTA) and a heat storage tube. As a result, erythritol, with a large latent heat of 344 kJ/kg at melting point of 117°C, high decomposition point of 160°C and excellent chemical stability under repeated phase change cycles was found to be the best PCM among them for the practical TH system. In the heat release experiments between liquid erythritol and flowing cold oil, we observed foaming phenomena of encapsulated oil, in which oil droplet was coated by solidification of PCM.

  14. Tensile properties of candidate structural materials for high power spallation sources at high helium contents

    Science.gov (United States)

    Jung, P.; Henry, J.; Chen, J.

    2005-08-01

    Low activation 9%Cr martensitic steels EUROFER97, pure tantalum, and low carbon austenitic stainless steel 316L were homogeneously implanted with helium to concentrations up to 5000 appm at temperatures from 70 °C to 400 °C. The specimens were tensile tested at room temperature and at the respective implantation temperatures. In all materials the helium caused an increased in strength and reduction in ductility, with both changes being generally larger at lower implantation and testing temperatures. After implantation some work hardening was retained in 316L and in tantalum, while it almost completely disappeared in EUROFER97. After tensile testing, fracture surfaces were analysed by scanning electron microscopy (SEM). Implantation caused reduction of necking, but up to concentrations of 2500 appm He fracture surface still showed transgranular ductile appearance. Completely brittle intergranular fracture was observed in tantalum at 9000 appm He and is also expected for EUROFER97 at this concentration, according to previous results on similar 9%Cr steels.

  15. Surface damage of TFTR protective plate candidate materials by energetic D+ irradiation

    International Nuclear Information System (INIS)

    Kaminsky, M.; Das, S.K.

    1979-01-01

    Experiments were conducted to determine the surface damage of ATJ graphite, V, Cu, and Type 316 stainless steel under 60-keV D + irradiation. The irradiations were conducted in the pulsed mode. For a total accumulated dose of 8.1 x 10 18 ions/cm 2 , blisters were readily seen for Cu surfaces, but no blisters were observed on Type 316 stainless steel and vanadium surfaces. For the case of ATJ graphite, the surface damage was observed in the form of ridges and grooves. In the case of copper, many large blisters with diameters ranging from 3.5 μm to 46 μm are observed in addition to some small ones (average diameter approx. 2 μm. The blister density of the large blisters is the highest in the case of copper (1.1 x 10 5 blisters/cm 2 ). These observations of blister formation are related to the differences in the premeability of deuterium in these materials. An examination of the cross section of the ridges in fractured samples of graphite indicates that they are not hollow. The mechanisms of formation of these ridges is not clear at present. 1 figure

  16. Laboratory scale development of coating for improving characteristics of candidate materials for fusion reactor

    International Nuclear Information System (INIS)

    Agarwala, R.P.

    1989-01-01

    Application of coatings of refractory low atomic number materials on to different components of Tokamak type controlled thermonuclear reactor are expected to provide a degree of design flexibility. The project envisages to deal with the challenging problem on laboratory scale. Coatings investigated include carbon, beryllium, boron, titanium carbide and alumina and substrates chosen have been 304, 316 stainless steels, monel-400, molybdenum, copper, graphite, etc. For their deposition, different techniques (e.g. evaporation, sputtering and their different variants) have been tried, appropriate ones chosen and their parameters optimized. The coating composition has been analyzed using X-ray diffraction (XRD), Auger electron spectroscopy (AES), X-ray photoelectron spectroscopy (XPS), Rutherford backscattering analysis (RBS) and secondary ions mass spectroscopy (SIMS). Surface morphology has been studied using scanning electron microscopy (SEM). Sebastian coating adherence tester has been used for adhesion measurement and Wilson's Tukon microhardness tester for their microhardness measurement. The coatings have been subjected to pulses from YAG laser to evaluate their thermal cycling behaviour. Deuterium ion bombardment (Energy: 20-120 keV; doses: 10 19 -9.3x10 20 ions/cm 2 ) behaviour has also been studied. In general, adherent and hard coatings capable of withstanding thermal cycling could be deposited. Out of the coatings studied, titanium carbide shows best results. The following pages are reprints and not mircrofiched: p. 25-32, 39-41, 57-81. Bibliographic description is on page 13

  17. Candidate solar cell materials for photovoltaic conversion in a solar power satellite /SPS/

    Science.gov (United States)

    Glaser, P. E.; Almgren, D. W.

    1978-01-01

    In recognition of the obstacles to solar-generated baseload power on earth, proposals have been made to locate solar power satellites in geosynchronous earth orbit (GEO), where solar energy would be available 24 hours a day during most of the time of the year. In an SPS, the electricity produced by solar energy conversion will be fed to microwave generators forming part of a planar phase-array transmitting antenna. The antenna is designed to precisely direct a microwave beam of very low intensity to one or more receiving antennas at desired locations on earth. At the receiving antenna, the microwave energy will be safely and efficiently reconverted to electricity and then be transmitted to consumers. An SPS system will include a number of satellites in GEO. Attention is given to the photovoltaic option for solar energy conversion in GEO, solar cell requirements, the availability of materials, the implication of large production volumes, requirements for high-volume manufacture of solar cell arrays, and the effects of concentration ratio on solar cell array area.

  18. Deuterium ion irradiation damage and deuterium trapping mechanism in candidate stainless steel material (JPCA2) for fusion reactor

    International Nuclear Information System (INIS)

    Ashizuka, Norihiro; Kurita, Takaaki; Yoshida, Naoaki; Fujiwara, Tadashi; Muroga, Takeo

    1987-01-01

    An improved austenitic stainless steel (JPCA), a candidate material for fusion reactor, is irradiated at room temperature with deuterium ion beams. Desorption spectra of deuterium gas is measured at various increased temperatures and defects formed under irradiation are observed by transmission electron microscopy to determine the mechanism of the thermal release of deuteriums and the characteristics of irradiation-induced defects involved in the process. In the deuterium deportion spectra observed, five release stages are found to exist at 90 deg C, 160 deg C, 220 deg C, 300 deg C and 400 deg C, referred to as Stage I, II, III, IV and V, respectively. Stage I is interpreted as representing the release of deuteriums trapped in point defects (presumably vacancies) formed under irradiation. The energy of desorption from the trapping sites is estimated at 0.8 eV. Stage II is concluded to be associated with the release of deuteriums trapped in a certain kind of existing defects. Stage III involves the release of deuteriums that are trapped in dislocations, dislocation loops or dislocated portions of stacking fault tetrahedra. This release occurs significantly in processed materials and other materials irradiated with high energy ion beams that may cause cascade damage. Stage IV is interpreted in terms of thermal decomposition of small deuterium clusters. Stage V is associated with the decomposition of rather large deuterium clusters grown on the {111} plane. (Nogami, K.)

  19. Investigation on candidates of principal facilities for exposure dose to public for the facilities using nuclear material

    International Nuclear Information System (INIS)

    Shimazaki, Yosuke; Sawahata, Hiroaki; Takada, Shoji; Fujimoto, Nozomu

    2015-01-01

    HTTR holds the nuclear fuel material use facilities in its reactor facilities, for the purpose of study on the fracture behavior of fuel and release behavior of fission products, development of high-performance fuel, and measurement of neutron flux. Due to the revision of the 'Act on the regulation of nuclear source material, nuclear fuel material and reactor', the facilities having the 'Important safety-related facilities' among the facilities applicable to the Enforcement Ordinance Article 41 (Article 41 facilities) has come to need to conform to the 'Regulations concerning standards for the location, structure, and equipment of used facilities and others'. In this case, actions such as modification by all possible means are required. The nuclear fuel substance use facilities of HTTR correspond to Article 41 facilities. So, whether it is a candidate for the 'Important safety-related facilities' has been examined. As a result, it is confirmed that the facilities are not correspond to the 'Important safety-related facilities', and it has been concluded that modification measures for the purpose of conforming to this approval standard rule are not necessary as of the present. (A.O.)

  20. Solution of problems with material nonlinearities with a coupled finite element/boundary element scheme using an iterative solver. Yucca Mountain Site Characterization Project

    International Nuclear Information System (INIS)

    Koteras, J.R.

    1996-01-01

    The prediction of stresses and displacements around tunnels buried deep within the earth is an important class of geomechanics problems. The material behavior immediately surrounding the tunnel is typically nonlinear. The surrounding mass, even if it is nonlinear, can usually be characterized by a simple linear elastic model. The finite element method is best suited for modeling nonlinear materials of limited volume, while the boundary element method is well suited for modeling large volumes of linear elastic material. A computational scheme that couples the finite element and boundary element methods would seem particularly useful for geomechanics problems. A variety of coupling schemes have been proposed, but they rely on direct solution methods. Direct solution techniques have large storage requirements that become cumbersome for large-scale three-dimensional problems. An alternative to direct solution methods is iterative solution techniques. A scheme has been developed for coupling the finite element and boundary element methods that uses an iterative solution method. This report shows that this coupling scheme is valid for problems where nonlinear material behavior occurs in the finite element region

  1. Characterisation of bentonites from Kutch, India and Milos, Greece - some candidate tunnel back-fill materials?

    International Nuclear Information System (INIS)

    Olsson, Siv; Karnland, Ola

    2009-12-01

    During the past decades comprehensive investigations have been made on bentonite clays in order to find optimal components of the multi-barrier system of repositories for radioactive waste. The present study gives a mineralogical characterisation of some selected bentonites, in order to supply some of the necessary background data on the bentonites for evaluating their potential as tunnel back-fill materials. Two bentonites from the island of Milos, Greece (Milos BF 04 and BF 08), and two bentonites from Kutch, India (Kutch BF 04 and BF 08) were analysed for their grain size distribution, cation exchange properties and chemical composition. The mineralogical composition was determined by X-ray diffraction analysis and evaluated quantitatively by use of the Siroquant software. Both the bulk bentonite and the 63 μm. The bentonite is distinguished by a high content of dolomite and calcite, which make up almost 25% of the bulk sample. The major accessory minerals are K-feldspars and plagioclase, whereas the content of sulphur-bearing minerals is very low (0.06% total S). Smectite makes up around 60% of the bulk sample, which has a CEC value of 73 meq/100 g. The pool of interlayer cations has a composition Mg>Ca>>Na>>K. The X-ray diffraction characteristics and the high potassium content (1.03% K 2 O) of the Na>Mg>>K. The 2 O) which indicates that also this smectite may be interstratified with a few percent illitic layers. Based on the charge distribution the smectite should be classified as montmorillonite but in this case Fe predominates over Mg in the octahedral sheet. The structural formula suggests that this smectite has the lowest total layer charge of the smectites examined. Kutch BF 04 contains essentially no particles >63 μm. The bentonite has a high content of titanium and iron-rich accessory minerals, such as anatase, magnetite, hematite and goethite. Other accessory minerals of significance are feldspars and quartz, whereas the content of sulphur

  2. Developing and Evaluating Candidate Materials for Generation IV Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Jang, Jin Sung; Kim, Sung Ho; Hwang Sung Sik and others

    2006-03-01

    High temperature mechanical behavior High temperature behavior of two F-M steels were investigated, considering the transient temperature range of the SCWR (above 800 .deg. C). T91 and T122 specimens were five times cyclically heat treated to the temperature 810 .deg. C and 845 .deg. C respectively. And the heat treatments were found to have little effect on the creep rupture behavior at 550, 600, or 650 .deg. C. However, the microstructural change was detected by the rapid hardness change after the holding the specimens at 840 .deg. C even for 10 sec. (by INL, previously ANL-W) A 20Cr Fe-base ODS alloy (MA956) was isothermally heat treated at 475 .deg. C for various times and then impact tested. The material was found to become very brittle after the heat treatment even for 100 hrs by the drastic decrease of the impact absorption energy (from 300 J to about the nil) and by the typically brittle fracture surface. (by KAIST) Corrosion and SCC Behavior in SCW (1) The corrosion behaviors of the F-M steels (T91, T92, and T122) and high Ni alloys (alloy 625, Alloy 690, and alloy 800H) and an ODS alloy (MA 956) were studied in the aerated SCW (8 ppm of D.O; dissolved oxygen) under 25 MPa from 300 to 600 .deg. C with an interval of 50 .deg. C. The test durations were 100, 200, and 500 hrs respectively. In general high Ni alloys were definitely more resistant to corrosion in SCW than F-M steels. As the Cr content increases the resistance of F-M steels to corrosion becomes better. The resistance of F-M steels to corrosion at 350 .deg. C, a subcritical temperature, was revealed to be comparatively similar to those at 550 .deg. C, a 200 .deg. C higher temperature. (2) The SCC resistance of F-M steels, T91 and T92, was evaluated by CERT (constant extension rate test) method. T91 specimens were tested at 500, 550 and 600 .deg. C in a fully deaerated SCW (below 10 ppb D.O), and SCC did not happen in the T91 specimens. T92 specimens were tested at 500 .deg. C in SCW of different

  3. Characterisation of bentonites from Kutch, India and Milos, Greece - some candidate tunnel back-fill materials?

    Energy Technology Data Exchange (ETDEWEB)

    Olsson, Siv; Karnland, Ola (Clay Technology AB, Lund (Sweden))

    2009-12-15

    During the past decades comprehensive investigations have been made on bentonite clays in order to find optimal components of the multi-barrier system of repositories for radioactive waste. The present study gives a mineralogical characterisation of some selected bentonites, in order to supply some of the necessary background data on the bentonites for evaluating their potential as tunnel back-fill materials. Two bentonites from the island of Milos, Greece (Milos BF 04 and BF 08), and two bentonites from Kutch, India (Kutch BF 04 and BF 08) were analysed for their grain size distribution, cation exchange properties and chemical composition. The mineralogical composition was determined by X-ray diffraction analysis and evaluated quantitatively by use of the Siroquant software. Both the bulk bentonite and the <1mum fraction were analyzed when relevant. Prior to the chemical analyses the <1 mum fractions were converted to homo-ionic clays and purified by dialysis. The chemical data were used for calculating the structural formula of the smectites. Milos BF 04 contains ca. 10% particles >63 mum. The bentonite is distinguished by a high content of dolomite and calcite, which make up almost 25% of the bulk sample. The major accessory minerals are K-feldspars and plagioclase, whereas the content of sulphur-bearing minerals is very low (0.06% total S). Smectite makes up around 60% of the bulk sample, which has a CEC value of 73 meq/100 g. The pool of interlayer cations has a composition Mg>Ca>>Na>>K. The X-ray diffraction characteristics and the high potassium content (1.03% K{sub 2}O) of the <1 mum fraction suggest that the smectite is interstratified with ca. 10% illitic layers. Based on the charge distribution the smectite should be classified as montmorillonite and according to the structural formula, Mg predominates over Fe in the octahedral sheet. However, remnants of Mg-carbonates, if present, may be a source of error in the formula calculation. Milos BF 08 has a

  4. Developing and Evaluating Candidate Materials for Generation IV Supercritical Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Jin Sung; Kim, Sung Ho; Hwang Sung Sik and others

    2006-03-15

    High temperature mechanical behavior High temperature behavior of two F-M steels were investigated, considering the transient temperature range of the SCWR (above 800 .deg. C). T91 and T122 specimens were five times cyclically heat treated to the temperature 810 .deg. C and 845 .deg. C respectively. And the heat treatments were found to have little effect on the creep rupture behavior at 550, 600, or 650 .deg. C. However, the microstructural change was detected by the rapid hardness change after the holding the specimens at 840 .deg. C even for 10 sec. (by INL, previously ANL-W) A 20Cr Fe-base ODS alloy (MA956) was isothermally heat treated at 475 .deg. C for various times and then impact tested. The material was found to become very brittle after the heat treatment even for 100 hrs by the drastic decrease of the impact absorption energy (from 300 J to about the nil) and by the typically brittle fracture surface. (by KAIST) Corrosion and SCC Behavior in SCW (1) The corrosion behaviors of the F-M steels (T91, T92, and T122) and high Ni alloys (alloy 625, Alloy 690, and alloy 800H) and an ODS alloy (MA 956) were studied in the aerated SCW (8 ppm of D.O; dissolved oxygen) under 25 MPa from 300 to 600 .deg. C with an interval of 50 .deg. C. The test durations were 100, 200, and 500 hrs respectively. In general high Ni alloys were definitely more resistant to corrosion in SCW than F-M steels. As the Cr content increases the resistance of F-M steels to corrosion becomes better. The resistance of F-M steels to corrosion at 350 .deg. C, a subcritical temperature, was revealed to be comparatively similar to those at 550 .deg. C, a 200 .deg. C higher temperature. (2) The SCC resistance of F-M steels, T91 and T92, was evaluated by CERT (constant extension rate test) method. T91 specimens were tested at 500, 550 and 600 .deg. C in a fully deaerated SCW (below 10 ppb D.O), and SCC did not happen in the T91 specimens. T92 specimens were tested at 500 .deg. C in SCW of different

  5. Development of sampling techniques for ITER Type B radwaste

    International Nuclear Information System (INIS)

    Hong, Kwon Pyo; Kim, Sung Geun; Jung, Sang Hee; Oh, Wan Ho; Park, Myung Chul; Kim, Hee Moon; Ahn, Sang Bok

    2016-01-01

    There are several difficulties and limitation in sampling activities. As the Type B radwaste components are mostly metallic(mostly stainless steel) and bulk(∼ 1 m in size and ∼ 100 mm in thickness), it is difficult in taking samples from the surface of Type B radwaste by remote operation. But also, sampling should be performed without use of any liquid coolant to avoid the spread of contamination. And all sampling procedures are carried in the hot cell red zone with remote operation. Three kinds of sampling techniques are being developed. They are core sampling, chip sampling, and wedge sampling, which are the candidates of sampling techniques to be applied to ITER hot cell. Object materials for sampling are stainless steel or Cu alloy block in order to simulate ITER Type B radwaste. The best sampling technique for ITER Type B radwaste among the three sampling techniques will be suggested in several months after finishing the related experiment

  6. Development of sampling techniques for ITER Type B radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Kwon Pyo; Kim, Sung Geun; Jung, Sang Hee; Oh, Wan Ho; Park, Myung Chul; Kim, Hee Moon; Ahn, Sang Bok [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    There are several difficulties and limitation in sampling activities. As the Type B radwaste components are mostly metallic(mostly stainless steel) and bulk(∼ 1 m in size and ∼ 100 mm in thickness), it is difficult in taking samples from the surface of Type B radwaste by remote operation. But also, sampling should be performed without use of any liquid coolant to avoid the spread of contamination. And all sampling procedures are carried in the hot cell red zone with remote operation. Three kinds of sampling techniques are being developed. They are core sampling, chip sampling, and wedge sampling, which are the candidates of sampling techniques to be applied to ITER hot cell. Object materials for sampling are stainless steel or Cu alloy block in order to simulate ITER Type B radwaste. The best sampling technique for ITER Type B radwaste among the three sampling techniques will be suggested in several months after finishing the related experiment.

  7. Photoelectron Yield and Photon Reflectivity from Candidate LHC Vacuum Chamber Materials with Implications to the Vacuum Chamber Design

    CERN Document Server

    Baglin, V; Gröbner, Oswald

    1998-01-01

    Studies of the photoelectron yield and photon reflectivity at grazing incidence (11 mrad) from candidate LHC vacuum chamber materials have been made on a dedicated beam line on the Electron Positron A ccumulator (EPA) ring at CERN. These measurements provide realistic input toward a better understanding of the electron cloud phenomena expected in the LHC. The measurements were made using synchrotro n radiation with critical photon energies of 194 eV and 45 eV; the latter corresponding to that of the LHC at the design energy of 7 TeV. The test materials are mainly copper, either, i) coated by co- lamination or by electroplating onto stainless steel, or ii) bulk copper prepared by special machining. The key parameters explored were the effect of surface roughness on the reflectivity and the pho toelectron yield at grazing photon incidence, and the effect of magnetic field direction on the yields measured at normal photon incidence. The implications of the results on the electron cloud phenom ena, and thus the L...

  8. Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers. Final report

    International Nuclear Information System (INIS)

    Vinson, D.W.; Bullen, D.B.

    1995-01-01

    One of the most significant factors impacting the performance of waste package container materials under repository relevant conditions is the thermal environment. This environment will be affected by the areal power density of the repository, which is dictated by facility design, and the dominant heat transfer mechanism at the site. The near-field environment will evolve as radioactive decay decreases the thermal output of each waste package. Recent calculations (Buscheck and Nitao, 1994) have addressed the importance of thermal loading conditions on waste package performance at the Yucca Mountain site. If a relatively low repository thermal loading design is employed, the temperature and relative humidity near the waste package may significantly affect the degradation of corrosion allowance barriers due to moist air oxidation and radiolytically enhanced corrosion. The purpose this report is to present a literature review of the potential degradation modes for moderately corrosion resistant nickel copper and nickel based candidate materials that may be applicable as alternate barriers for the ACD systems in the Yucca Mountain environment. This report presents a review of the corrosion of nickel-copper alloys, summaries of experimental evaluations of oxidation and atmospheric corrosion in nickel-copper alloys, views of experimental studies of aqueous corrosion in nickel copper alloys, a brief review of galvanic corrosion effects and a summary of stress corrosion cracking in these alloys

  9. ITER licensing

    International Nuclear Information System (INIS)

    Gordon, C.W.

    2005-01-01

    ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

  10. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  11. Water-cooled Pb-17Li test blanket module for ITER: impact of the structural material grade on the neutronic responses

    Energy Technology Data Exchange (ETDEWEB)

    Vella, G.; Aiello, G.; Oliveri, E. [Palermo Univ. (Italy). Dipt. di Ingegneria Nucl.; Fuetterer, M.A.; Giancarli, L. [CEA - Saclay, DRN/DMT/SERMA, Gif-sur-Yvette (France); Tavassoli, F. [CEA - Saclay, CEREM, Gif-sur-Yvette (France)

    1998-10-01

    The water-cooled lithium lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a test blanket module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL-TBM has been inserted into an existing ITER model accounting for a proper D-T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material damage through DPA and He and H gas production rate, radial distribution of tritium production rate and total tritium production in the module. The impact of using natural lithium on the TBM system operation has also been evaluated. (orig.) 13 refs.

  12. HCV INFECTION THROUGH PERFORATING AND CUTTING MATERIAL AMONG CANDIDATES FOR BLOOD DONATION IN BELÉM, BRAZILIAN AMAZON

    Directory of Open Access Journals (Sweden)

    Rubenilson Caldas Valois

    2014-12-01

    Full Text Available This study evaluated epidemiological factors for HCV infection associated with sharing perforating and cutting instruments among candidates for blood donation (CBD in the city of Belém, Pará, Brazilian Amazon. Two definitions of HCV infection cases were used: anti-HCV positivity shown by EIA, and HCV-RNA detection by PCR. Infected and uninfected CBD completed a questionnaire about possible risk factors associated with sharing perforating and cutting instruments. The information was evaluated using simple and multiple logistic regressions. Between May and November 2010, 146 (1.1% persons with anti-HCV antibodies and 106 (0.8% with HCV-RNA were detected among 13,772 CBD in Belém. Risk factors associated with HCV infection based on the EIA (model 1 and PCR (model 2 results were: use of needles and syringes sterilized at home; shared use of razors at home, sharing of disposable razors in barbershops, beauty salons etc.; and sharing manicure and pedicure material. The models of HCV infection associated with sharing perforating and cutting instruments should be taken into account by local and regional health authorities and by those of other countries with similar cultural practices, in order to provide useful information to guide political and public strategies to control HCV transmission.

  13. Impact of phase stability on the corrosion behavior of the austenitic candidate materials for NNWSI [Nevada Nuclear Waste Storage Investigations

    International Nuclear Information System (INIS)

    Bullen, D.B.; Gdowski, G.E.; McCright, R.D.

    1987-10-01

    The Nuclear Waste Management Program at Lawrence Livermore National Laboratory is responsible for the development of the waste package design to meet the Nuclear Regulatory Commission licensing requirements for the Nevada Nuclear Waste Storage Investigations (NNWSI) Project. The metallic container component of the waste package is required to assist in providing substantially complete containment of the waste for a period of up to 1000 years. Long term phase stability of the austenitic candidate materials (304L and 316L stainless steels and alloy 825) over this time period at moderate temperatures (100-250 0 C) can impact the mechanical and corrosion behavior of the metal barrier. A review of the technical literature with respect to phase stability of 304L, 316L and 825 is presented. The impact of martensitic transformations, carbide precipitation and intermediate (σ, chi, and eta) phase formation on the mechanical properties and corrosion behavior of these alloys at repository relevant conditions is discussed. The effect of sensitization on intergranular stress corrosion cracking (IGSCC) of each alloy is also addressed. A summary of the impact of phase stability on the degradation of each alloy in the proposed repository environment is included. 32 refs., 6 figs

  14. The influence of Saccharomyces cerevisiae enzyme ratio on preparation virgin coconut oil for candidate in-house reference materials

    Science.gov (United States)

    Rohyami, Yuli; Anjani, Rafika Debby; Purwanti, Napthalina Putri

    2017-03-01

    Virgin coconut oil is an excellent product which has result of oil processing business opportunities in the international market. Standardization of virgin coconut oil necessary to satisfy the requirements industry needs. This research is expected as procedure preparation of reference materials. Preparation of virgin coconut oil by Sacharomycescerevisiaeenzyme. Based on the results of this study concluded that the ratio of Saccharomyces cerevisiae can affect the yield of virgin coconut oil produced. The preparation of virgin coconut oil enzymatically using a variety of mass ratio of 0.001 to 0.006% is obtained yield average of 12.40%. The optimum separation of virgin coconut oil on the use of enzymes with a mass ratio of 0.002%. The average water content at a ratio of 0.002% is 0.04 % with a value of uncertainty is 0.005%. The average iodine number in virgin coconut oil produced is 2.4403 ± 0,1974 grams of iodine per 100 grams of oil and optimum iodine number is obtained from the manufacturing process virgin coconut oil with a ratio of 0.006% Saccharomyces cerevisiae. Sacharomycescerevisiae with a ratio of 0.002% results virgin coconut oil with acid number 0.3068 ± 0.1098%. The peroxide value of virgin coconut oil between 0.0108 ± 0.009 to 0.0114 ± 0015milli-equivalent per kilograms. Organoleptic test results and test chemical parameters can be used as the test data that can be developed in prototype preparation of candidate in-house reference material in the testing standards of quality virgin coconut oil.

  15. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  16. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  17. Survey of total error of precipitation and homogeneous HDL-cholesterol methods and simultaneous evaluation of lyophilized saccharose-containing candidate reference materials for HDL-cholesterol

    NARCIS (Netherlands)

    C.M. Cobbaert (Christa); H. Baadenhuijsen; L. Zwang (Louwerens); C.W. Weykamp; P.N. Demacker; P.G.H. Mulder (Paul)

    1999-01-01

    textabstractBACKGROUND: Standardization of HDL-cholesterol is needed for risk assessment. We assessed for the first time the accuracy of HDL-cholesterol testing in The Netherlands and evaluated 11 candidate reference materials (CRMs). METHODS: The total error (TE) of

  18. ITER ITA newsletter. No. 21, April 2005

    International Nuclear Information System (INIS)

    2005-05-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about Russian federation Participant Team's activity in the area of preparation for ITER construction and information about International Fusion materials irradiation Facility(IRMIF) project and prospects for implementation

  19. ITER EDA newsletter. V. 8, no. 11

    International Nuclear Information System (INIS)

    1999-11-01

    This ITER EDA Newsletter contains summary reports on the eleventh meeting of the ITER diagnostic expert group in Cadarache, France, on the ITER JCT presentation at the international conference on fusion reactor materials in Colorado Springs, USA and on the seventh workshop on plasma edge theory in fusion devices in Tajimi, Japan. Individual abstracts are prepared for the three contributions

  20. Behavior of divertor and first wall armour materials at plasma heat fluxes relevant to ITER ELMs and disruptions

    Directory of Open Access Journals (Sweden)

    D.V. Kovalenko

    2017-08-01

    Full Text Available The paper presents the main results of numerous experiments carried out over the past 10 years at QSPA-T and QSPA-Be plasma guns in support of ITER. Special targets made of pure W, W-1%La2O3 and two types of Be (TGP-56FW and S65-C were tested under the series of repeated plasma stream and photonic flux impact. Maximum heat load on the target surface was up to 2.5MJ/m2 in the case of plasma testing and was equal to 0.5MJ/m2 in the case of photonic flux testing. Pulse waveform was rectangular with tpulse= 0.5ms. It was found that the main erosion mechanisms of W and Be under plasma stream impact are the melt layer movement, the ejection of droplets and the cracks formation. As a result of repeated photonic fluxes a regular, “corrugated” structure are eventually formed on the Be target surface. Study of erosion products of W formed under plasma stream impact on the W target has shown that the D/W atomic ratio in the deposited W films during pulsed events may be the same or even higher than that for stationary processes.

  1. An iterative method for near-field Fresnel region polychromatic phase contrast imaging

    Science.gov (United States)

    Carroll, Aidan J.; van Riessen, Grant A.; Balaur, Eugeniu; Dolbnya, Igor P.; Tran, Giang N.; Peele, Andrew G.

    2017-07-01

    We present an iterative method for polychromatic phase contrast imaging that is suitable for broadband illumination and which allows for the quantitative determination of the thickness of an object given the refractive index of the sample material. Experimental and simulation results suggest the iterative method provides comparable image quality and quantitative object thickness determination when compared to the analytical polychromatic transport of intensity and contrast transfer function methods. The ability of the iterative method to work over a wider range of experimental conditions means the iterative method is a suitable candidate for use with polychromatic illumination and may deliver more utility for laboratory-based x-ray sources, which typically have a broad spectrum.

  2. Ceramic breeder materials

    International Nuclear Information System (INIS)

    Johnson, C.E.

    1990-01-01

    The breeding blanket is a key component of the fusion reactor because it directly involves tritium breeding and energy extraction, both of which are critical to development of fusion power. The lithium ceramics continue to show promise as candidate breeder materials. This promise was recognized by the International Thermonuclear Reactor (ITER) design team in its selection of ceramics as the first option for the ITER breeder material. Blanket design studies have indicated properties in the candidate materials data base that need further investigation. Current studies are focusing on tritium release behavior at high burnup, changes in thermophysical properties with burnup, compatibility between the ceramic breeder and beryllium multiplier, and phase changes with burnup. Laboratory and in-reactor tests, some as part of an international collaboration for development of ceramic breeder materials, are underway. 32 refs., 1 fig., 1 tab

  3. Processing of W-Cu functionally graded materials (FGM) through the powder metallurgy route: application as plasma facing components for ITER-like thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Raharijaona, J.J.

    2009-11-01

    The aim of this study was to study and optimize the sintering of W-Cu graded composition materials, for first wall of ITER-like thermonuclear reactor application. The graded composition in the material generates graded functional properties (Functionally Graded Materials - FGM). Rough thermomechanical calculations have shown the interest of W-Cu FGM to improve the lifetime of Plasma Facing Components (PFC). To process W-Cu FGM, powder metallurgy route was analyzed and optimized from W-CuO powder mixtures. The influence of oxide reduction on the sintering of powder mixtures was highlighted. An optimal heating treatment under He/H 2 atmosphere was determined. The sintering mechanisms were deduced from the analysis of the effect of the Cu-content. Sintering of W-Cu materials with a graded composition and grain size has revealed two liquid migration steps: i) capillary migration, after the Cu-melting and, ii) expulsion of liquid, at the end of sintering, from the dense part to the porous part, due to the continuation of W-skeleton sintering. These two steps were confirmed by a model based on capillary pressure calculation. In addition, thermal conductivity measurements were conducted on sintered parts and showed values which gradually increase with the Cu-content. Hardness tests on a polished cross-section in the bulk are consistent with the composition profiles obtained and the differential grain size. (author)

  4. Study of deuterium retention in/release from ITER-relevant Be-containing mixed material layers implanted at elevated temperatures

    Energy Technology Data Exchange (ETDEWEB)

    Sugiyama, K., E-mail: kazuyoshi.sugiyama@ipp.mpg.de [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Porosnicu, C. [National Institute for Laser, Plasma and Radiation Physics, EURATOM-MEdC Association, 077125 Bucharest (Romania); Jacob, W.; Roth, J.; Dürbeck, Th. [Max-Planck-Institut für Plasmaphysik, EURATOM Association, D-85748 Garching (Germany); Jepu, I.; Lungu, C.P. [National Institute for Laser, Plasma and Radiation Physics, EURATOM-MEdC Association, 077125 Bucharest (Romania)

    2013-07-15

    D implantation into Be-containing mixed material layers: Be, Be–W (W: ∼6 at.%) and Be–C (C: ∼50 at.%), was performed at elevated temperatures. The temperature dependence of D retention varied depending on the admixed element. D retention in Be and Be–W layers decreases with increasing implantation temperature, while the Be–C layers maintained rather high D retention in the present investigated temperature range (up to 623 K). D desorption behaviour from Be–C suggests the contribution of C–D bonds to D retention. W admixture into Be can significantly suppress D retention in Be. Long-term isothermal annealing at 513 and 623 K for D removal was also performed to simulate the ITER-wall-baking scenario. Even extended annealing at temperatures comparable to or lower than the implantation temperature does not lead to a significant release of retained D.

  5. Evaluation of Method-Specific Extraction Variability for the Measurement of Fatty Acids in a Candidate Infant/Adult Nutritional Formula Reference Material.

    Science.gov (United States)

    Place, Benjamin J

    2017-05-01

    To address community needs, the National Institute of Standards and Technology has developed a candidate Standard Reference Material (SRM) for infant/adult nutritional formula based on milk and whey protein concentrates with isolated soy protein called SRM 1869 Infant/Adult Nutritional Formula. One major component of this candidate SRM is the fatty acid content. In this study, multiple extraction techniques were evaluated to quantify the fatty acids in this new material. Extraction methods that were based on lipid extraction followed by transesterification resulted in lower mass fraction values for all fatty acids than the values measured by methods utilizing in situ transesterification followed by fatty acid methyl ester extraction (ISTE). An ISTE method, based on the identified optimal parameters, was used to determine the fatty acid content of the new infant/adult nutritional formula reference material.

  6. Outgassing rates before, during and after bake-out for various vacuum and first wall candidate materials of a large tokamak device

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Gomay, J.; Sugiyama, Y.; Mizuno, M.; Komiya, S.; Tazima, T.

    1977-01-01

    Outgassing rates of vacuum wall candidate materials; stainless steel SS-304L and YUS-170, Inconel-625 and Hastelloy-X, and first wall materials; molybdenum, pyrolytic graphite and silicon carbide are measured before, during and after a bake-out at 500 0 C. The outgassing rate from the inside wall of the cylinder made of each material is estimated from the pressure difference between before and after a calibrated orifice. The ultimate outgassing rates of SS-304L and pyrolytic graphite, and YUS-170 Inconel-625, Hastelloy-X and molybdenum are the orders of 10 -10 and 10 -11 Pa.l.s -1 cm -2 , respectively

  7. US ITER Management Plan

    International Nuclear Information System (INIS)

    1991-12-01

    This US ITER Management Plan is the plan for conducting the Engineering Design Activities within the US. The plan applies to all design, analyses, and associated physics and technology research and development (R ampersand D) required to support the program. The plan defines the management considerations associated with these activities. The plan also defines the management controls that the project participants will follow to establish, implement, monitor, and report these activities. The activities are to be conducted by the project in accordance with this plan. The plan will be updated to reflect the then-current management approach required to meet the project objectives. The plan will be reviewed at least annually for possible revision. Section 2 presents the ITER objectives, a brief description of the ITER concept as developed during the Conceptual Design Activities, and comments on the Engineering Design Activities. Section 3 discusses the planned international organization for the Engineering Design Activities, from which the tasks will flow to the US Home Team. Section 4 describes the US ITER management organization and responsibilities during the Engineering Design Activities. Section 5 describes the project management and control to be used to perform the assigned tasks during the Engineering Design Activities. Section 6 presents the references. Several appendices are provided that contain detailed information related to the front material

  8. Linear iterative near-field phase retrieval (LIPR) for dual-energy x-ray imaging and material discrimination.

    Science.gov (United States)

    Li, Heyang Thomas; Kingston, Andrew M; Myers, Glenn R; Beeching, Levi; Sheppard, Adrian P

    2018-01-01

    Near-field x-ray refraction (phase) contrast is unavoidable in many lab-based micro-CT imaging systems. Quantitative analysis of x-ray refraction (a.k.a. phase retrieval) is in general an under-constrained problem. Regularizing assumptions may not hold true for interesting samples; popular single-material methods are inappropriate for heterogeneous samples, leading to undesired blurring and/or over-sharpening. In this paper, we constrain and solve the phase-retrieval problem for heterogeneous objects, using the Alvarez-Macovski model for x-ray attenuation. Under this assumption we neglect Rayleigh scattering and pair production, considering only Compton scattering and the photoelectric effect. We formulate and test the resulting method to extract the material properties of density and atomic number from single-distance, dual-energy imaging of both strongly and weakly attenuating multi-material objects with polychromatic x-ray spectra. Simulation and experimental data are used to compare our proposed method with the Paganin single-material phase-retrieval algorithm, and an innovative interpretation of the data-constrained modeling phase-retrieval technique.

  9. Plasma cleaning of ITER first mirrors

    Science.gov (United States)

    Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

    2017-12-01

    Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

  10. A highly efficient silole-containing dithienylethene with excellent thermal stability and fatigue resistance: a promising candidate for optical memory storage materials.

    Science.gov (United States)

    Chan, Jacky Chi-Hung; Lam, Wai Han; Yam, Vivian Wing-Wah

    2014-12-10

    Diarylethene compounds are potential candidates for applications in optical memory storage systems and photoswitchable molecular devices; however, they usually show low photocycloreversion quantum yields, which result in ineffective erasure processes. Here, we present the first highly efficient photochromic silole-containing dithienylethene with excellent thermal stability and fatigue resistance. The photochemical quantum yields for photocyclization and photocycloreversion of the compound are found to be high and comparable to each other; the latter of which is rarely found in diarylethene compounds. These would give rise to highly efficient photoswitchable material with effective writing and erasure processes. Incorporation of the silole moiety as a photochromic dithienylethene backbone also was demonstrated to enhance the thermal stability of the closed form, in which the thermal backward reaction to the open form was found to be negligible even at 100 °C, which leads to a promising candidate for use as photoswitchable materials and optical memory storage.

  11. Structural materials for high-heat flux applications

    International Nuclear Information System (INIS)

    Rybin, V.V.; Smith, D.L.

    1991-01-01

    The structural materials for the ITER, (International Thermonuclear Experimental Reactor) divertor must perform reliably under complex and diverse operating requirements. Only a limited number of materials offer a potential for meeting these requirements for the wide temperature range of interest. The candidate materials considered in the ITER design activity include copper, molybdenum, niobium alloys. Molybdenum alloys being considered include dilute alloys of the TZM type and the Mo-Re system. Niobium alloys under consideration include Nb-V-Zr and Nb-Zr systems. Copper alloys being considered include precipitation strengthened alloys of the Glidcop and MAGT type, alloys of Cu-Mo system and dispersion hardened bronzes. The projected operating conditions for the ITER divertor and the criteria for evaluating the candidate materials are reviewed. This paper summarizes the data base and presents recent experimental results on these candidate divertor structural alloys

  12. ITER council proceedings: 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Continuing the ITER EDA, two further ITER Council Meetings were held since the publication of ITER EDA documentation series no, 20, namely the ITER Council Meeting on 27-28 February 2001 in Toronto, and the ITER Council Meeting on 18-19 July in Vienna. That Meeting was the last one during the ITER EDA. This volume contains records of these Meetings, including: Records of decisions; List of attendees; ITER EDA status report; ITER EDA technical activities report; MAC report and advice; Final report of ITER EDA; and Press release

  13. A study of the photocatalytic effects of aqueous suspensions of platinized semiconductor materials on the reaction rates of candidate redox reactions

    Science.gov (United States)

    Miles, A. M.

    1982-01-01

    The effectiveness of powdered semiconductor materials in photocatalyzing candidate redox reactions was investigated. The rate of the photocatalyzed oxidation of cyanide at platinized TiO2 was studied. The extent of the cyanide reaction was followed directly using an electroanalytical method (i.e. differential pulse polarography). Experiments were performed in natural or artificial light. A comparison was made of kinetic data obtained for photocatalysis at platinized powders with rate data for nonplatinized powders.

  14. Charge, spin and orbital order in the candidate multiferroic material LuFe{sub 2}O{sub 4}

    Energy Technology Data Exchange (ETDEWEB)

    Groot, Joost de

    2012-06-28

    This thesis is a detailed study of the magnetic, structural and orbital order parameters of the candidate multiferroic material LuFe{sub 2}O{sub 4}. Multiferroic oxides with a strong magnetoelectric coupling are of high interest for potential information technology applications, but they are rare because the traditional mechanism of ferroelectricity is incompatible with magnetism. Consequently, much attention is focused on various unconventional mechanisms of ferroelectricity. Of these, ferroelectricity originating from charge ordering (CO) is particularly intriguing because it potentially combines large electric polarizations with strong magneto-electric coupling. However, examples of oxides where this mechanism occurs are exceedingly rare and none is really well understood. LuFe{sub 2}O{sub 4} is often cited as the prototypical example of CO-based ferroelectricity. In this material, the order of Fe valences has been proposed to render the triangular Fe/O bilayers polar by making one of the two layers rich in Fe{sup 2+} and the other rich in Fe{sup 3+}, allowing for a possible ferroelectric stacking of the individual bilayers. Because of this new mechanism for ferroelectricity, and also because of the high transition temperatures of charge order (T{sub CO} {proportional_to}320K) and ferro magnetism (T{sub N}{proportional_to}240 K) LuFe{sub 2}O{sub 4} has recently attracted increasing attention. Although these polar bilayers are generally accepted in the literature for LuFe{sub 2}O{sub 4}, direct proof is lacking. An assumption-free experimental determination of whether or not the CO in the Fe/O bilayers is polar would be crucial, given the dependence of the proposed mechanism of ferroelectricity from CO in LuFe{sub 2}O{sub 4} on polar bilayers. This thesis starts with a detailed characterization of the macroscopic magnetic properties, where growing ferrimagnetic contributions observed in magnetization could be ascribed to increasing oxygen off-stoichiometry. The

  15. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  16. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    Energy Technology Data Exchange (ETDEWEB)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr. [Lawrence Livermore National Lab., CA (United States); Gdowski, G.E. [KMI, Inc., Albuquerque, NM (United States)

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices.

  17. Selection of candidate container materials for the conceptual waste package design for a potential high level nuclear waste repository at Yucca Mountain

    International Nuclear Information System (INIS)

    Van Konynenburg, R.A.; Halsey, W.G.; McCright, R.D.; Clarke, W.L. Jr.; Gdowski, G.E.

    1993-02-01

    Preliminary selection criteria have been developed, peer-reviewed, and applied to a field of 41 candidate materials to choose three alloys for further consideration during the advanced conceptual design phase of waste package development for a potential high level nuclear waste repository at Yucca Mountain, Nevada. These three alloys are titanium grade 12, Alloy C-4, and Alloy 825. These selections are specific to the particular conceptual design outlined in the Site Characterization Plan. Other design concepts that may be considered in the advanced conceptual design phase may favor other materials choices

  18. ITER EDA newsletter. V. 7, no. 5

    International Nuclear Information System (INIS)

    1998-05-01

    This newsletter contains the articles 'The materials selection in ITER and the first materials workshop', 'US fusion community discussion on fusion strategies', 'ITER central solenoid model coil heat treatment complete and assembly started' and 'Programme of the 17th IAEA fusion energy conference'. There is also a note in memoriam of Hiroschi Shibata, who died on the 5th of June 1998

  19. Design of the ITER vacuum vessel

    International Nuclear Information System (INIS)

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

    1995-01-01

    The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement.The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2-2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4-0.7 m.The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200-300 MN due to VDE and to allow for differential thermal expansion.The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology. (orig.)

  20. Quantitative mineralogy and preliminary pore-water chemistry of candidate buffer and backfill materials for a nuclear fuel waste disposal vault

    International Nuclear Information System (INIS)

    Quigley, R.M.

    1984-07-01

    The quantitative mineralogy of seven candidate buffer and backfill materials for a nuclear fuel waste disposal vault is presented. Two of the materials were coarse grained: one a blended very pure silica sand, and the other a crushed plagioclase-rich granite or granodiorite. Five materials were fine-grained soils containing abundant clay minerals. Of these, three were fairly pure, Cretaceous, ash-derived bentonites that contained up to 3 percent of soluble sulphates; one was a freshwater glacial clay containing 59 percent interlayered smectite-illite; and one was a crushed Paleozoic shale containing abundant illite and chlorite. The adsorbed cation regimes and the pore-water chemistry of the clays are discussed

  1. Structural stability at high pressure, electronic, and magnetic properties of BaFZnAs: A new candidate of host material of diluted magnetic semiconductors

    International Nuclear Information System (INIS)

    Chen Bi-Juan; Deng Zheng; Wang Xian-Cheng; Feng Shao-Min; Yuan Zhen; Zhang Si-Jia; Liu Qing-Qing; Jin Chang-Qing

    2016-01-01

    The layered semiconductor BaFZnAs with the tetragonal ZrCuSiAs-type structure has been successfully synthesized. Both the in-situ high-pressure synchrotron x-ray diffraction and the high-pressure Raman scattering measurements demonstrate that the structure of BaFZnAs is stable under pressure up to 17.5 GPa at room temperature. The resistivity and the magnetic susceptibility data show that BaFZnAs is a non-magnetic semiconductor. BaFZnAs is recommended as a candidate of the host material of diluted magnetic semiconductor. (special topic)

  2. Testing of Candidate Polymeric Materials for Compatibility with Pure Alternate Pretreat as Part of the Universal Waste Management System (UWMS)

    Science.gov (United States)

    Wingard, C. D.

    2018-01-01

    The Universal Waste Management System (UWMS) is an improved Waste Collection System for astronauts living and working in low Earth orbit spacecraft. Polymeric materials used in water recovery on International Space Station are regularly exposed to phosphoric acid-treated 'pretreated' urine. Polymeric materials used in UWMS are not only exposed to pretreated urine, but also to concentrated phosphoric acid with oxidizer before dilution known as 'pure pretreat.' Samples of five different polymeric materials immersed in pure pretreat for 1 year were tested for liquid compatibility by measuring changes in storage modulus with a dynamic mechanical analyzer.

  3. Down-selection of candidate alloys for further testing of advanced replacement materials for LWR core internals

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States). Applied Physics Program; Leonard, Keith J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Tan, Lizhen [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-09-01

    Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) Light Water Reactor Sustainability Program to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to identify and develop advanced alloys with superior degradation resistance in light water reactor (LWR)-relevant environments by 2024.

  4. ITER council proceedings: 1998

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains documents of the 13th and the 14th ITER council meeting as well as of the 1st extraordinary ITER council meeting. Documents of the ITER meetings held in Vienna and Yokohama during 1998 are also included. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics are given

  5. ITER diagnostic system: Vacuum interface

    International Nuclear Information System (INIS)

    Patel, K.M.; Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L.; Drevon, J.M.; Encheva, A.; Kashchuk, Y.; Maquet, Ph.; Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J.

    2013-01-01

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10 −7 Pa, irrespective of plasma operation, and a leak rate of less than 10 −10 Pa m 3 s −1 . In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described

  6. ITER diagnostic system: Vacuum interface

    Energy Technology Data Exchange (ETDEWEB)

    Patel, K.M., E-mail: Kaushal.Patel@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Udintsev, V.S.; Hughes, S.; Walker, C.I.; Andrew, P.; Barnsley, R.; Bertalot, L. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Drevon, J.M. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Encheva, A. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France); Kashchuk, Y. [Institution “PROJECT CENTER ITER”, 1, Akademika Kurchatova pl., Moscow (Russian Federation); Maquet, Ph. [Bertin Technologies, BP 22, 13762 Aix-en Provence cedex 3 (France); Pearce, R.; Taylor, N.; Vayakis, G.; Walsh, M.J. [ITER Organization, Route de Vinon sur Verdon, 13115 St Paul-Lez-Durance (France)

    2013-10-15

    Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10{sup −7} Pa, irrespective of plasma operation, and a leak rate of less than 10{sup −10} Pa m{sup 3} s{sup −1}. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions

  7. ITER Council proceedings: 1993

    International Nuclear Information System (INIS)

    1994-01-01

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ''ITER EDA Agreement and Protocol 2'' (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tasks; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings

  8. ITER council proceedings: 2000

    International Nuclear Information System (INIS)

    2001-01-01

    No ITER Council Meetings were held during 2000. However, two ITER EDA Meetings were held, one in Tokyo, January 19-20, and one in Moscow, June 29-30. The parties participating in these meetings were those that partake in the extended ITER EDA, namely the EU, the Russian Federation, and Japan. This document contains, a/o, the records of these meetings, the list of attendees, the agenda, the ITER EDA Status Reports issued during these meetings, the TAC (Technical Advisory Committee) reports and recommendations, the MAC Reports and Advice (also for the July 1999 Meeting), the ITER-FEAT Outline Design Report, the TAC Reports and Recommendations both meetings), Site requirements and Site Design Assumptions, the Tentative Sequence of technical Activities 2000-2001, Report of the ITER SWG-P2 on Joint Implementation of ITER, EU/ITER Canada Proposal for New ITER Identification

  9. ITER council proceedings: 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ``ITER EDA Agreement and Protocol 2`` (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tastes; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings.

  10. Investigation on fabrication of SiC/SiC composite as a candidate material for fuel sub-assembly

    International Nuclear Information System (INIS)

    Lee, Jae-Kwang; Naganuma, Masayuki; Park, Joon-Soo; Kohyama, Akira

    2005-01-01

    The possibility of SiC/SiC (Silicon carbide fiber reinforced Silicon carbide) composites application for fuel sub-assembly of Fast Breeder Reactor was investigated. To select a raw material of SiC/SiC composites, a few kinds of SiC nano powder was estimated by SEM observation and XRD analysis. Furthermore, SiC monolithic was sintered from them and estimated by flexural test. SiC nano-powder which showed good sinterability, it was used for fabrication of SiC/SiC composites by Hot Pressing method. From the sintering condition of 1800, 1820degC temperature and 15, 20 MPa pressure, SiC/SiC composite was fabricated and then estimated by tensile test. SiC/SiC composite, which made by 1820degC and 20 MPa condition, showed the highest mechanical strength by the monotonic tensile test. SiC/SiC composite, which made by 1800degC and 15 MPa condition, showed a stable fracture behavior at the monotonic and cyclic tensile test. And then, the hoop stress of ideal model of SiC/SiC composites was discussed. It was confirmed that applicability of SiC/SiC composites by Hot Pressing method for fuel sub-assembly structural material. To make it real attractive one, to maintain the reliability and safety as a high temperature structural material, the design and process study on SiC/Sic composites material will be continued. (author)

  11. Experimental demonstration of radiation effects on the performance of a stirling-alternator convertor and candidate materials evaluation

    Science.gov (United States)

    Mireles, Omar R.

    Free-piston Stirling power convertors are under consideration by NASA for service in the Advanced Stirling Radioisotope Generator (ASRG) and Fission Surface Power (FSP) systems to enable aggressive exploration missions by providing a reliable and constant power supply. The ASRG must withstand environmental radiation conditions, while the FSP system must tolerate a mixed neutron and gamma-ray environment resulting from self-irradiation. Stirling-alternators utilize rare earth magnets and a variety of organic materials whose radiation limits dominate service life estimates and shielding requirements. The project objective was to demonstrate the performance of the alternator, identify materials that exhibit excessive radiation sensitivity, identify radiation tolerant substitutes, establish empirical dose limits, and demonstrate the feasibility of cost effective nuclear and radiation tests by selection of the appropriate personnel and test facilities as a function of hardware maturity. The Stirling Alternator Radiation Test Article (SARTA) was constructed from linear alternator components of a Stirling convertor and underwent significant pre-exposure characterization. The SARTA was operated at the Sandia National Laboratories Gamma Irradiation Facility to a dose of over 40 Mrad. Operating performance was within nominal variation, although modestly decreasing trends occurred in later runs as well as the detection of an electrical fault after the final exposure. Post-irradiation disassembly and internal inspection revealed minimal degradation of the majority of the organic components. Radiation testing of organic material coupons was conducted since the majority of the literature was inconsistent. These inconsistencies can be attributed to testing at environmental conditions vastly different than those Stirling-alternator organics will experience during operation. Samples were irradiated at the Texas A&M TRIGA reactor to above expected FSP neutron fluence. A thorough

  12. Characterize and Homogeneity Test of Phthalate and Phosphate Buffer as Reference Materials Candidate for Determination Ofacidic Value

    OpenAIRE

    Sujarwo, Sujarwo; Nuryatini, Nuryatini

    2013-01-01

    Measurement of pH is closely related to everyday life, such as in the area of environment, food and beverage, pharmaceuticals, and even metabolism process in human body. To ensure the quality of pH measurement, pH metershould be calibrated regularly and checked each time before used. Calibration and performance checking of pH meter requires traceable standard. The standard should be linked to CRM (Certified Reference Materials). Research Center for Chemistry - Indonesian Institute of Sciences...

  13. Effect of Eu magnetism on the electronic properties of the candidate Dirac material EuMnBi2

    Science.gov (United States)

    May, Andrew F.; McGuire, Michael A.; Sales, Brian C.

    2014-08-01

    The crystal structure and physical properties of the layered material EuMnBi2 have been characterized by measurements on single crystals. EuMnBi2 is isostructural with the Dirac material SrMnBi2 based on single-crystal x-ray diffraction, crystallizing in the I4/mmm space group (No. 139). Magnetic susceptibility measurements suggest antiferromagnetic (AFM) ordering of moments on divalent Eu ions near TN=22 K. For low fields, the ordered Eu moments are aligned along the c axis, and a spin flop is observed near 5.4 T at 5 K. The moment is not saturated in an applied field of 13 T at 5 K, which is uncommon for compounds containing Eu2+. The magnetic behavior suggests an anisotropy enhancement via interaction between Eu and the Mn moments that appear to be ordered antiferromagnetically below ≈310 K. A large increase in the magnetoresistance is observed across the spin flop, with absolute magnetoresistance reaching ≈650% at 5 K and 12 T. Hall effect measurements reveal a decrease in the carrier density below TN, which implies a manipulation of the Fermi surface by magnetism on the sites surrounding the Bi square nets that lead to Dirac cones in this family of materials.

  14. Consistency Study About Critical Thinking Skill of PGSD Students (Teacher Candidate of Elementary School) on Energy Material

    Science.gov (United States)

    Wijayanti, M. D.; Raharjo, S. B.; Saputro, S.; Mulyani, S.

    2017-09-01

    This study aims to examine the consistency of critical thinking ability of PGSD students in Energy material. The study population is PGSD students in UNS Surakarta. Samples are using cluster random sampling technique obtained by 101 students. Consistency of student’s response in knowing the critical thinking ability of PGSD students can be used as a benchmark of PGSD students’ understanding to see the equivalence of IPA problem, especially in energy material presented with various phenomena. This research uses descriptive method. Data are obtained through questionnaires and interviews. The research results that the average level of critical thinking in this study is divided into 3 levels, i.e.: level 1 (54.85%), level 2 (19.93%), and level 3 (25.23%). The data of the research result affect to the weak of students’ Energy materials’ understanding. In addition, indicators identify that assumptions and arguments analysis are also still low. Ideally, the consistency of critical thinking ability as a whole has an impact on the expansion of students’ conceptual understanding. The results of the study may become a reference to improve the subsequent research in order to obtain positive changes in the ability of critical thinking of students who directly improve the concept of students’ better understanding, especially in energy materials at various real problems occured.

  15. Supplementary Material for: In silico screening for candidate chassis strains of free fatty acid-producing cyanobacteria

    KAUST Repository

    Motwalli, Olaa Amin

    2017-01-01

    Abstract Background Finding a source from which high-energy-density biofuels can be derived at an industrial scale has become an urgent challenge for renewable energy production. Some microorganisms can produce free fatty acids (FFA) as precursors towards such high-energy-density biofuels. In particular, photosynthetic cyanobacteria are capable of directly converting carbon dioxide into FFA. However, current engineered strains need several rounds of engineering to reach the level of production of FFA to be commercially viable; thus new chassis strains that require less engineering are needed. Although more than 120 cyanobacterial genomes are sequenced, the natural potential of these strains for FFA production and excretion has not been systematically estimated. Results Here we present the FFA SC (FFASC), an in silico screening method that evaluates the potential for FFA production and excretion of cyanobacterial strains based on their proteomes. A literature search allowed for the compilation of 64 proteins, most of which influence FFA production and a few of which affect FFA excretion. The proteins are classified into 49 orthologous groups (OGs) that helped create rules used in the scoring/ranking of algorithms developed to estimate the potential for FFA production and excretion of an organism. Among 125 cyanobacterial strains, FFASC identified 20 candidate chassis strains that rank in their FFA producing and excreting potential above the specifically engineered reference strain, Synechococcus sp. PCC 7002. We further show that the top ranked cyanobacterial strains are unicellular and primarily include Prochlorococcus (order Prochlorales) and marine Synechococcus (order Chroococcales) that cluster phylogenetically. Moreover, two principal categories of enzymes were shown to influence FFA production the most: those ensuring precursor availability for the biosynthesis of lipids, and those involved in handling the oxidative stress associated to FFA synthesis

  16. Comparison of the leading candidate combinations of blanket materials, thermodynamic cycles, and tritium systems for full scale fusion power plants

    International Nuclear Information System (INIS)

    Fraas, A.P.

    1975-01-01

    The many possible combinations of blanket materials, tritium generation and recovery systems, and power conversion systems were surveyed and a comprehensive set of designs were generated by using a common set of ground rules that include all of the boundary conditions that could be envisioned for a full-scale commercial fusion power plant. Particular attention was given to the effects of blanket temperature on power plant cycle efficiency and economics, the interdependence of the thermodynamic cycle and the tritium recovery system, and to thermal and pressure stresses in the blanket structure. The results indicate that, of the wide variety of systems that have been considered, the most promising employs lithium recirculated in a closed loop within a niobium blanket structure and cooled with boiling potassium or cesium. This approach gives the simplest and lowest cost tritium recovery system, the lowest pressure and thermal stresses, the simplest structure with the lowest probability of a leak, the greatest resistance to damage from a plasma energy dump, and the lowest rate of plasma contamination by either outgassing or sputtering. The only other blanket materials combination that appears fairly likely to give a satisfactory tritium generation and recovery system is a lithium-beryllium fluoride-Incoloy blanket, and even this system involves major uncertainties in the effectiveness, size, and cost of the tritium recovery system. Further, the Li 2 BeF 4 blanket system has the disadvantage that the world reserves of beryllium are too limited to support a full-blown fusion reactor economy, its poor thermal conductivity leads to cooling difficulties and a requirement for a complex structure with intricate cooling passages, and this inherently leads to an expansive blanket with a relatively high probability of leaks. The other blanket materials combinations yield even less attractive systems

  17. Methods to estimate equipment and materials that are candidates for removal during the decontamination of fuel processing facilities

    International Nuclear Information System (INIS)

    Duncan, D.R.; Valero, O.J.; Hyre, R.A.; Pottmeyer, J.A.; Millar, J.S.; Reddick, J.A.

    1995-02-01

    The methodology presented in this report provides a model for estimating the volume and types of waste expected from the removal of equipment and other materials during Decontamination and Decommissioning (D and D) of canyon-type fuel reprocessing facilities. This methodology offers a rough estimation technique based on a comparative analysis for a similar, previously studied, reprocessing facility. This approach is especially useful as a planning tool to save time and money while preparing for final D and D. The basic methodology described here can be extended for use at other types of facilities, such as glovebox or reactor facilities

  18. CHARACTERISTIC OF MECHANICS TEACHING MATERIALS FOR INCREASING STUDENTS OF PHYSICS TEACHER CANDIDATES REPRESENTATION ABILITY ON VERBAL, MATHEMATICAL, PICTURE, AND GRAPHIC

    Directory of Open Access Journals (Sweden)

    I Ketut Mahardika

    2014-01-01

    Full Text Available Penelitian ini adalah tentang karakteristik pengajaran mekanik untuk meningkatkan kemampuan representasi mahasiswa calon guru fisika pada verbal, matematika, gambar, dan grafis. Masalah penelitian adalah bagaimana karakteristik bahan ajar mekanik dapat membantu mahasiswa calon guru fisika dalam meningkatkan kemampuan representasi mereka pada verbal, matematika, gambar, dan grafis. Masalah penelitian ini diterjemahkan ke dalam dua pertanyaan. Pertama, bagaimana bahan ajar mekanik terdiri dari contoh aplikasi prinsip mekanika dalam gambar peristiwa fisika? Kedua, bagaimana pola penulisan bahan ajar mekanik bagi mahasiswa calon guru fisika? Penelitian ini merupakan pengembangan bahan ajar mekanika, menggunakan metode campuran model investigasi, yaitu: pengumpulan data kualitatif, membangun pengajaran mekanik, rancangan materi sehingga akan siap untuk divalidasi. Subjek penelitian diharapkan mampu meningkatkan kemampuan representasi pada verbal, matematika, gambar, dan grafis. Responden penelitian ini adalah mahasiswa fisika calon guru yang mengambil subjek fisika dasar fakultas keguruan dan pendidikan, Universitas Jember. Adapun teknik pengumpulan data adalah: observasi, angket, tes dan dokumentasi. Data dijabarkan secara kualitatif. Hasil penelitian menunjukkan bahwa bahan ajar mekanik terdiri dari contoh aplikasi prinsip mekanika dalam gambar acara fisika dan memiliki pola urutan bab, judul sub bab, deskripsi materi, contoh, uji kemampuan dan bibliografi. Kata Kunci: bahan pengajaran mekanik, calon guru fisika, kemampuan representasi pada verbal, matematika, gambar, dan grafik

  19. Degradation mode survey candidate titanium-base alloys for Yucca Mountain project waste package materials. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Gdowski, G.E.

    1997-12-01

    The Yucca Mountain Site Characterization Project (YMP) is evaluating materials from which to fabricate high-level nuclear waste containers (hereafter called waste packages) for the potential repository at Yucca Mountain, Nevada. Because of their very good corrosion resistance in aqueous environments titanium alloys are considered for container materials. Consideration of titanium alloys is understandable since about one-third (in 1978) of all titanium produced is used in applications where corrosion resistance is of primary importance. Consequently, there is a considerable amount of data which demonstrates that titanium alloys, in general, but particularly the commercial purity and dilute {alpha} grades, are highly corrosion resistant. This report will discuss the corrosion characteristics of Ti Gr 2, 7, 12, and 16. The more highly alloyed titanium alloys which were developed by adding a small Pd content to higher strength Ti alloys in order to give them better corrosion resistance will not be considered in this report. These alloys are all two phase ({alpha} and {beta}) alloys. The palladium addition while making these alloys more corrosion resistant does not give them the corrosion resistance of the single phase {alpha} and near-{alpha} (Ti Gr 12) alloys.

  20. ITER council proceedings: 1995

    International Nuclear Information System (INIS)

    1996-01-01

    Records of the 8. ITER Council Meeting (IC-8), held on 26-27 July 1995, in San Diego, USA, and the 9. ITER Council Meeting (IC-9) held on 12-13 December 1995, in Garching, Germany, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the ITER Interim Design Report Package and Relevant Documents. Figs, tabs

  1. Validation of a liquid chromatographic method for determination of related substances in a candidate certified reference material of captopril

    Directory of Open Access Journals (Sweden)

    Raquel Nogueira

    2011-06-01

    Full Text Available This paper describes the validation of a reversed-phase high performance liquid chromatography method (RP-HPLC with diode array detection (DAD for determination of related substances (impurities from organic synthesis and degradation products of captopril according to the Brazilian Pharmacopeia IV. The aim of this study was to guarantee the method accuracy for quantification of related substances, an essential requisite to determine, using the mass balance approach, the captopril content in the first Brazilian certified reference material (CRM of an active pharmaceutical ingredient (API, developed by Inmetro. The captopril instability in solution is discussed and the captopril content determined by mass balance is compared to the results from titration and differential scanning calorimetry (DSC.Este artigo descreve a validação de método de cromatografia líquida de alta eficiência em fase reversa (CLAE-RP com detector de fotodiodos (DAD para determinação de substâncias relacionadas (impurezas orgânicas de síntese e produtos de degradação de captopril segundo Farmacopéia Brasileira IV ed. Este estudo teve como objetivo garantir que o método é capaz de quantificar com exatidão o teor de substâncias relacionadas, um requisito essencial para que o teor de captopril seja determinado por balanço de massa no primeiro material de referência certificado (MRC de fármacos brasileiro, o qual foi desenvolvido pelo Inmetro. A instabilidade do captopril em solução é discutida em detalhes e o teor de captopril determinado por balanço de massa é comparado com aqueles obtidos por titulação e por calorimetria exploratória diferencial (DSC.

  2. ITER council proceedings: 1999

    International Nuclear Information System (INIS)

    1999-01-01

    In 1999 the ITER meeting in Cadarache (10-11 March 1999) and the Programme Directors Meeting in Grenoble (28-29 July 1999) took place. Both meetings were exclusively devoted to ITER engineering design activities and their agendas covered all issues important for the development of ITER. This volume presents the documents of these two important meetings

  3. ITER council proceedings: 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Records of the 10. ITER Council Meeting (IC-10), held on 26-27 July 1996, in St. Petersburg, Russia, and the 11. ITER Council Meeting (IC-11) held on 17-18 December 1996, in Tokyo, Japan, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the cost review and safety analysis. Figs, tabs

  4. ITER EDA technical activities

    International Nuclear Information System (INIS)

    Aymar, R.

    1998-01-01

    Six years of technical work under the ITER EDA Agreement have resulted in a design which constitutes a complete description of the ITER device and of its auxiliary systems and facilities. The ITER Council commented that the Final Design Report provides the first comprehensive design of a fusion reactor based on well established physics and technology

  5. ITER radio frequency systems

    International Nuclear Information System (INIS)

    Bosia, G.

    1998-01-01

    Neutral Beam Injection and RF heating are two of the methods for heating and current drive in ITER. The three ITER RF systems, which have been developed during the EDA, offer several complementary services and are able to fulfil ITER operational requirements

  6. People's Republic of China joins ITER

    International Nuclear Information System (INIS)

    Huo Yuping

    2003-01-01

    The People's Republic of China is the largest developing country with a projected population of 1.6 - 2 billion people and an energy consumption growing from the current 1.3 Billion Tons Coal Equivalent (TCE) to more than 4 Billion TCE by 2050. This large demand needs to be accommodated in a sustainable way, requiring energy generation in an environmentally friendly way. Fusion is one of the most promising candidates to solve this important issue. This explains why in the second half of 2002, the ITER Participants' delegations to the ITER Negotiations received expression of interest from the People's Republic of China in the possibility of Chinese participation in ITER, including joining the ongoing Negotiations. The speed with which the Chinese authorities had made their decision to participate in the ITER Negotiations was impressive. The Prime Minister and the State Council had already confirmed their decision to apply to join ITER as soon as possible, and Mr. Xu Guanhua, Chinese Minister of Science and Technology, wrote on behalf of his government, on 10 January 2003, to the four heads of delegation in the ITER Negotiations, requesting that China participate in the present ITER Negotiations, pointing out that China intends to provide a substantial contribution to the Project, comparable to what is currently envisaged by some of the participants in the present Negotiations

  7. Development of core sampling technique for ITER Type B radwaste

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. G.; Hong, K. P.; Oh, W. H.; Park, M. C.; Jung, S. H.; Ahn, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Type B radwaste (intermediate level and long lived radioactive waste) imported from ITER vacuum vessel are to be treated and stored in basement of hot cell building. The Type B radwaste treatment process is composed of buffer storage, cutting, sampling/tritium measurement, tritium removal, characterization, pre-packaging, inspection/decontamination, and storage etc. The cut slices of Type B radwaste components generated from cutting process undergo sampling process before and after tritium removal process. The purpose of sampling is to obtain small pieces of samples in order to investigate the tritium content and concentration of Type B radwaste. Core sampling, which is the candidates of sampling technique to be applied to ITER hot cell, is available for not thick (less than 50 mm) metal without use of coolant. Experimented materials were SS316L and CuCrZr in order to simulate ITER Type B radwaste. In core sampling, substantial secondary wastes from cutting chips will be produced unavoidably. Thus, core sampling machine will have to be equipped with disposal system such as suction equipment. Core sampling is considered an unfavorable method for tool wear compared to conventional drilling.

  8. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER; Hacia la demostracion de capacidad de produccion industrial del acero ferritico-martensitico espanol ASTURFER a escala de demanda ITER

    Energy Technology Data Exchange (ETDEWEB)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-07-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

  9. Fe-Cluster Compounds of Chalcogenides: Candidates for Rare-Earth-Free Permanent Magnet and Magnetic Nodal-Line Topological Material.

    Science.gov (United States)

    Zhao, Xin; Wang, Cai-Zhuang; Kim, Minsung; Ho, Kai-Ming

    2017-12-04

    Fe-cluster-based crystal structures are predicted for chalcogenides Fe 3 X 4 (X = S, Se, Te) using an adaptive genetic algorithm. Topologically different from the well-studied layered structures of iron chalcogenides, the newly predicted structures consist of Fe clusters that are either separated by the chalcogen atoms or connected via sharing of the vertex Fe atoms. Using first-principles calculations, we demonstrate that these structures have competitive or even lower formation energies than the experimentally synthesized Fe 3 X 4 compounds and exhibit interesting magnetic and electronic properties. In particular, we show that Fe 3 Te 4 can be a good candidate as a rare-earth-free permanent magnet and Fe 3 S 4 can be a magnetic nodal-line topological material.

  10. Development of a Equipment to Measure Gas Transport Properties: Application to Study Mixtures of Candidates Buffer Materials for Low-Medium Level Waste Repositories

    Energy Technology Data Exchange (ETDEWEB)

    Martin, P.L.; Barcala, J.M.; Oller, J.C.

    2002-07-01

    This report describes the design, the construction and a testing of a system set-up for the measurement of gas transport, created at CIEMAT, and its application to study mixtures of candidate buffer materials for Low-Medium Level Waste Repertories. The measure of the gas flows is carried on by mass flow meters of several ranges, white the pressure of the applied within the sample is controlled. Two National l Instrument's acquisition system that permits the control and recording of the parameters. A specific application developed for this test, with National Instruments LabWIEW DSC, permits to mange the system. A client interface lets to follow the experiment course from a remote location through Internet. (Author) 21 refs.

  11. Assessment of homogeneity of candidate reference material at the nanogram level and investigation on representativeness of single particle analysis using electron probe X ray microanalysis

    International Nuclear Information System (INIS)

    Ro, Chul-Un; Hoornaerta, S.; Griekena, R. van

    2002-01-01

    Particulate samples of a candidate reference material are evaluated on their homogeneity from bottle to bottle using electron probe X ray microanalysis technique. The evaluation on the homogeneity is done by the utilization of the Kolmogorov-Smirnov statistics to the processing of the quantitative electron probe X ray microanalysis data. Due to a limitation, existing even in computer controlled electron probe X ray microanalysis, in terms of analysis time and expenses, the number of particles analyzed is much smaller compared to that in the sample. Therefore, it is investigated whether this technique provides representative analysis results for the characteristics of the sample, even though a very small portion of the sample is really analyzed. Furthermore, the required number of particles for the analysis, to insure a certain level of reproducibility, e.g. 5% relative standard deviation, is determined by the application of the Ingamells sampling theory. (author)

  12. The mechanical properties of fluoride salts at elevated temperatures. [candidate thermal energy storage materials for solar dynamic systems

    Science.gov (United States)

    Raj, S. V.; Whittenberger, J. D.

    1989-01-01

    The deformation behavior of CaF2 and LiF single crystals compressed in the 111 and the 100 line directions, respectively, are compared with the mechanical properties of polycrystalline LiF-22 (mol pct) CaF2 eutectic mixture in the temperature range 300 to 1275 K for strain rates varying between 7 x 10 to the -7th and 0.2/s. The true stress-strain curves for the single crystals were found to exhibit three stages in an intermediate range of temperatures and strain rates, whereas those for the eutectic showed negative work-hardening rates after a maximum stress. The true stress-strain rate data for CaF2 and LiF-22 CaF2 could be represented by a power-law relation with the strain rate sensitivities lying between 0.05 and 0.2 for both materials. A similar relation was found to be unsatisfactory in the case of LiF.

  13. Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in U.S

    International Nuclear Information System (INIS)

    Farmer, J.C.; McCright, R.D.

    1989-01-01

    Three ion-based to nickel-based austenitic alloys and three copper-based alloys are being considered in the United States as candidate materials for the fabrication of high-level radioactive waste containers. The austenitic alloys are Types 304L and 316L stainless steels as well as the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper) CDA 613 (Cu7Al), and CDA 715 (Cu-30Ni). Waste in the forms of spent fuel assemblies from reactors and borosilicate glass will be sent to a proposed repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and in gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys

  14. Temperature Effects on the Mechanical Properties of Candidate SNS Target Container Materials after Proton and Neutron Irradiation; TOPICAL

    International Nuclear Information System (INIS)

    Byun, T.S.

    2001-01-01

    This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54 to 2.53 dpa. Irradiation temperatures were in the range 30 to 100 C. Tensile testing was performed at room temperature (20 C) and 164 C to study the effects of test temperature on the tensile properties. Test materials displayed significant radiation-induced hardening and loss of ductility due to irradiation. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative strain hardening. In the EC316LN stainless steel, increasing the test temperature from 20 C to 164 C decreased the strength by 13 to 18% and the ductility by 8 to 36%. The tensile data for the EC316LN stainless steel irradiated in spallation conditions were in line with the values in a database for 316 stainless steels for doses up to 1 dpa irradiated in fission reactors at temperatures below 200 C. However, extra strengthening induced by helium and hydrogen contents is evident in some specimens irradiated to above about 1 dpa. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. It was estimated that the 316 stainless steels would retain more than 1% true stains to necking at 164 C after irradiation to 5 dpa. A calculation using reduction of area (RA) measurements and stress-strain data predicted positive strain hardening during plastic instability

  15. It was the demonstration of industrial steel production capacity ferritic-martensitic Spanish ASTURFER scale demand ITER

    International Nuclear Information System (INIS)

    Coto, R.; Serrano, M.; Moran, A.; Rodriguez, D.; Artimez, J. A.; Belzunce, J.; Sedano, L.

    2013-01-01

    Reduced Activation Ferritic-Martensitic (RAFM) structural steels are considered as candidate materials with notable possibilities to be incorporated to fusion reactor ITER, nowadays under construction, and future fusion reactor DEMO, involving a notable forecasting of supply materials, with a considerable limitation due to the few number of furnishes currently on the market. The manufacture at an industrial scale of the ASTURFER steel, developed at laboratory scale by ITMA Materials Technology and the Structural Materials Division of the Technology Division of CIEMAT would be a significant business opportunity for steelwork companies.

  16. ITER-FEAT safety

    International Nuclear Information System (INIS)

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  17. ITER council proceedings: 1997

    International Nuclear Information System (INIS)

    1997-01-01

    This volume of the ITER EDA Documentation Series presents records of the 12th ITER Council Meeting, IC-12, which took place on 23-24 July, 1997 in Tampere, Finland. The Council received from the Parties (EU, Japan, Russia, US) positive responses on the Detailed Design Report. The Parties stated their willingness to contribute to fulfil their obligations in contributing to the ITER EDA. The summary discussions among the Parties led to the consensus that in July 1998 the ITER activities should proceed for additional three years with a general intent to enable an efficient start of possible, future ITER construction

  18. Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

    1991-01-01

    An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

  19. ITER EDA newsletter. V. 2, Nos. 7/8

    International Nuclear Information System (INIS)

    1993-01-01

    This ITER EDA (Engineering Design Activities) Newsletter issue includes a description of the ITER Design Integration Division, and reports on the 5th IAEA Technical Committee Meeting on Developments in Fusion Safety held in Toronto, Canada, 7 - 11 June 1993, and on the International Atomic Energy Agency's Atomic and Plasma-Material Interaction Data Activities in support of the ITER Engineering Design Activities

  20. Reliability of activation cross sections for estimation of shutdown dose rate in the ITER port cell and port interspace

    Science.gov (United States)

    García, Raquel; García, Mauricio; Ogando, Francisco; Pampin, Raúl; Sanz, Javier

    2017-09-01

    This paper explores the quality of available activation cross section (XS) data for accurate Shutdown Dose Rate (SDDR) prediction in the ITER Port Cell and Port Interspace areas, where different maintenance activities are foreseen. For this purpose the EAF library (2007 and 2010 versions) has been investigated, as it is typically used by the ITER community. Based on both reports/papers on SDDR in ITER and own calculations, major nuclides contributing to the SDDR coming from the activation of i) relevant materials placed in ITER and ii) candidate materials for the bioshield plug as L2N and barite concretes, are identified. Then, relevant production pathways are obtained. EAF XS quality for all pathways is checked following the procedure used for validating and testing the successive EAF versions. Also, possible improvements from using the TENDL-2015 library are assessed by comparing EAF and TENDL XS with available differential experimental data from EXFOR. Results point out that most of the activation XS related to materials currently placed in ITER are reliable, and only a few need improvement. Also, many of the XS related to both L2N and barite concretes need further work for validation.

  1. New Ablation-Resistant Material Candidate for Hypersonic Applications: Synthesis, Composition, and Oxidation Resistance of HfIr3-Based Solid Solution.

    Science.gov (United States)

    Lozanov, Victor V; Baklanova, Natalya I; Bulina, Natalia V; Titov, Anatoly T

    2018-04-18

    The peculiarities of the solid-state interaction in the HfC-Ir system have been studied within the 1000-1600 °C temperature range using a set of modern analytical techniques. It was stated that the interaction of HfC with iridium becomes noticeable at temperatures as low as 1000-1100 °C and results in the formation of HfIr 3 -based substitutional solid solution. The homogeneity range of the HfIr 3± x phase was evaluated and refined as HfIr 2.43 -HfIr 3.36 . The durability of the HfIr 3 -based system under extreme environmental conditions was studied. It was shown that the HfIr 3 -based material displays excellent ablation resistance under extreme environmental conditions. The benefits of the new designed material result from its relative oxygen impermeability and special microstructure similar to superalloys. The results obtained in this work allow us to consider HfIr 3 as a very promising candidate for extreme applications.

  2. Synthesis and characterization of polymer matrix composite material with combination of ZnO filler and nata de coco fiber as a candidate of semiconductor material

    Science.gov (United States)

    Saputra, Asep Handaya; Anindita, Hana Nabila

    2015-12-01

    Synthesis of semiconductor composite using acrylic matrix filled with ZnO and nata de coco fiber has been conducted in this research. The purpose of this research is to obtain semiconductor composite material that has a good mechanical strength and thermal resistance. In situ polymerization method is used in this research and the composites are ready to be characterized after 12 hours. The main parameter that is characterized is the electric conductivity of the composite. Additional parameters are also characterized such as composite's elastic modulus and glass transition temperature. The composites that has been made in this research can be classified as semiconductor material because the conductivity is in the range of 10-8-103 S/cm. In general the addition of ZnO and nata de coco filler can increase the conductivity of the composite. The highest semiconductor characteristic in acrylic/ZnO composite is obtained from 30% volume filler that reach 3.4 x 10-7 S/cm. Similar with acrylic/ZnO composite, in acrylic/nata de coco fiber composite the highest semiconductor characteristic is also obtained from 30% volume filler that reach 1.15 x 10-7 S/cm. Combination of 20% volume of ZnO, 10% volume of nata de coco, and 70% volume of acrylic resulting in composite with electric conductivity of 1.92 x 10-7 S/cm. In addition, combination of ZnO and nata de coco fiber as filler in composite can also improve the characteristic of composite where composite with 20% volume of ZnO filler and 10% volume of nata de coco fiber resulting in composite with elastic modulus of 1.79 GPa and glass transition temperature of 175.73°C which is higher than those in acrylic/ZnO composite.

  3. Advanced scenarios for ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2004-07-01

    In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

  4. ITER EDA newsletter. V. 4, no. 11

    International Nuclear Information System (INIS)

    1995-11-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains a report on the Ninth Meeting of the ITER Management Advisory Committee held in St. Petersburg, Russia, on November 3, 1995; a report on the Seventh International Conference on Fusion Reactor Materials held at Obninsk, Russia, 25-29 September, 1995; on the presentation of the ITER Project during a symposium on fusion energy held at Champaign, Illinois, USA, October 1-5, 1995; and on two meetings on ITER diagnostics, i.e., an international workshop on diagnostics for ITER held in Varenna, Italy, 28 August - 1 September, 1995; followed by the Third Diagnostics Expert Group Workshop held September 4-5 in the same location

  5. ITER site selection studies in Spain

    International Nuclear Information System (INIS)

    Medrano, M.; Alejaldre, C.; Doncel, J.; Garcia, A.; Ibarra, A.; Jimenez, J.A.; Sanchez de Mora, M.A.; Alcala, F.; Diez, J.E.; Dominguez, M.; Albisu, F.

    2003-01-01

    The studies carried out to evaluate and select a candidate site for International Thermonuclear Experimental Reactor (ITER) construction in Spain are presented in this paper. The ITER design, completed in July 2001, considered a number of technical requirements that must be fulfilled by the selected site. Several assumptions concerning the ITER site were made in order to carry on the design before final site selection. In the studies undertaken for ITER site selection in Spain, the referred technical requirements and assumptions were applied across the whole of Spain and two areas were identified as being preferential. These areas are on the Mediterranean coast and are situated in the Catalan and Valencian regions. A comparative evaluation based on technical characteristics for the concrete plots, proposed within the preferential areas, has been done. The result of these studies was the selection of a site that was deemed to be the most competitive--Vandellos (Tarragona)--and it was proposed to the European Commission for detailed studies in order to be considered as a possible European site for ITER construction. Another key factor for hosting ITER in Spain, is the licensing process. The present status is summarised in this paper

  6. Iteration and accelerator dynamics

    International Nuclear Information System (INIS)

    Peggs, S.

    1987-10-01

    Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

  7. US ITER limiter module design

    International Nuclear Information System (INIS)

    Mattas, R.F.; Billone, M.; Hassanein, A.

    1996-08-01

    The recent U.S. effort on the ITER (International Thermonuclear Experimental Reactor) shield has been focused on the limiter module design. This is a multi-disciplinary effort that covers design layout, fabrication, thermal hydraulics, materials evaluation, thermo- mechanical response, and predicted response during off-normal events. The results of design analyses are presented. Conclusions and recommendations are also presented concerning, the capability of the limiter modules to meet performance goals and to be fabricated within design specifications using existing technology

  8. Future plan of ITER

    International Nuclear Information System (INIS)

    Kitsunezaki, Akio

    1998-01-01

    In cooperation of four countries, Japan, USA, EU and Russia, ITER plan has been proceeding as ''the conceptual design activities'' from 1988 to 1990 and ''the industrial design activities'' since 1992. To construct ITER, the legal and work side of ITER operation has been investigated by four countries. However, their economic conditions have been changed to be wrong. So that, construction of ITER can not begin after end of industrial design activities in 1998. Accordingly, they determined to continue the industrial design activities more three years in order to study low cost options and to test the superconductive model·coil. (S.Y.)

  9. ITER test programme

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Casini, G.

    1991-01-01

    ITER has been designed to operate in two phases. The first phase which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER. 15 refs, figs and tabs

  10. ITER plasma facing components

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  11. Effects of a range of machined and ground surface finishes on the simulated reactor helium corrosion of several candidate structural materials

    International Nuclear Information System (INIS)

    Thompson, L.D.

    1981-02-01

    This report discusses the corrosion behavior of several candidate reactor structural alloys in a simulated advanced high-temperature gas-cooled reactor (HTGR) environment over a range of lathe-machined and centerless-ground surface finishes. The helium environment contained 50 Pa H 2 /5 Pa CO/5 Pa CH 4 / 2 O (500 μatm H 2 /50 μatm CO/50 μatm CH 4 / 2 O) at 900 0 C for a total exposure of 3000 h. The test alloys included two vacuum-cast superalloys (IN 100 and IN 713LC); a centrifugally cast austenitic alloy (HK 40); three wrought high-temperature alloys (Alloy 800H, Hastelloy X, and Inconel 617); and a nickel-base oxide-dispersion-strengthened alloy (Inconel MA 754). Surface finish variations did not affect the simulated advanced-HTGR corrosion behavior of these materials. Under these conditions, the availability of reactant gaseous impurities controls the kinetics of the observed gas-metal interactions. Variations in the near-surface activities and mobilities of reactive solute elements, such as chromium, which might be expected to be affected by changes in surface finish, do not seem to greatly influence corrosion in this simulated advanced HTGR environment. 18 figures, 4 tables

  12. United States rejoin ITER

    International Nuclear Information System (INIS)

    Roberts, M.

    2003-01-01

    Upon pressure from the United States Congress, the US Department of Energy had to withdraw from further American participation in the ITER Engineering Design Activities after the end of its commitment to the EDA in July 1998. In the years since that time, changes have taken place in both the ITER activity and the US fusion community's position on burning plasma physics. Reflecting the interest in the United States in pursuing burning plasma physics, the DOE's Office of Science commissioned three studies as part of its examination of the option of entering the Negotiations on the Agreement on the Establishment of the International Fusion Energy Organization for the Joint Implementation of the ITER Project. These were a National Academy Review Panel Report supporting the burning plasma mission; a Fusion Energy Sciences Advisory Committee (FESAC) report confirming the role of ITER in achieving fusion power production, and The Lehman Review of the ITER project costing and project management processes (for the latter one, see ITER CTA Newsletter, no. 15, December 2002). All three studies have endorsed the US return to the ITER activities. This historical decision was announced by DOE Secretary Abraham during his remarks to employees of the Department's Princeton Plasma Physics Laboratory. The United States will be working with the other Participants in the ITER Negotiations on the Agreement and is preparing to participate in the ITA

  13. ITER and the fusion reactor: status and challenge to technology

    International Nuclear Information System (INIS)

    Lackner, K.

    2001-01-01

    Fusion has a high potential, but requires an integrated physics and technology effort without precedence in non-military R and D, the basic physics feasibility demonstration will be concluded with ITER, although R and D for efficiency improvement will continue. The essential technological issues remaining at the start of ITER operation concern materials questions: first wall components and radiation tolerant (low activation materials). This paper comprised just the copy of the slides presentation with the following subjects: magnetic confinement fusion, the Tokamak, progress in Tokamak performance, ITER: its geneology, physics basis-critical issues, cutaway of ITER-FEAT, R and D - divertor cassette (L-5), differences power plant-ITER, challenges for ITER and fusion plants, main technological problems (plasma facing materials), structural and functional materials for fusion power plants, ferritic steels, EUROFER development, improvements beyond ferritic steels, costing among others. (nevyjel)

  14. ITER CTA newsletter. No. 3

    International Nuclear Information System (INIS)

    2001-11-01

    This ITER CTA newsletter comprises reports of Dr. P. Barnard, Iter Canada Chairman and CEO, about the progress of the first formal ITER negotiations and about the demonstration of details of Canada's bid on ITER workshops, and Dr. V. Vlasenkov, Project Board Secretary, about the meeting of the ITER CTA project board

  15. ITER EDA Newsletter. V.4, no.1

    International Nuclear Information System (INIS)

    1995-01-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on (i) the seventh ITER Council Meeting held at the Naka Joint Work Site on 14-15 December 1994, (ii) the ''Confinement Modelling and Database Expert Group Workshop'' held in Seville, Spain, 3-4 October 1994, and (iii) the first meeting of the International Organizing Committee for the Seventh International Fusion Reactor Materials Conference

  16. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  17. ITER council proceedings: 1992

    International Nuclear Information System (INIS)

    1994-01-01

    At the signing of the ITER EDA Agreement on July, 1992, each of the Parties presented to the Director General the names of their designated members of the ITER Council. Upon receiving those names, the Director General stated that the ITER Engineering Design Activities were ''ready to begin''. The next step in this process was the convening of the first meeting of the ITER Council. The first meeting of the Council, held in Vienna, was opened by Director General Hans Blix. The second meeting was held in Moscow, the formal seat of the Council. This volume presents records of these first two Council meetings and, together with the previous volumes on the text of the Agreement and Protocol 1 and the preparations for their signing respectively, represents essential information on the evolution of the ITER EDA

  18. Full Dynamic Analysis of Mooring Solution Candidates - First Iteration

    DEFF Research Database (Denmark)

    Thomsen, Jonas Bjerg; Ferri, Francesco

    This report covers an initial full dynamic analysis of the mooring solutions for the four wave energy converters in the project “Mooring Solutions for Large Wave Energy Converters”. The analysis tends to provide the first understanding of the layouts and provide discussion on what parameters that...

  19. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  20. Extensive characterisation of advanced manufacturing solutions for the ITER Central Solenoid pre-compression system

    CERN Document Server

    Langeslag, S.A.E.; Libeyre, P.; Marcinek, D.J.; Zhang, Z.

    2015-01-01

    The ITER Central Solenoid (CS), positioned in the center of the ITER tokamak, will provide a magnetic field, contributing to the confinement of the plasma. The 13 m high CS consists of a vertical stack of 6 independently driven modules, dynamically activated. Resulting opposing currents can lead to high separation forces. A pre-compression structure is implemented to counteract these opposing forces, by realising a continuous 180 MN coil-to-coil contact loading. Preload is applied by mechanical fastening via 9 subunits, positioned along the coil stack, each consisting of 2 outer and 1 inner tie plate. The tie plates therefore need to feature outstanding mechanical behaviour in a large temperature range. High strength, Nitronic®-50 type F XM-19 austenitic stainless steel is selected as candidate material. The linearised stress distribution reaches approximately 250 MPa, leading to a required yield strength of 380 MPa at room temperature. Two different manufacturing methods are being studied for the procuremen...

  1. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.N. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)]. E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [EFDA-CSU, Max-Planck-Institut fuer Plasmaphysik, D-85748 Garching (Germany); Pestchanyi, S.E. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2007-06-15

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  2. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    Science.gov (United States)

    Bazylev, B. N.; Janeschitz, G.; Landman, I. S.; Loarte, A.; Pestchanyi, S. E.

    2007-06-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated.

  3. Melt damage simulation of W-macrobrush and divertor gaps after multiple transient events in ITER

    International Nuclear Information System (INIS)

    Bazylev, B.N.; Janeschitz, G.; Landman, I.S.; Loarte, A.; Pestchanyi, S.E.

    2007-01-01

    Tungsten in the form of macrobrush structure is foreseen as one of two candidate materials for the ITER divertor and dome. In ITER, even for moderate and weak ELMs when a thin shielding layer does not protect the armour surface from the dumped plasma, the main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. The melt erosion of W-macrobrush targets with different geometry of brush surface under the heat loads caused by weak ELMs is numerically investigated using the modified code MEMOS. The optimal angle of brush surface inclination that provides a minimum of surface roughness is estimated for given inclination angles of impacting plasma stream and given parameters of the macrobrush target. For multiple disruptions the damage of the dome gaps and the gaps between divertor cassettes caused by the radiation impact is estimated

  4. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  5. ITER towards the construction

    International Nuclear Information System (INIS)

    Shimomura, Y.

    2005-01-01

    The ITER Project has been significantly developed in the last few years in preparation for its construction. The ITER Participant's Negotiators have developed the Joint Implementation Agreement (JIA), ready for finalisation following selection of the construction site and nomination of the project's Director General. The ITER International Team and Participant Teams have continued technical and organisational preparations. Construction will be able to start immediately after the international ITER organisation is established, following signature of the JIA. The Project is strongly supported by the governments of the Participants as well as by the scientific community. The real negotiations, including siting and the final details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. Their wish to host ITER construction is too strong to allow convergence to a single site considering the ITER device in isolation. A broader collaboration among the Parties is therefore being contemplated, covering complementary activities to help accelerate fusion development towards a viable power source, and allow the Participants to reach a conclusion on ITER siting. This report reviews these preparations, and the status of negotiations

  6. Iterative solutions of finite difference diffusion equations

    International Nuclear Information System (INIS)

    Menon, S.V.G.; Khandekar, D.C.; Trasi, M.S.

    1981-01-01

    The heterogeneous arrangement of materials and the three-dimensional character of the reactor physics problems encountered in the design and operation of nuclear reactors makes it necessary to use numerical methods for solution of the neutron diffusion equations which are based on the linear Boltzmann equation. The commonly used numerical method for this purpose is the finite difference method. It converts the diffusion equations to a system of algebraic equations. In practice, the size of this resulting algebraic system is so large that the iterative methods have to be used. Most frequently used iterative methods are discussed. They include : (1) basic iterative methods for one-group problems, (2) iterative methods for eigenvalue problems, and (3) iterative methods which use variable acceleration parameters. Application of Chebyshev theorem to iterative methods is discussed. The extension of the above iterative methods to multigroup neutron diffusion equations is also considered. These methods are applicable to elliptic boundary value problems in reactor design studies in particular, and to elliptic partial differential equations in general. Solution of sample problems is included to illustrate their applications. The subject matter is presented in as simple a manner as possible. However, a working knowledge of matrix theory is presupposed. (M.G.B.)

  7. Physical and dosimetrical characterization of 4He and 16O beam interacting with tissue-like and candidates-shielding materials

    Science.gov (United States)

    La Tessa, Chiara; Zeitlin, Cary; Rusek, Adam; Durante, Marco; Schuy, Christoph; Sivertz, Michael

    2012-07-01

    The permanence of human in space has increased in the last decades with the establishment of space stations orbiting permanently around the Earth; furthermore, future plans are likely to include extended human missions in deep space outside the geomagnetosphere and settlements on other planets. The extensive exposure to the radiation environment represents one of the major limitations to space exploration due to its relation with severe health risks. The unfeasibility to stop the external radiation entirely motivates the investigation of shields able to minimize the total absorbed and equivalent dose to which the astronauts are exposed. The process of nuclear fragmentation plays a key role in this topic being the major responsible for modifying the radiation field that enters the spacecraft. Theoretical predictions on the dose received in a given scenario rely heavily on the accuracy of fragmentation cross sections and their uncertainties can be a central factor in limiting mission feasibility and duration. The interaction of 160 MeV/u Helium and 360 MeV/u Oxygen beams with water has been investigated in this work. The total charge-changing cross section has been estimated from the measurement of the attenuation of the primary ions in the target. For different target thicknesses, the yield and energy spectrum of charged and unchanged particles has been measured at several angles with respect to the primary beam direction. At the same position, microdosimetric spectra have been collected to characterize the quality of the radiation field and estimate the absorbed dose. Furthermore, total and partial-change-changing cross sections in candidate shielding materials are presented and compared with the results for water.

  8. Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in the US: A literature review

    International Nuclear Information System (INIS)

    Farmer, J.C.; McCright, R.D.

    1988-01-01

    Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys. Though all three austenitic candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these forms of localized attack. Both types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented for Alloy 825 under comparable conditions. Gamma irradiation has been found to enhance SCC of Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while microbiologically induced corrosion effects have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. Of the copper-based alloys, CDA 715 has the best overall resistance to localized attack. Its resistance to pitting is comparable to that of CDA 613 and superior to that of CDA 102. Observed rates of dealloying in CDA 715 are less than those observed in CDA 613 by orders of magnitude. The resistance of CDA 715 to SCC in tarnishing ammonical environments is comparable to that of CDA 102 and superior to that of CDA 613. Its resistance to SCC in nontarnishing ammonical environments is comparable to that of CDA 613 and superior to that of CDA 102. 22 refs., 8 figs., 4 tabs

  9. Corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

    International Nuclear Information System (INIS)

    Nakazono, Yoshihisa; Iwai, Takeo; Abe, Hiroaki

    2009-01-01

    The supercritical water-cooled reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy. Supercritical Water (SCW) has never been used in nuclear power applications. There are numerous potential problems, particularly with materials. As the operating temperature of SCWR will be between 553 K and 893 K with a pressure of 25 MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel has been developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. Austenitic Fe-base steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel was selected for possible use in supercritical water systems. The corrosion data of PNC1520 in SCW is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in SCW. The SCW corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520T) by using a SCW autoclave. The 1520S and 1520T are the first trial production materials of SCWR cladding candidate material in our group. Corrosion and compatibility tests on the austenitic 1520S and 1520T steels in supercritical water were performed at 673, 773 and 600degC with exposures up to 1000 h. We have evaluated the amount of weight gain, weight loss and weight of scale after the corrosion test in SCW for 1520S and 1520T austenitic steels. After 1000 h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m 2 at 400degC and 500degC. But 1520T weight increases more and weight loss than 1520S at 600degC. The SEM observation result of the surface after 1000 h corrosion of an test

  10. Advanced gas cooled nuclear reactor materials evaluation and development program. Selection of candidate alloys. Vol. 1. Advanced gas cooled reactor systems definition

    International Nuclear Information System (INIS)

    Marvin, M.D.

    1978-01-01

    Candidate alloys for a Very High Temperature Reactor (VHTR) Nuclear Process Heal (NPH) and Direct Cycle Helium Turbine (DCHT) applications in terms of the effect of the primary coolant exposure and thermal exposure were evaluated

  11. Perl Modules for Constructing Iterators

    Science.gov (United States)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  12. Dark matter candidates

    International Nuclear Information System (INIS)

    Turner, M.S.

    1989-01-01

    One of the simplest, yet most profound, questions we can ask about the Universe is, how much stuff is in it, and further what is that stuff composed of? Needless to say, the answer to this question has very important implications for the evolution of the Universe, determining both the ultimate fate and the course of structure formation. Remarkably, at this late date in the history of the Universe we still do not have a definitive answer to this simplest of questions---although we have some very intriguing clues. It is known with certainty that most of the material in the Universe is dark, and we have the strong suspicion that the dominant component of material in the Cosmos is not baryons, but rather is exotic relic elementary particles left over from the earliest, very hot epoch of the Universe. If true, the Dark Matter question is a most fundamental one facing both particle physics and cosmology. The leading particle dark matter candidates are: the axion, the neutralino, and a light neutrino species. All three candidates are accessible to experimental tests, and experiments are now in progress. In addition, there are several dark horse, long shot, candidates, including the superheavy magnetic monopole and soliton stars. 13 refs

  13. Overview and status of ITER internal components

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  14. Overview and status of ITER internal components

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  15. ITER Council proceedings: April 1988 - August 1989

    International Nuclear Information System (INIS)

    1990-01-01

    The ITER documentation series, of which this is the sixth report, began with a concise record of the decisions and actions taken in establishing ITER. The contents of that first report include the Terms of Reference Concerning Conceptual Design Activities for an International Thermonuclear Experimental Reactor. The Terms of Reference, which were part of the IAEA's invitation to the prospective parties, formally describe how the co-operative work of the four Parties in the specified activities is directed and managed. The first report in the series also covered activities from the initial meeting of the ITER Quadripartite Initiative Committee in March 1987 through March 1988. The present report is intended to make available in convenient form the essential information on ''landmark'' events in the direction of the ITER activities from the first meeting of the ITER Council (IC), in April 1988, through the letter report by the Council following their fourth meeting in July 1989. This report therefore covers approximately the first half of the Conceptual Design Activities, which are to be concluded in December 1990. THe next section of this report provides, for convenient reference, an overview of the organization and schedule that were adopted for the ITER activities through 1990. The sections that follow contain, for each of the four IC meetings during the period covered by this report, a copy of the official record of the transactions. The written reports of the ITER Management Committee (IMC) and the ITER Scientific and Technical Advisory Committee (ISTAC) to the IC in connection with the meetings are represented by summaries, prepared by the ITER Secretariat. These summaries include direct quotations of especially significant statements in the reports. In general, other supporting material is included only if it is of more than transitory significance. 1 fig and tabs

  16. ITER definition phase

    International Nuclear Information System (INIS)

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned as a fusion device which would demonstrate the scientific and technological feasibility of fusion power. As a first step towards achieving this goal, the European Community, Japan, the Soviet Union, and the United States of America have entered into joint conceptual design activities under the auspices of the International Atomic Energy Agency. A brief summary of the Definition Phase of ITER activities is contained in this report. Included in this report are the background, objectives, organization, definition phase activities, and research and development plan of this endeavor in international scientific collaboration. A more extended technical summary is contained in the two-volume report, ''ITER Concept Definition,'' IAEA/ITER/DS/3. 2 figs, 2 tabs

  17. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  18. An assessment of disruption erosion in ITER environment

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1994-01-01

    The behavior of divertor materials during a major disruption in ITER is very important for the successful and reliable operation of the reactor. Erosion of material surfaces due to the thermal energy dump can severely limit the lifetime of the plasma facing components therefore degrading reactor economic feasibility. A comprehensive numerical model recently developed is used in this analysis in which all major physical processes taking place during plasma-material interactions are included. Models to account for material thermal evolution, plasma-vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud are implemented in a self-consistent manner to realistically assess the disruption damage. The extent of the self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. The aim of this study is to estimate the divertor lifetime for a range of reactor conditions. Candidate materials such as beryllium and graphite are both considered in this analysis. The dependence of the divertor disruption lifetime on the characteristics of plasma-vapor interaction zone for incident plasma ions and electrons is analyzed and discussed. The effect of uncertainties in reactor disruption conditions on the overall divertor erosion lifetime is also analyzed

  19. Development of real time monitoring for ITER first wall erosion

    International Nuclear Information System (INIS)

    Berryman, Ian.; Pallaras, Luke; Thomson, Laura; Wang, Michael; Riley, Daniel P.

    2009-01-01

    Full text: This project aims to contribute to the current research on the first wall erosion diagnostic for the ITER fusion reactor. The plasma-facing first wall tiles of the ITER tokamak reactor are exposed to an expected neutron flux of O. 7 8 M W/m2 and a thermal load of O. 5M W/m 2 during operation. Instabilities in the magnetically confined plasma, such as edge-Iocalised modes, cause the plasma to come into direct contact with the first wall. The resulting thermal loads can vaporise and ablate the tile material. Moreover, a flux of high-energy neutrons produced during the fusion process results in a range of radiation effects. Therefore, a diagnostic is necessary to monitor the extent and rate of damage caused to the first wall. We have considered and critically assessed the viability of six alternative diagnostic methods, encompassing both established and novel concepts. From these, a design featuring embedded conducting elements was selected as the strongest candidate, as by monitoring electrical signals it has the potential to detect both bulk erosion and radiation damage.

  20. ITER EDA and technology

    International Nuclear Information System (INIS)

    Baker, C.C.

    2001-01-01

    The year 1998 was the culmination of the six-year Engineering Design Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project. The EDA results in design and validating technology R and D, plus the associated effort in voluntary physics research, is a significant achievement and major milestone in the history of magnetic fusion energy development. Consequently, the ITER EDA was a major theme at this Conference, contributing almost 40 papers

  1. Toward construction of ITER

    International Nuclear Information System (INIS)

    Shimomura, Yasuo

    2005-01-01

    The ITER Project has been significantly developed in the past years in preparation for its construction. The ITER Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project's Director General (DG). The ITER International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international ITER organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the ITER facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on ITER siting. (author)

  2. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

    The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

  3. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jay Jayakumar, R.; Tsuji, H.; Ohsaki, O.

    2001-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  4. ITER central solenoid manufacturing R and D

    International Nuclear Information System (INIS)

    Jayakumar, R.J.; Tsuji, H.; Ohsaki, O.

    1999-01-01

    The International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA) includes the development of high performance superconductors, high current joints between superconducting cables and insulating materials. Also in the EDA, the resulting products of this R and D are incorporated in a Central Solenoid Model Coil which utilizes full size conductors. The manufacturing of the model coil and components has led to the development of the design, materials, tooling and process which are fully applicable to the manufacture of the ITER relevant CS coil. The R and D is essentially complete and final stages of the CS Model Coil manufacturing are underway. (author)

  5. ITER CTA newsletter. No. 6

    International Nuclear Information System (INIS)

    2002-01-01

    This ITER CTA Newsletter issue comprises information about the following ITER Meetings: The second negotiation meeting on the joint implementation of ITER, held in Tokyo(Japan) on 22-23 January 2002, and an international ITER symposium on burning plasma science and technology, held the day later after the second negotiation meeting at the same place

  6. ITER CTA newsletter. No. 2

    International Nuclear Information System (INIS)

    2001-10-01

    This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

  7. Simulation of dust production in ITER transient events

    Energy Technology Data Exchange (ETDEWEB)

    Pestchanyi, S. [Forschungszentrum Karlsruhe (Germany)

    2007-07-01

    The tritium retention problem is a critical issue for the tokamak ITER performance. Tritium is trapped in redeposited T-C layers and at the surface of carbon dust, where it is retained in form of various hydrocarbons. The area of dust surface and hence, the amount of tritium deposited on the surface depends on the dust amount and of the dust sizes. The carbon dust appears as a result of brittle destruction at the surface of the carbon fibre composite (CFC) which is now the reference armour material for the most loaded part of tokamak divertor. Stationary heat flux on the ITER divertor armour does not cause its brittle destruction and does not produce dust. However, according to the modern understanding of tokamak fusion devices performance, the most attractive regime of ITER operation is the ELMy H mode. This regime is associated with a repetitive short time increase of heat flux at the CFC divertor armour of 2-3 orders of magnitude over its stationary value during edge localized modes (ELMs). Under influence of these severe heat shocks CFC armour can crack due to the thermostress, producing a dust of carbon. Besides, a carbon dust produced during disruptions due to brittle destruction of the armour under influence of thermoshock. Most of the modern tokamaks do not produce the ELMs powerful enough to cause CFC brittle destruction at the divertor surface, except of very special regimes in JET. This is why the CFC erosion and dust production could be investigated now only theoretically and experimentally in plasma guns and electron beam facilities. Simulation of the CFC brittle destruction has been done using the code PEGASUS already developed and tested in FZK for simulation of erosion for ITER candidate materials under the heat shocks. After upgrades the code was used for simulation of the amount of carbon dust particles and of the distribution of their sizes. The code has been tested against available experimental data from the plasma gun MK-200UG and from the

  8. Simulation of dust production in ITER transient events

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2007-01-01

    The tritium retention problem is a critical issue for the tokamak ITER performance. Tritium is trapped in redeposited T-C layers and at the surface of carbon dust, where it is retained in form of various hydrocarbons. The area of dust surface and hence, the amount of tritium deposited on the surface depends on the dust amount and of the dust sizes. The carbon dust appears as a result of brittle destruction at the surface of the carbon fibre composite (CFC) which is now the reference armour material for the most loaded part of tokamak divertor. Stationary heat flux on the ITER divertor armour does not cause its brittle destruction and does not produce dust. However, according to the modern understanding of tokamak fusion devices performance, the most attractive regime of ITER operation is the ELMy H mode. This regime is associated with a repetitive short time increase of heat flux at the CFC divertor armour of 2-3 orders of magnitude over its stationary value during edge localized modes (ELMs). Under influence of these severe heat shocks CFC armour can crack due to the thermostress, producing a dust of carbon. Besides, a carbon dust produced during disruptions due to brittle destruction of the armour under influence of thermoshock. Most of the modern tokamaks do not produce the ELMs powerful enough to cause CFC brittle destruction at the divertor surface, except of very special regimes in JET. This is why the CFC erosion and dust production could be investigated now only theoretically and experimentally in plasma guns and electron beam facilities. Simulation of the CFC brittle destruction has been done using the code PEGASUS already developed and tested in FZK for simulation of erosion for ITER candidate materials under the heat shocks. After upgrades the code was used for simulation of the amount of carbon dust particles and of the distribution of their sizes. The code has been tested against available experimental data from the plasma gun MK-200UG and from the

  9. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huguet, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.R.; Shimada, M.; Aymar, R.; Chuyanov, V.A.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties. (author)

  10. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    International Nuclear Information System (INIS)

    Hong, Bong Guen; In, S. R.; Bae, Y. D.

    2006-02-01

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  11. Establishment of design and fabrication technology and domestic qualification for ITER blanket system

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Bong Guen; In, S. R.; Bae, Y. D. (and others)

    2006-02-15

    To obtain and analyze the detailed design and manufacturing technology of the blanket system for each components, the related data are collected through the various sources. And also, design processes and results of the FWs, shield blocks, and TBMs are investigated. From these analysis of the blanket R and D status of each party, we develop the KO R and D plan and it is used in the selection of manufacturing method and the materials. For the ITA16-10 subtask1, we had the official agreement with ITER IT in December 2004 for the qualification of the FW panel fabrication methods and to establish the NDT methods for the FW panel. From the technical reports we published, we compare the manufacturing methods and the proposed material for each component according to the parties. Be is proposed as a plasma facing material and most parties have interest in S-65C. Cu alloy is proposed as a heat sink material and DSCu or CuCrZr are investigated now. For the structural material, stainless steel such as SS316L(N) is investigated internationally. HIP and brazing are proposed as the manufacturing methods. In order to establish the blanket system technology, design contents of shield block by ITER IT and other parties were investigated through participating the international workshop and meeting, dispatching the researcher to the ITER IT or other parties to collect the drafting and 3D modeling files. The modification items of blanket design were investigated and a researcher was dispatched in the ITER IT and participated in the analysis on cooling problem in shield block such as front header and drilled manifold. To investigate the development status of TBM, we participated the 14th TBWG meeting and proposed the KO HCSB and HCML as candidates. And also, we obtain the R and D results of other parties and make document about the R and D status of other parties for the TBM. Finally, we establish the KO TBM R and D plan and proposed it to ITER IT and other parties. In which, the

  12. Irradiation testing of stainless steel plate material and weldments. Report on ITER Task T14, Part B. Tensile properties after 0.5 and 5 dpa at 350 and 500 K

    International Nuclear Information System (INIS)

    Rensman, J.W.; Boskeljon, J.; Horsten, M.G.; De Vries, M.I.

    1997-10-01

    The tensile properties of unirradiated and neutron irradiated type 316L(N)-SPH stainless steel plate, EB weldments, 16-8 TIG-weldments, and full 16-8 TIG-deposits have been measured. Miniature 4 mm diameter test specimens of the European Reference Heat 1 and 2 (ERH), and 4 mm and some 8 mm diameter specimens of the weldments mentioned above, were irradiated in the High Flux Reactor (HFR) in Petten, The Netherlands, simulating the first wall conditions by a combination of high displacement damage with high amounts of helium. The irradiation conditions were 0.5 and 5 displacements per atom (dpa) at 350K and 0.5 and 5 dpa at 500K. Testing temperatures ranged from 300K to 850K. This work was performed as part of the European Fusion Technology Programme for ITER as 'Irradiation testing of stainless steel' The report contains the experimental conditions and summarises the results. The tensile properties of the unirradiated ERH's 1 and 2 plate materials were found to differ slightly but significantly: ERH2 has a lower UTS, but higher yield strength and ductility than ERH1. The plate materials have lower yield strength in the unirradiated condition than all of the weldments (EB, TIG-weld and TIG-deposit), accompanied by a higher ductility of the plate materials. When irradiated at 350K the differences in strength between the plate and weld materials decrease, but the ductility of the plate remains higher than that of the weldments. A saturation of irradiation damage has taken place already at about 0.5 dpa. When irradiated at 500K the plate material continuously hardens up to 5 dpa, where it has lost all uniform plastic ductility. The weldments show similar but less dramatic hardening and loss of ductility as the plate material for both irradiation conditions. 54 figs., 17 tabs., 21 refs

  13. The ITER activity

    International Nuclear Information System (INIS)

    Glass, A.J.

    1991-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a collaboration among four parties, the United States, the Soviet Union, Japan, and the European Communities, to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes. ITER will demonstrate this through the construction of a tokamak fusion reactor capable of generating 1000 megawatts of fusion power. The ITER project has three missions, as follows: (1) Physics mission -- to demonstrate ignition and controlled burn, with pulse durations from 200 to 1000 S; (2) Technology mission -- to demonstrate the technologies essential to a reactor in an integrated system, operating with high reliability and availability in pulsed operation, with steady-state operation as the ultimate goal; and (3) Testing mission -- to test nuclear and high-heat-flux components at flux levels for 1 mw/m 2 , and fluences of order 1 mw-yr/m 2

  14. Earthly sun called ITER

    International Nuclear Information System (INIS)

    Pozdeyev, Mikhail

    2002-01-01

    Full text: Participating in the film are Academicians Velikhov and Glukhikh, Mr. Filatof, ITER Director from Russia, Mr. Sannikov from Kurchatov Institute. The film tells about the starting point of the project (Mr. Lavrentyev), the pioneers of the project (Academicians Tamme, Sakharov, Artsimovich) and about the situation the project is standing now. Participating in [ITER now are the US, Russia, Japan and the European Union. There are two associated members as well - Kazakhstan and Canada. By now the engineering design phase has been finished. Computer animation used in the video gives us the idea how the first thermonuclear reactor based on famous Russian TOKOMAK works. (author)

  15. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  16. Iterated multidimensional wave conversion

    International Nuclear Information System (INIS)

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-01-01

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  17. Physics fundamentals for ITER

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.

    1999-01-01

    The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an 'empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should

  18. Physics research needs for ITER

    International Nuclear Information System (INIS)

    Sauthoff, N.R.

    1995-01-01

    Design of ITER entails the application of physics design tools that have been validated against the world-wide data base of fusion research. In many cases, these tools do not yet exist and must be developed as part of the ITER physics program. ITER's considerable increases in power and size demand significant extrapolations from the current data base; in several cases, new physical effects are projected to dominate the behavior of the ITER plasma. This paper focuses on those design tools and data that have been identified by the ITER team and are not yet available; these needs serve as the basis for the ITER Physics Research Needs, which have been developed jointly by the ITER Physics Expert Groups and the ITER design team. Development of the tools and the supporting data base is an on-going activity that constitutes a significant opportunity for contributions to the ITER program by fusion research programs world-wide

  19. Proposal for the use of new materials in the TOKAMAK building cover; Contrato de ingenieria/arquitectura para el proyecto ITER

    Energy Technology Data Exchange (ETDEWEB)

    Chiva, L.

    2011-07-01

    It was considered relevant and innovative to apply new structural materials to the construction of the roof of the building that lodged the TOKAMAK reactor, with the aim of achieving a severe reduction of the weight of the roof structure that result in greater ease of mounting, minor charges on the walls and foundations of the building and a reduced impact on the distribution of masses of the building scheme.

  20. The physics role of ITER

    International Nuclear Information System (INIS)

    Rutherford, P.H.

    1997-04-01

    Experimental research on the International Thermonuclear Experimental Reactor (ITER) will go far beyond what is possible on present-day tokamaks to address new and challenging issues in the physics of reactor-like plasmas. First and foremost, experiments in ITER will explore the physics issues of burning plasmas--plasmas that are dominantly self-heated by alpha-particles created by the fusion reactions themselves. Such issues will include (i) new plasma-physical effects introduced by the presence within the plasma of an intense population of energetic alpha particles; (ii) the physics of magnetic confinement for a burning plasma, which will involve a complex interplay of transport, stability and an internal self-generated heat source; and (iii) the physics of very-long-pulse/steady-state burning plasmas, in which much of the plasma current is also self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achieving and sustaining burning plasma regimes in a tokamak necessarily requires plasmas that are larger than those in present experiments and have higher energy content and power flow, as well as much longer pulse length. Accordingly, the experimental program on ITER will embrace the study of issues of plasma physics and plasma-materials interactions that are specific to a reactor-scale fusion experiment. Such issues will include (i) confinement physics for a tokamak in which, for the first time, the core-plasma and the edge-plasma are simultaneously in a reactor-like regime; (ii) phenomena arising during plasma transients, including so-called disruptions, in regimes of high plasma current and thermal energy; and (iii) physics of a radiative divertor designed for handling high power flow for long pulses, including novel plasma and atomic-physics effects as well as materials science of surfaces subject to intense plasma interaction. Experiments on ITER will be conducted by researchers in control rooms situated at major

  1. Overview of the JET results in support to ITER

    DEFF Research Database (Denmark)

    Litaudon, X.; Abduallev, S.; Abhangi, M.

    2017-01-01

    The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scal...

  2. Iterative List Decoding

    DEFF Research Database (Denmark)

    Justesen, Jørn; Høholdt, Tom; Hjaltason, Johan

    2005-01-01

    We analyze the relation between iterative decoding and the extended parity check matrix. By considering a modified version of bit flipping, which produces a list of decoded words, we derive several relations between decodable error patterns and the parameters of the code. By developing a tree...... of codewords at minimal distance from the received vector, we also obtain new information about the code....

  3. ITER power electrical networks

    International Nuclear Information System (INIS)

    Sejas Portela, S.

    2011-01-01

    The ITER project (International Thermonuclear Experimental Reactor) is an international effort to research and development to design, build and operate an experimental facility to demonstrate the scientific and technological possibility of obtaining useful energy from the physical phenomenon known as nuclear fusion.

  4. ITER conceptual design report

    International Nuclear Information System (INIS)

    1991-01-01

    Results of the International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity (CDA) are reported. This report covers the Terms of Reference for the project: defining the technical specifications, defining future research needs, define site requirements, and carrying out a coordinated research effort coincident with the CDA. Refs, figs and tabs

  5. Nuclear analysis for ITER

    International Nuclear Information System (INIS)

    Santoro, R.T.; Iida, H.; Khripunov, V.; Petrizzi, L.; Sato, S.; Sawan, M.; Shatalov, G.; Schipakin, O.

    2001-01-01

    This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shut-down. (author)

  6. ITER at Cadarache

    International Nuclear Information System (INIS)

    2005-06-01

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  7. ITER conceptual design

    International Nuclear Information System (INIS)

    Tomabechi, K.; Gilleland, J.R.; Sokolov, Yu.A.; Toschi, R.

    1991-01-01

    The Conceptual Design Activities of the International Thermonuclear Experimental Reactor (ITER) were carried out jointly by the European Community, Japan, the Soviet Union and the United States of America, under the auspices of the International Atomic Energy Agency. The European Community provided the site for joint work sessions at the Max-Planck-Institut fuer Plasmaphysik in Garching, Germany. The Conceptual Design Activities began in the spring of 1988 and ended in December 1990. The objectives of the activities were to develop the design of ITER, to perform a safety and environmental analysis, to define the site requirements as well as the future research and development needs, to estimate the cost and manpower, and to prepare a schedule for detailed engineering design, construction and operation. On the basis of the investigation and analysis performed, a concept of ITER was developed which incorporated maximum flexibility of the performance of the device and allowed a variety of operating scenarios to be adopted. The heart of the machine is a tokamak having a plasma major radius of 6 m, a plasma minor radius of 2.15 m, a nominal plasma current of 22 MA and a nominal fusion power of 1 GW. The conceptual design can meet the technical objectives of the ITER programme. Because of the success of the Conceptual Design Activities, the Parties are now considering the implementation of the next phase, called the Engineering Design Activities. (author). Refs, figs and tabs

  8. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huget, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.; Shimada, M.; Aymar, R.; Chuyanov, V.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first 10 years' operation will be devoted primarily to physics issues at low neutron fluence and the following 10 years' operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes such as inductive high Q modes, long pulse hybrid modes, non-inductive steady-state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours per day but also in involving the world-wide fusion communities and in promoting scientific competition among the Parties. (author)

  9. ITER fuel cycle

    International Nuclear Information System (INIS)

    Leger, D.; Dinner, P.; Yoshida, H.

    1991-01-01

    Resulting from the Conceptual Design Activities (1988-1990) by the parties involved in the International Thermonuclear Experimental Reactor (ITER) project, this document summarizes the design requirements and the Conceptual Design Descriptions for each of the principal subsystems and design options of the ITER Fuel Cycle conceptual design. The ITER Fuel Cycle system provides for the handling of all tritiated water and gas mixtures on ITER. The system is subdivided into subsystems for fuelling, primary (torus) vacuum pumping, fuel processing, blanket tritium recovery, and common processes (including isotopic separation, fuel management and storage, and processes for detritiation of solid, liquid, and gaseous wastes). After an introduction describing system function and conceptual design procedure, a summary of the design is presented including a discussion of scope and main parameters, and the fuel design options for fuelling, plasma chamber vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary and common processes. Design requirements are defined and design descriptions are given for the various subsystems (fuelling, plasma vacuum pumping, fuel cleanup, blanket tritium recovery, and auxiliary/common processes). The document ends with sections on fuel cycle design integration, fuel cycle building layout, safety considerations, a summary of the research and development programme, costing, and conclusions. Refs, figs and tabs

  10. Advances in iterative methods

    International Nuclear Information System (INIS)

    Beauwens, B.; Arkuszewski, J.; Boryszewicz, M.

    1981-01-01

    Results obtained in the field of linear iterative methods within the Coordinated Research Program on Transport Theory and Advanced Reactor Calculations are summarized. The general convergence theory of linear iterative methods is essentially based on the properties of nonnegative operators on ordered normed spaces. The following aspects of this theory have been improved: new comparison theorems for regular splittings, generalization of the notions of M- and H-matrices, new interpretations of classical convergence theorems for positive-definite operators. The estimation of asymptotic convergence rates was developed with two purposes: the analysis of model problems and the optimization of relaxation parameters. In the framework of factorization iterative methods, model problem analysis is needed to investigate whether the increased computational complexity of higher-order methods does not offset their increased asymptotic convergence rates, as well as to appreciate the effect of standard relaxation techniques (polynomial relaxation). On the other hand, the optimal use of factorization iterative methods requires the development of adequate relaxation techniques and their optimization. The relative performances of a few possibilities have been explored for model problems. Presently, the best results have been obtained with optimal diagonal-Chebyshev relaxation

  11. ITER neutral beam system

    International Nuclear Information System (INIS)

    Mondino, P.L.; Di Pietro, E.; Bayetti, P.

    1999-01-01

    The Neutral Beam (NB) system for the International Thermonuclear Experimental Reactor (ITER) has reached a high degree of integration with the tokamak and with the rest of the plant. Operational requirements and maintainability have been considered in the design. The paper considers the integration with the tokamak, discusses design improvements which appear necessary and finally notes R and D progress in key areas. (author)

  12. Iterative software kernels

    Energy Technology Data Exchange (ETDEWEB)

    Duff, I.

    1994-12-31

    This workshop focuses on kernels for iterative software packages. Specifically, the three speakers discuss various aspects of sparse BLAS kernels. Their topics are: `Current status of user lever sparse BLAS`; Current status of the sparse BLAS toolkit`; and `Adding matrix-matrix and matrix-matrix-matrix multiply to the sparse BLAS toolkit`.

  13. ITER Safety and Licensing

    International Nuclear Information System (INIS)

    Girard, J-.P; Taylor, N.; Garin, P.; Uzan-Elbez, J.; GULDEN, W.; Rodriguez-Rodrigo, L.

    2006-01-01

    The site for the construction of ITER has been chosen in June 2005. The facility will be implemented in Europe, south of France close to Marseille. The generic safety scheme is now under revision to adapt the design to the host country regulation. Even though ITER will be an international organization, it will have to comply with the French requirements in the fields of public and occupational health and safety, nuclear safety, radiation protection, licensing, nuclear substances and environmental protection. The organization of the central team together with its partners organized in domestic agencies for the in-kind procurement of components is a key issue for the success of the experimentation. ITER is the first facility that will achieve sustained nuclear fusion. It is both important for the experimental one-of-a-kind device, ITER itself, and for the future of fusion power plants to well understand the key safety issues of this potential new source of energy production. The main safety concern is confinement of the tritium, activated dust in the vacuum vessel and activated corrosion products in the coolant of the plasma-facing components. This is achieved in the design through multiple confinement barriers to implement the defence in depth approach. It will be demonstrated in documents submitted to the French regulator that these barriers maintain their function in all postulated incident and accident conditions. The licensing process started by examination of the safety options. This step has been performed by Europe during the candidature phase in 2002. In parallel to the final design, and taking into account the local regulations, the Preliminary Safety Report (RPrS) will be drafted with support of the European partner and others in the framework of ITER Task Agreements. Together with the license application, the RPrS will be forwarded to the regulatory bodies, which will launch public hearings and a safety review. Both processes must succeed in order to

  14. Behaviour of the ASDEX pressure gauge at high neutral gas pressure and applications for ITER

    International Nuclear Information System (INIS)

    Scarabosio, A.; Haas, G.

    2008-01-01

    The ASDEX Pressure Gauge is, at present, the main candidate for in-vessel neutral pressure measurement in ITER. Although the APG output is found to saturate at around 15 Pa, below the ITER requirement of 20 Pa. We show, here, that with small modifications of the gauge geometry and potentials settings we can achieve satisfactory behaviour up to 30 Pa at 6 T

  15. Status of ITER

    International Nuclear Information System (INIS)

    Aymar, R.

    2002-01-01

    At the end of engineering design activities (EDA) in July 2001, all the essential elements became available to make a decision on construction of ITER. A sufficiently detailed and integrated engineering design now exists for a generic site, has been assessed for feasibility, and costed, and essential physics and technology R and D has been carried out to underpin the design choices. Formal negotiations have now begun between the current participants--Canada, Euratom, Japan, and the Russian Federation--on a Joint Implementation Agreement for ITER which also establishes the legal entity to run ITER. These negotiations are supported on technical aspects by Coordinated Technical Activities (CTA), which maintain the integrity of the project, for the good of all participants, and concentrate on preparing for procurement by industry of the longest lead items, and for formal application for a construction license with the host country. This paper highlights the main features of the ITER design. With cryogenically-cooled magnets close to neutron-generating plasma, the design of shielding with adequate access via port plugs for auxiliaries such as heating and diagnostics, and of remote replacement and refurbishing systems for in-vessel components, are particularly interesting nuclear technology challenges. Making a safety case for ITER to satisfy potential regulators and to demonstrate, as far as possible at this stage, the environmental attractiveness of fusion as an energy source, is also important. The paper gives illustrative details on this work, and an update on the progress of technical preparations for construction, as well as the status of the above negotiations

  16. Krylov iterative methods and synthetic acceleration for transport in binary statistical media

    International Nuclear Information System (INIS)

    Fichtl, Erin D.; Warsa, James S.; Prinja, Anil K.

    2009-01-01

    In particle transport applications there are numerous physical constructs in which heterogeneities are randomly distributed. The quantity of interest in these problems is the ensemble average of the flux, or the average of the flux over all possible material 'realizations.' The Levermore-Pomraning closure assumes Markovian mixing statistics and allows a closed, coupled system of equations to be written for the ensemble averages of the flux in each material. Generally, binary statistical mixtures are considered in which there are two (homogeneous) materials and corresponding coupled equations. The solution process is iterative, but convergence may be slow as either or both materials approach the diffusion and/or atomic mix limits. A three-part acceleration scheme is devised to expedite convergence, particularly in the atomic mix-diffusion limit where computation is extremely slow. The iteration is first divided into a series of 'inner' material and source iterations to attenuate the diffusion and atomic mix error modes separately. Secondly, atomic mix synthetic acceleration is applied to the inner material iteration and S 2 synthetic acceleration to the inner source iterations to offset the cost of doing several inner iterations per outer iteration. Finally, a Krylov iterative solver is wrapped around each iteration, inner and outer, to further expedite convergence. A spectral analysis is conducted and iteration counts and computing cost for the new two-step scheme are compared against those for a simple one-step iteration, to which a Krylov iterative method can also be applied.

  17. An Asdex-type divertor for ITER

    International Nuclear Information System (INIS)

    Fowler, T.K.

    1989-01-01

    An Asdex-type local divertor is proposed for ITER consisting of a copper poloidal field coil adjacent to the plasma. Estimates indicate that the power consumption is acceptable. Advantages would be a much reduced heat load not very sensitive to magnetic perturbations. A disadvantage is the finite lifetime under neutron bombardment that would require periodic replacement of the divertor coils in a reactor, but probably not in ITER because of its limited fluence. Another disadvantage would be poorer blanket coverage unless the divertor coil itself incorporates breeding material. 3 figs

  18. ITER EDA status

    International Nuclear Information System (INIS)

    Aymar, R.

    2001-01-01

    The Project has focused on drafting the Plant Description Document (PDD), which will be published as the Technical Basis for the ITER Final Design Report (FDR), and its related documentation in time for the ITER review process. The preparations have involved continued intensive detailed design work, analyses and assessments by the Home Teams and the Joint Central Team, who have co-operated closely and efficiently. The main technical document has been completed in time for circulation, as planned, to TAC members for their review at TAC-17 (19-22 February 2001). Some of the supporting documents, such as the Plant Design Specification (PDS), Design Requirements and Guidelines (DRG1 and DRG2), and the Plant Safety Requirement (PSR) are also available for reference in draft form. A summary paper of the PDD for the Council's information is available as a separate document. A new documentation structure for the Project has been established. This hierarchical structure for documentation facilitates the entire organization in a way that allows better change control and avoids duplications. The initiative was intended to make this documentation system valid for the construction and operation phases of ITER. As requested, the Director and the JCT have been assisting the Explorations to plan for future joint technical activities during the Negotiations, and to consider technical issues important for ITER construction and operation for their introduction in the draft of a future joint implementation agreement. As charged by the Explorers, the Director has held discussions with the Home Team Leaders in order to prepare for the staffing of the International Team and Participants Teams during the Negotiations (Co-ordinated Technical Activities, CTA) and also in view of informing all ITER staff about their future directions in a timely fashion. One important element of the work was the completion by the Parties' industries of costing studies of about 83 ''procurement packages

  19. ITER EDA newsletter. V. 5, no.1

    International Nuclear Information System (INIS)

    1996-01-01

    This issue of the ITER EDA (Engineering Design Activities) Newsletter contains reports on the RF-Based ITER JCT (Joint Central Team) Support Design Team (by N. Kornev), the third international workshop on plasma disruptions (by Dr. A. Hassanein and Dr. V. Litunovski) held at Obninsk, Russia, September 28-29, 1995, and the IAEA Advisory Group Meeting on Completion of Fendl-1 and the start of Fendl-2 (by Dr. A.B. Pashchenko); the Fendl library is a comprehensive collection of high-quality nuclear data, selected from the various existing national data libraries, covering the necessary nuclear input data for all physics and engineering aspects of the material development, design, operation, and safety of the ITER project in its current EDA phase

  20. Ceramics radiation effects issues for ITER

    International Nuclear Information System (INIS)

    Zinkle, S.J.

    1993-01-01

    The key radiation effects issues associated with the successful operation of ceramic materials in components of the planned International Thermonuclear Experimental Reactor (ITER) are discussed. Radiation-induced volume changes and degradation of the mechanical properties should not be a serious issue for the fluences planned for ITER. On the other hand, radiation-induced electrical degradation effects may severely limit the allowable exposure of ceramic insulators. Degradation of the loss tangent and thermal conductivity may also restrict the location of some components such as ICRH feedthrough insulators to positions far away from the first wall. In-situ measurements suggest that the degradation of physical properties in ceramics during irradiation is greater than that measured in postirradiation tests. Additional in-situ data during neutron irradiation are needed before engineering designs for ITER can be finalized

  1. ITER EDA newsletter. V. 10, no. 4

    International Nuclear Information System (INIS)

    2001-04-01

    This ITER EDA Newsletter presents an overview of the Fourteenth Meeting of the ITER Physics Expert Group on Diagnostics which was held at the Institute for Plasma Physics, Juelich, Germany, 21-23 March 2001. The summary of the Meeting covers the discussions of the Expert Group as well as developments reported on similar meetings concerning ongoing work in diagnostic design and ITER relevant diagnostic development work which took place nearly at the same time. In addition, the outline of the material treated at the International Workshop on the Confinement Database and Modelling Expert Group in collaboration with the Edge and Pedestal Physics Expert Group which was held on 2-6 April 2001 at the Plasma Physics Research Centre of Lausanne (CRPP) Switzerland is presented

  2. ITER CTA newsletter. No. 4

    International Nuclear Information System (INIS)

    2001-12-01

    This ITER CTA Newsletter contains information about the organization of the ITER Co-ordinated Technical Activities (CTA) International Team as the follow-up of the ITER CTA project board meeting in Toronto on 7 November 2001. It also includes a summary on the start of the international tokamak physics activity by Dr. D. Campbell, Chair of the ITPA Co-ordinating Committee

  3. ITER CTA newsletter. No. 9

    International Nuclear Information System (INIS)

    2002-06-01

    This ITER CTA newsletter contains information about the Fourth Negotiations Meeting on the Joint Implementation of ITER held in Cadarache, France on 4-6 June 2002 and about the meeting of the ITER CTA Project Board which took place on the occasion of the N4 Meeting at Cadarache on 3-4 June 2002

  4. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held on 23 February in Garching, Germany. The main topics were: the consideration of the report by the Director on the ITER EDA Status, the review of the Work Programme, the review of the Joint Fund, the review of a schedule of ITER meetings, and the arrangements for termination and wind-up of the EDA

  5. ITER CTA newsletter. No. 1

    International Nuclear Information System (INIS)

    2001-01-01

    This ITER CTA newsletter comprises reports on ITER co-ordinated technical activities, information about the Meeting of the ITER CTA project board which took place in Vienna on 16 July 2001, and the Meeting of the expert group on MHD, disruptions and plasma control which was held on 25-26 June 2001 in Funchal, Madeira

  6. Status of the ITER EDA

    International Nuclear Information System (INIS)

    Aymar, R.

    2000-01-01

    This article summarizes progress made in the ITER Engineering Design Activities in the period between the ITER Meeting in Tokyo (January 2000) and June 2000. Topics: Termination of EDA, Joint Central Team and Support, Task Assignments, ITER Physics, Urgent and High Priority Physics Research Areas

  7. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola

    2014-05-02

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  8. Iterative supervirtual refraction interferometry

    KAUST Repository

    Al-Hagan, Ola; Hanafy, Sherif M.; Schuster, Gerard T.

    2014-01-01

    In refraction tomography, the low signal-to-noise ratio (S/N) can be a major obstacle in picking the first-break arrivals at the far-offset receivers. To increase the S/N, we evaluated iterative supervirtual refraction interferometry (ISVI), which is an extension of the supervirtual refraction interferometry method. In this method, supervirtual traces are computed and then iteratively reused to generate supervirtual traces with a higher S/N. Our empirical results with both synthetic and field data revealed that ISVI can significantly boost up the S/N of far-offset traces. The drawback is that using refraction events from more than one refractor can introduce unacceptable artifacts into the final traveltime versus offset curve. This problem can be avoided by careful windowing of refraction events.

  9. ITER technical basis

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties.

  10. ITER technical basis

    International Nuclear Information System (INIS)

    2002-01-01

    Following on from the Final Report of the EDA(DS/21), and the summary of the ITER Final Design report(DS/22), the technical basis gives further details of the design of ITER. It is in two parts. The first, the Plant Design specification, summarises the main constraints on the plant design and operation from the viewpoint of engineering and physics assumptions, compliance with safety regulations, and siting requirements and assumptions. The second, the Plant Description Document, describes the physics performance and engineering characteristics of the plant design, illustrates the potential operational consequences foe the locality of a generic site, gives the construction, commissioning, exploitation and decommissioning schedule, and reports the estimated lifetime costing based on data from the industry of the EDA parties

  11. Towards the procurement of the ITER divertor

    International Nuclear Information System (INIS)

    Merola, M.; Tivey, R.; Martin, A.; Pick, M.

    2006-01-01

    The procurement of the ITER divertor is planned to start in 2009. On the basis of the present common understanding of the sharing of the ITER components, the Japanese Participating Team (JAPT) will supply the outer vertical target, the Russian Federation (RF) PT the dome liner and will perform the high heat flux testing, the EU PT will supply the inner vertical targets and the cassette bodies, including final assembly of the divertor plasma-facing components (PFCs). The manufacturing of the PFCs of the ITER divertor represents a challenging endeavor due to the high technologies which are involved, and due to the unprecedented series production. To mitigate the associated risks, special arrangements need to be put in place prior to and during procurement to ensure quality and to keep to the time schedule. Before procurement can start, an ITER review of the qualification and production capability of each candidate PT is planned. Well in advance of the assumed start of the procurement, each PT which would like to contribute to the divertor PFC procurement, should first demonstrate its technical qualification to carry out the procurement with the required quality, and in an efficient and timely manner. Appropriate precautions, like subdivision of the procurement into stages, are also to be adopted during the procurement phase to mitigate the consequences of possible unexpected manufacturing problems. In preparation for writing the procurement specification for the vertical targets, the topic of setting acceptance criteria is also being addressed. This activity has the objective of defining workable acceptance criteria for the PFC armour joints. A complete set of analyses is also in progress to assess the latest design modifications against the design requirements. This task includes neutronic, shielding, thermo-mechanical and electromagnetic analyses. More than half of the ITER plasma parameters that must be measured and the related diagnostics are located in the

  12. Conformable variational iteration method

    Directory of Open Access Journals (Sweden)

    Omer Acan

    2017-02-01

    Full Text Available In this study, we introduce the conformable variational iteration method based on new defined fractional derivative called conformable fractional derivative. This new method is applied two fractional order ordinary differential equations. To see how the solutions of this method, linear homogeneous and non-linear non-homogeneous fractional ordinary differential equations are selected. Obtained results are compared the exact solutions and their graphics are plotted to demonstrate efficiency and accuracy of the method.

  13. Iterated Leavitt Path Algebras

    International Nuclear Information System (INIS)

    Hazrat, R.

    2009-11-01

    Leavitt path algebras associate to directed graphs a Z-graded algebra and in their simplest form recover the Leavitt algebras L(1,k). In this note, we introduce iterated Leavitt path algebras associated to directed weighted graphs which have natural ± Z grading and in their simplest form recover the Leavitt algebras L(n,k). We also characterize Leavitt path algebras which are strongly graded. (author)

  14. ICP (ITER Collaborative Platform)

    Energy Technology Data Exchange (ETDEWEB)

    Capuano, C.; Carayon, F.; Patel, V. [ITER, 13 - St. Paul-Lez Durance (France)

    2009-07-01

    The ITER organization has the necessity to manage a massive amount of data and processes. Each team requires different process and databases often interconnected with those of others teams. ICP is the current central ITER repository of structured and unstructured data. All data in ICP is served and managed via a web interface that provides global accessibility with a common user friendly interface. This paper will explain the model used by ICP and how it serves the ITER project by providing a robust and agile platform. ICP is developed in ASP.NET using MSSQL Server for data storage. It currently houses 15 data driven applications, 150 different types of record, 500 k objects and 2.5 M references. During European working hours the system averages 150 concurrent users and 20 requests per second. ICP connects to external database applications to provide a single entry point to ITER data and a safe shared storage place to maintain this data long-term. The Core model provides an easy to extend framework to meet the future needs of the Organization. ICP follows a multi-tier architecture, providing logical separation of process. The standard three-tier architecture is expanded, with the data layer separated into data storage, data structure, and data access components. The business or applications logic layer is broken up into a common business functionality layer, a type specific logic layer, and a detached work-flow layer. Finally the presentation tier comprises a presentation adapter layer and an interface layer. Each layer is built up from small blocks which can be combined to create a wide range of more complex functionality. Each new object type developed gains access to a wealth of existing code functionality, while also free to adapt and extend this. The hardware structure is designed to provide complete redundancy, high availability and to handle high load. This document is composed of an abstract followed by the presentation transparencies. (authors)

  15. Metrology for ITER Assembly

    International Nuclear Information System (INIS)

    Bogusch, E.

    2006-01-01

    The overall dimensions of the ITER Tokamak and the particular assembly sequence preclude the use of conventional optical metrology, mechanical jigs and traditional dimensional control equipment, as used for the assembly of smaller, previous generation, fusion devices. This paper describes the state of the art of the capabilities of available metrology systems, with reference to the previous experience in Fusion engineering and in other industries. Two complementary procedures of transferring datum from the primary datum network on the bioshield to the secondary datum s inside the VV with the desired accuracy of about 0.1 mm is described, one method using the access directly through the ports and the other using transfer techniques, developed during the co-operation with ITER/EFDA. Another important task described is the development of a method for the rapid and easy measurement of the gaps between sectors, required for the production of the customised splice plates between them. The scope of the paper includes the evaluation of the composition and cost of the systems and team of technical staff required to meet the requirements of the assembly procedure. The results from a practical, full-scale demonstration of the methodologies used, using the proposed equipment, is described. This work has demonstrated the feasibility of achieving the necessary accuracies for the successful building of ITER. (author)

  16. The ITER tritium systems

    International Nuclear Information System (INIS)

    Glugla, M.; Antipenkov, A.; Beloglazov, S.; Caldwell-Nichols, C.; Cristescu, I.R.; Cristescu, I.; Day, C.; Doerr, L.; Girard, J.-P.; Tada, E.

    2007-01-01

    ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task

  17. Neutron cameras for ITER

    International Nuclear Information System (INIS)

    Johnson, L.C.; Barnes, C.W.; Batistoni, P.

    1998-01-01

    Neutron cameras with horizontal and vertical views have been designed for ITER, based on systems used on JET and TFTR. The cameras consist of fan-shaped arrays of collimated flight tubes, with suitably chosen detectors situated outside the biological shield. The sight lines view the ITER plasma through slots in the shield blanket and penetrate the vacuum vessel, cryostat, and biological shield through stainless steel windows. This paper analyzes the expected performance of several neutron camera arrangements for ITER. In addition to the reference designs, the authors examine proposed compact cameras, in which neutron fluxes are inferred from 16 N decay gammas in dedicated flowing water loops, and conventional cameras with fewer sight lines and more limited fields of view than in the reference designs. It is shown that the spatial sampling provided by the reference designs is sufficient to satisfy target measurement requirements and that some reduction in field of view may be permissible. The accuracy of measurements with 16 N-based compact cameras is not yet established, and they fail to satisfy requirements for parameter range and time resolution by large margins

  18. ITER concept definition. V.2

    International Nuclear Information System (INIS)

    1989-01-01

    Volume II of the two volumes describing the concept definition of the International Thermonuclear Experimental Reactor deals with the ITER concept in technical depth, and covers all areas of design of the ITER tokamak. Included are an assessment of the current database for design, scoping studies, rationale for concepts selection, performance flexibility, the ITER concept, the operations and experimental/testing program, ITER parameters and design phase schedule, and research and development specific to ITER. This latter includes a definition of specific research and development tasks, a division of tasks among members, specific milestones, required results, and schedules. Figs and tabs

  19. ITER CTA newsletter. No. 10

    International Nuclear Information System (INIS)

    2002-07-01

    This ITER CTA newsletter issue comprises the ITER backgrounder, which was approved as an official document by the participants in the Negotiations on the ITER Implementation agreement at their fourth meeting, held in Cadarache from 4-6 June 2002, and information about two ITER meetings: one is the third meeting of the ITER parties' designated Safety Representatives, which took place in Cadarache, France from 6-7 June 2002, and the other is the second meeting of the International Tokamak Physics Activity (ITPA) topical group on diagnostics, which was held at General Atomics, San Diego, USA, from 4-8 March 2002

  20. Micro-homogeneity of candidate reference materials: Results from an intercomparison study for the Analytical Quality Control Services (AQCS) of the IAEA

    International Nuclear Information System (INIS)

    Rossbach, M.; Kniewald, G.

    2002-01-01

    The IAEA Analytical Quality Control Services (AQCS) has made available two single cell algae materials IAEA-392 and IAEA-393 as well as an urban dust IAEA-396 to study their use for analytical sample sizes in the milligram range and below. Micro-analytical techniques such as PIXE and μ-PIXE, solid sampling AAS, scanning electron microprobe X-ray analysis and INAA were applied to the determination of trace elements on the basis of μg to mg amounts of the selected materials. The comparability of the mean values as well as the reproducibility of successive measurements is being evaluated in order to compare relative homogeneity factors for many elements in the investigated materials. From the reported results it seems that the algae materials IAEA-392 and IAEA-393 are extremely homogeneous biological materials for a number of elements with an extraordinary sharp particle size distribution below 10 μm. A similar situation seems to hold for the urban dust material IAEA-396 which had been air-jet milled to a particle size distribution around 4 μm. The introduction of these materials as CRMs with very small amounts needed to determine the certified concentrations will help to meet the needs of micro-analytical techniques for natural matrix reference materials. (author)

  1. ITER EDA newsletter. V. 7, no. 7

    International Nuclear Information System (INIS)

    1998-07-01

    This newsletter contains the articles: 'Extraordinary ITER council meeting', 'ITER EDA final safety meeting' and 'Summary report of the 3rd combined workshop of the ITER confinement and transport and ITER confinement database and modeling expert groups'

  2. Spirit and prospects of ITER

    Energy Technology Data Exchange (ETDEWEB)

    Velikhov, E.P. [Kurchatov Institute of Atomic Energy, Moscow (Russian Federation)

    2002-10-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  3. Spirit and prospects of ITER

    International Nuclear Information System (INIS)

    Velikhov, E.P.

    2002-01-01

    ITER is the unique and the most straightforward way to study the burning plasma science in the nearest future. ITER has a firm physics ground based on the results from the world tokamaks in terms of confinement, stability, heating, current drive, divertor, energetic particle confinement to an extend required in ITER. The flexibility of ITER will allow the exploration of broad operation space of fusion power, beta, pulse length and Q values in various operational scenarios. Success of the engineering R and D programs has demonstrated that all party has an enough capability to produce all the necessary equipment in agreement with the specifications of ITER. The acquired knowledge and technologies in ITER project allow us to demonstrate the scientific and technical feasibility of a fusion reactor. It can be concluded that ITER must be constructed in the nearest future. (author)

  4. Citizen Candidates Under Uncertainty

    OpenAIRE

    Eguia, Jon X.

    2005-01-01

    In this paper we make two contributions to the growing literature on "citizen-candidate" models of representative democracy. First, we add uncertainty about the total vote count. We show that in a society with a large electorate, where the outcome of the election is uncertain and where winning candidates receive a large reward from holding office, there will be a two-candidate equilibrium and no equilibria with a single candidate. Second, we introduce a new concept of equilibrium, which we te...

  5. ITER EDA newsletter. V. 10, special issue

    International Nuclear Information System (INIS)

    2001-07-01

    This ITER EDA Newsletter includes summaries of the reports of ITER EDA JCT Physics unit about ITER physics R and D during the Engineering Design Activities (EDA), ITER EDA JCT Naka JWC ITER technology R and D during the EDA, and Safety, Environment and Health group of ITER EDA JCT, Garching JWS on EDA activities related to safety

  6. ITER: the first experimental fusion reactor

    International Nuclear Information System (INIS)

    Rebut, P.H.

    1995-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (''the parties''). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world.The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500MW of fusion power.The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER.The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies.The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. (orig.)

  7. The JET ITER-like wall experiment: First results and lessons for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

    2013-10-15

    Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

  8. Fusion for Energy: The European joint undertaking for ITER and the development of fusion energy

    International Nuclear Information System (INIS)

    Diegele, E.

    2009-01-01

    Materials development in nuclear fusion for in-vessel components, i.e. for breeder blankets and divertors, has a history of more than two decades. It is the specific in-service and loading conditions and the consequentially required properties in combination with safety standards and social-economic demands that create a unique set of specifications. Objectives of Fusion for Energy (F4E) include: 1) To provide Europe's contribution to the ITER international fusion energy project; 2) To implement the Broader Approach agreement between Euratom and Japan; 3) To prepare for the construction and demonstration of fusion reactors (DEMO). Consequently, activities in F4E focus on structural materials for the first generations of breeder blankets, i.e. ITER Test Blanket Modules (TBM) and DEMO, whereas a Fusion Materials Topical Group implemented under EFDA coordinates R and D on physically based modelling of irradiation effects and R and D in the longer term (new and /or higher risk materials). The paper focuses on martensitic-ferritic steels and (i) reviews briefly the challenges and the rationales for the decisions taken in the past, (ii) analyses the status of the main activities of development and qualification, (iii) indicates unresolved issues, and (iv) outlines future strategies and needs and their implications. Due to the exposure to intense high energy neutron flux, the main issue for breeder materials is high radiation resistance. The First Wall of a breeder blanket should survive 3-5 full power years or, respectively in terms of irradiation damage, typically 50-70 dpa for DEMO and double figures for a power plant. Even though the objective is to have the materials and key fabrication technologies needed for DEMO fully developed and qualified within the next two decades, a major part of the task has to be completed much earlier. Tritium breeding test blanket modules will be installed in ITER with the objective to test DEMO relevant technologies in fusion

  9. The ITER divertor concept

    International Nuclear Information System (INIS)

    Janeschitz, G.; Borrass, K.; Federici, G.; Igitkhanov, Y.; Kukushkin, A.; Pacher, H.D.; Pacher, G.W.; Sugihara, M.

    1995-01-01

    The ITER divertor must exhaust most of the alpha particle power and the He ash at acceptable erosion rates. The high recycling regime of the ITER-CDA for present parameters would yield high power loads and erosion rates on conventional targets. Improvement by radiation in the SOL at constant pressure is limited in principle. To permit a higher radiation fraction, the plasma pressure along the field must be reduced by more than a factor 10, reducing also the target ion flux. This pressure reduction can be obtained by strong plasma-neutral interaction below the X-point. Under these conditions T e in the divertor can be reduced to <5 eV along a flame like ionisation front by impurity radiation and CX losses. Downstream of the front, neutrals undergo more CX or i-n collisions than ionisation events, resulting in significant momentum loss via neutrals to the divertor chamber wall. The pressure reduction by this mechanism depends on the along-field length for neutral-plasma interaction, the parallel power flux, the neutral density, the ratio of neutral-neutral collision length to the plasma-wall distance and on the Mach number of ions and neutrals. A supersonic transition in the main plasma-neutral interaction region, expected to occur near the ionisation front, would be beneficial for momentum removal. The momentum transfer fraction to the side walls is calculated: low Knudsen number is beneficial. The impact of the different physics effects on the chosen geometry and on the ITER divertor design and the lifetime of the various divertor components are discussed. ((orig.))

  10. Iteration of adjoint equations

    International Nuclear Information System (INIS)

    Lewins, J.D.

    1994-01-01

    Adjoint functions are the basis of variational methods and now widely used for perturbation theory and its extension to higher order theory as used, for example, in modelling fuel burnup and optimization. In such models, the adjoint equation is to be solved in a critical system with an adjoint source distribution that is not zero but has special properties related to ratios of interest in critical systems. Consequently the methods of solving equations by iteration and accumulation are reviewed to show how conventional methods may be utilized in these circumstances with adequate accuracy. (author). 3 refs., 6 figs., 3 tabs

  11. Divertor armour issues: lifetime, safety and influence on ITER performance

    International Nuclear Information System (INIS)

    Pestchanyi, S.

    2009-01-01

    Comprehensive simulations of the ITER divertor armour vaporization and brittle destruction under ELMs of different sizes have revealed that the erosion rate of CFC armour is intolerable for an industrial reactor, but it can be considerably reduced by the armour fibre structure optimization. The ITER core contamination with carbon is tolerable for medium size ELMs, but large type I ELM can run the confinement into the disruption. Erosion of tungsten, an alternative armour material, under ELMs influence is satisfactory, but the danger of the core plasma contamination with tungsten is still not enough understood and potentially it could be very dangerous. Vaporization of tungsten, its cracking and dust production during ELMs are rather urgent issues to be investigated for proper choice of the divertor armour material for ITER. However, the erosion rate under action of the disruptive heat loads is tolerable for both armour materials assuming few hundred disruptions falls out during ITER lifetime

  12. Feedback and rotational stabilization of resistive wall modes in ITER

    International Nuclear Information System (INIS)

    Liu Yueqiang; Bondeson, A.; Chu, M.S.; La Haye, R.J.; Favez, J.-Y.; Lister, J.B.; Gribov, Y.; Gryaznevich, M.; Hender, T.C.; Howell, D.F.

    2005-01-01

    Different models have been introduced in the stability code MARS-F in order to study the damping effect of resistive wall modes (RWM) in rotating plasmas. Benchmark of MARS-F calculations with RWM experiments on JET and D3D indicates that the semi-kinetic damping model is a good candidate for explaining the damping mechanisms. Based on these results, the critical rotation speeds required for RWM stabilization in an advanced ITER scenario are predicted. Active feedback control of the n = 1 RWM in ITER is also studied using the MARS-F code. (author)

  13. Development of ITER Tritium Storage Material

    Energy Technology Data Exchange (ETDEWEB)

    Chung, H. C.; Kim, K. R.; Paek, S. W.; Shim, M.; Noh, B

    2007-01-15

    The ZrCo getter beds are built of a primary vessel which contains the ZrCo powder and of a secondary outer vessel. The purpose of the secondary outer vessel is to capture permeated or leaked tritium and to present a good thermal insulation when properly evacuated. A third volume, a helium filled loop, is installed in the primary volume to remove the decay heat and is used to perform tritium accountancy measurements. In this report the authors verified that ZrCo can be used safely under a low pressure and temperature.

  14. iterClust: a statistical framework for iterative clustering analysis.

    Science.gov (United States)

    Ding, Hongxu; Wang, Wanxin; Califano, Andrea

    2018-03-22

    In a scenario where populations A, B1 and B2 (subpopulations of B) exist, pronounced differences between A and B may mask subtle differences between B1 and B2. Here we present iterClust, an iterative clustering framework, which can separate more pronounced differences (e.g. A and B) in starting iterations, followed by relatively subtle differences (e.g. B1 and B2), providing a comprehensive clustering trajectory. iterClust is implemented as a Bioconductor R package. andrea.califano@columbia.edu, hd2326@columbia.edu. Supplementary information is available at Bioinformatics online.

  15. (02.2) Scoping experiments; (02.3) long-term corrosion testing and properties evaluation of candidate waste package basket material

    International Nuclear Information System (INIS)

    VanKonynenburg, R. A.

    1996-01-01

    The work described in this activity plan addresses Information Need 2.7.3 of the Yucca Mountain Site Characterization Plan (l), which reads Determination that the design criteria in lOCFR60.130 through 60.133 and any appropriate additional design objectives pertaining to criticality control have been met. This work falls under section WBS 1.2.2.5 2 (Basket Materials) of WBS 1.2.2.5 (Waste Package Materials) in the Work Breakdown Structure of the Yucca Mountain Site Characterization Project

  16. Study of wall conditioning in tokamaks with application to ITER

    International Nuclear Information System (INIS)

    Kogut, Dmitri

    2014-01-01

    Thesis is devoted to studies of performance and efficiency of wall conditioning techniques in fusion reactors, such as ITER. Conditioning is necessary to control the state of the surface of plasma facing components to ensure plasma initiation and performance. Conditioning and operation of the JET tokamak with ITER-relevant material mix is extensively studied. A 2D model of glow conditioning discharges is developed and validated; it predicts reasonably uniform discharges in ITER. In the nuclear phase of ITER operation conditioning will be needed to control tritium inventory. It is shown here that isotopic exchange is an efficient mean to eliminate tritium from the walls by replacing it with deuterium. Extrapolations for tritium removal are comparable with expected retention per a nominal plasma pulse in ITER. A 1D model of hydrogen isotopic exchange in beryllium is developed and validated. It shows that fluence and temperature of the surface influence efficiency of the isotopic exchange. (author) [fr

  17. ITER assembly and maintenance

    International Nuclear Information System (INIS)

    Honda, T.; Davis, F.; Lousteau, D.

    1991-01-01

    This document is intended to describe the work conducted by the ITER Assembly and Maintenance (A and M) Design Unit and the supporting home teams during the ITER Conceptual Design Activities, carried out from 1988 through 1990. Its content consists of two main sections, i.e., Chapter III, which describes the identified tasks to be performed by the A and M system and a general description of the required equipment; and Chapter IV, which provides a more detailed description of the equipment proposed to perform the assigned tasks. A two-stage R and D program is now planned, i.e., (1) a prototype equipment functional tests using full scale mock-ups and (2) a full scale integration demonstration test facility with real components (vacuum vessel with ports, blanket modules, divertor modules, armor tiles, etc.). Crucial in-vessel and ex-vessel operations and the associated remote handling equipment, including handling of divertor plates and blanket modules will be demonstrated in the first phase, whereby the database needed to proceed with the engineering phase will be acquired. The second phase will demonstrate the ability of the overall system to execute the required maintenance procedures and evaluate the performance of the prototype equipment

  18. A single-source precursor route to anisotropic halogen-doped zinc oxide particles as a promising candidate for new transparent conducting oxide materials

    Directory of Open Access Journals (Sweden)

    Daniela Lehr

    2015-11-01

    Full Text Available Numerous applications in optoelectronics require electrically conducting materials with high optical transparency over the entire visible light range. A solid solution of indium oxide and substantial amounts of tin oxide for electronic doping (ITO is currently the most prominent example for the class of so-called TCOs (transparent conducting oxides. Due to the limited, natural occurrence of indium and its steadily increasing price, it is highly desired to identify materials alternatives containing highly abundant chemical elements. The doping of other metal oxides (e.g., zinc oxide, ZnO is a promising approach, but two problems can be identified. Phase separation might occur at the required high concentration of the doping element, and for successful electronic modification it is mandatory that the introduced heteroelement occupies a defined position in the lattice of the host material. In the case of ZnO, most attention has been attributed so far to n-doping via substitution of Zn2+ by other metals (e.g., Al3+. Here, we present first steps towards n-doped ZnO-based TCO materials via substitution in the anion lattice (O2− versus halogenides. A special approach is presented, using novel single-source precursors containing a potential excerpt of the target lattice 'HalZn·Zn3O3' preorganized on the molecular scale (Hal = I, Br, Cl. We report about the synthesis of the precursors, their transformation into halogene-containing ZnO materials, and finally structural, optical and electronic properties are investigated using a combination of techniques including FT-Raman, low-T photoluminescence, impedance and THz spectroscopies.

  19. Rokkasho: Japanese site for ITER

    International Nuclear Information System (INIS)

    Ohtake, S.; Yamaguchi, V.; Matsuda, S.; Kishimoto, H.

    2003-01-01

    The Atomic Energy Commission of Japan authorized ITER as the core machine of the Third Phase Basic Program of Fusion Energy Development. After a series of discussions in the Atomic Energy Commission and the Council of Science and Technology Policy, Japanese Government concluded formally with the Cabinet Agreement on 31 May 2002 that Japan should participate in the ITER Project and offer the Rokkasho-Mura site for construction of ITER to the Negotiations among Canada (CA), the European Union (EU), Japan (JA), and the Russian Federation (RF). The JA site proposal is now under the international assessment in the framework of the ITER Negotiations. (author)

  20. IAEA activities related to ITER

    International Nuclear Information System (INIS)

    Dolan, T.J.; Schneider, U.

    2001-01-01

    As agreed between the IAEA and the ITER Parties, special sessions are dedicated to ITER at the IAEA Fusion Energy Conferences. At the 18th IAEA Fusion Energy Conference, held on 4-10 October 2000 in Sorrento, Italy, in the Artsimovich-Kadomtsev Memorial opening session there were special lectures by Carlo Rubbia (President, ENEA, Italy), A. Arima (Japan), and E.P. Velikhov (Russia); an overview talk on ITER by R. Aymar (ITER Director); and a talk on the FTU experiment by F. Romanelli. In total, 573 participants from 34 countries presented 389 papers (including 11 post-deadline papers and the 4 summaries)

  1. ITER CTA newsletter. No. 13, October 2002

    International Nuclear Information System (INIS)

    2002-11-01

    This ITER CTA newsletter issue comprises concise information about an ITER related meeting concerning the joint implementation of ITER - the fifth ITER Negotiations Meeting - which was held in Toronto, Canada, 19-20 September, 2002, and information about assessment of the possible ITER site in Clarington, Ontario, Canada, which was the subject of the first official stage of the Joint Assessment of Specific Sites (JASS) for the ITER Project. This assessment was completed just before the Fifth ITER Negotiations Meeting

  2. Status and plans for U.S. ITER studies

    International Nuclear Information System (INIS)

    Doggett, J.N.

    1992-01-01

    The United States' participation in the International Thermonuclear Experimental Reactor (ITER) began in late 1987 when the initiative to start a cooperative program among the four Parties-the Soviet Union, Japan, the European Community, and the United States-was initiated. Participation then continued through the start of joint Work in May 1988 until the conclusion of the Conceptual Design Activities (CDA) in December 1990. In the period between the conclusion of the CDA and the agreement to execute the Engineering Design Activities (EDA), the U.S. ITER Home Team continued to do work on the design, executed additional research and development (R and D) and participated in the preparations for the EDA. Activities included one major design study on a High-Aspect-Ratio Design (HARD) and input to the National ITER Technical Review, the ITER Steering Committee-U.S. (ISCUS), Special Working Group 1 (SWG-1), and the Fusion Energy Advisory Committee's Panel 1 (FEAC-1). Research and development was continued in areas of work that were identified as critical-path elements by an international panel chartered by the four ITER Parties near the end of the CDA. During the interim period, the U.S. Home Team Management (HTM) was in the process of organizing to support the EDA both at home and in the central design sites. The major efforts have been in producing a management plan, establishing memorandums of agreement with the performing institutions for ITER tasks, establishing an industrial council, and producing a list of candidates who are qualified, willing, and available to serve on the joint Central Team or to participate in ITER home tasks. The author describes the conclusion of the CDA and the interim U.S. ITER activities and will give an indication of US involvement in the EDA

  3. ITER waste management

    International Nuclear Information System (INIS)

    Rosanvallon, S.; Na, B.C.; Benchikhoune, M.; Uzan, J. Elbez; Gastaldi, O.; Taylor, N.; Rodriguez, L.

    2010-01-01

    ITER will produce solid radioactive waste during its operation (arising from the replacement of components and from process and housekeeping waste) and during decommissioning (de-activation phase and dismantling). The waste will be activated by neutrons of energies up to 14 MeV and potentially contaminated by activated corrosion products, activated dust and tritium. This paper describes the waste origin, the waste classification as a function of the French national agency for radioactive waste management (ANDRA), the optimization process put in place to reduce the waste radiotoxicity and volumes, the estimated waste amount based on the current design and maintenance procedure, and the overall strategy from component removal to final disposal anticipated at this stage of the project.

  4. Iterated crowdsourcing dilemma game

    Science.gov (United States)

    Oishi, Koji; Cebrian, Manuel; Abeliuk, Andres; Masuda, Naoki

    2014-02-01

    The Internet has enabled the emergence of collective problem solving, also known as crowdsourcing, as a viable option for solving complex tasks. However, the openness of crowdsourcing presents a challenge because solutions obtained by it can be sabotaged, stolen, and manipulated at a low cost for the attacker. We extend a previously proposed crowdsourcing dilemma game to an iterated game to address this question. We enumerate pure evolutionarily stable strategies within the class of so-called reactive strategies, i.e., those depending on the last action of the opponent. Among the 4096 possible reactive strategies, we find 16 strategies each of which is stable in some parameter regions. Repeated encounters of the players can improve social welfare when the damage inflicted by an attack and the cost of attack are both small. Under the current framework, repeated interactions do not really ameliorate the crowdsourcing dilemma in a majority of the parameter space.

  5. ITER cooling systems

    International Nuclear Information System (INIS)

    Natalizio, A.; Hollies, R.E.; Sochaski, R.O.; Stubley, P.H.

    1992-06-01

    The ITER reference system uses low-temperature water for heat removal and high-temperature helium for bake-out. As these systems share common equipment, bake-out cannot be performed until the cooling system is drained and dried, and the reactor cannot be started until the helium has been purged from the cooling system. This study examines the feasibility of using a single high-temperature fluid to perform both heat removal and bake-out. The high temperature required for bake-out would also be in the range for power production. The study examines cost, operational benefits, and impact on reactor safety of two options: a high-pressure water system, and a low-pressure organic system. It was concluded that the cost savings and operational benefits are significant; there are no significant adverse safety impacts from operating either the water system or the organic system; and the capital costs of both systems are comparable

  6. Divertor development for ITER

    International Nuclear Information System (INIS)

    Janeschitz, G.; Ando, T.; Antipenkov, A.; Barabash, V.; Chiocchio, S.; Federici, G.; Ibbott, C.; Jakeman, R.; Matera, R.; Martin, E.; Parker, R.; Tivey, R.; Pacher, H.D.

    1998-01-01

    The requirements for the ITER divertor design, i.e. power and He ash exhaust, neutral leakage control, lifetime, disruption load resistance and exchange by remote handling, are described in this paper. These requirements and the physics requirements for detached and semi-attached operation result in the vertical target configuration. This is realised by a concept incorporating 60 cassettes carrying the high heat flux components. The armour choice for these components is CFC monoblock in the strike zone near at the lower part of the vertical target, and a W brush elsewhere. Cooling is by swirl tubes or hypervapotrons depending on the component. The status of the heat sink and joining technology R and D is given. Finally, the resulting design of the high heat flux components is presented. (orig.)

  7. Joint assessment of specific sites for ITER begins at Clarington

    International Nuclear Information System (INIS)

    Stewart, M.J.

    2002-01-01

    Clarington, Ontario, Canada was the subject of the first official stage of the Joint Assessment of Specific Sites (JASS) for the ITER Project. The Assessment is part of the Negotiations process and is being conducted by an ad-hoc group of the Negotiators with representatives from Canada, the European Union, Japan and Russian Federation, supported by the ITER international team. The evaluation was conducted over four days through a series of visits to the site itself, a review of materials included in Canada's submission to host ITER, presentations from group leading Canada's offer and experts on specific aspects of the offer

  8. ITER-FEAT - outline design report. Report by the ITER Director. ITER meeting, Tokyo, January 2000

    International Nuclear Information System (INIS)

    2001-01-01

    It is now possible to define the key elements of ITER-FEAT. This report provides the results, to date, of the joint work of the Special Working Group in the form of an Outline Design Report on the ITER-FEAT design which, subject to the views of ITER Council and of the Parties, will be the focus of further detailed design work and analysis in order to provide to the Parties a complete and fully integrated engineering design within the framework of the ITER EDA extension

  9. Standardization, Calibration, and Evaluation of Tantalum-Nano rGO-SnO2 Composite as a Possible Candidate Material in Humidity Sensors

    Science.gov (United States)

    Karthick, Subbiah; Lee, Han-Seung; Kwon, Seung-Jun; Natarajan, Rethinam; Saraswathy, Velu

    2016-01-01

    The present study focuses the development and the evaluation of humidity sensors based on reduced graphene oxide—tin oxide (rGO-SnO2) nanocomposites, synthesized by a simple redox reaction between GO and SnCl2. The physico-chemical characteristics of the nanocomposites were analyzed by XRD, TEM, FTIR, and Raman spectroscopy. The formation of SnO2 crystal phase was observed through XRD. The SnO2 crystal phase anchoring to the graphene sheet was confirmed through TEM images. For the preparation of the sensors, tantalum substrates were coated with the sensing material. The sensitivity of the fabricated sensor was studied by varying the relative humidity (RH) from 11% to 95% over a period of 30 days. The dependence of the impedance and of the capacitance with RH of the sensor was measured with varying frequency ranging from 1 kHz to 100 Hz. The long-term stability of the sensor was measured at 95% RH over a period of 30 days. The results proved that rGO-SnO2 nanocomposites are an ideal conducting material for humidity sensors due to their high sensitivity, rapid response and recovery times, as well as their good long-term stability. PMID:27941598

  10. Standardization, Calibration, and Evaluation of Tantalum-Nano rGO-SnO2 Composite as a Possible Candidate Material in Humidity Sensors

    Directory of Open Access Journals (Sweden)

    Subbiah Karthick

    2016-12-01

    Full Text Available The present study focuses the development and the evaluation of humidity sensors based on reduced graphene oxide—tin oxide (rGO-SnO2 nanocomposites, synthesized by a simple redox reaction between GO and SnCl2. The physico-chemical characteristics of the nanocomposites were analyzed by XRD, TEM, FTIR, and Raman spectroscopy. The formation of SnO2 crystal phase was observed through XRD. The SnO2 crystal phase anchoring to the graphene sheet was confirmed through TEM images. For the preparation of the sensors, tantalum substrates were coated with the sensing material. The sensitivity of the fabricated sensor was studied by varying the relative humidity (RH from 11% to 95% over a period of 30 days. The dependence of the impedance and of the capacitance with RH of the sensor was measured with varying frequency ranging from 1 kHz to 100 Hz. The long-term stability of the sensor was measured at 95% RH over a period of 30 days. The results proved that rGO-SnO2 nanocomposites are an ideal conducting material for humidity sensors due to their high sensitivity, rapid response and recovery times, as well as their good long-term stability.

  11. ITER CTA newsletter. No. 8

    International Nuclear Information System (INIS)

    2002-05-01

    This ITER CTA newsletter contains information about the Third Negotiations Meeting on the Joint Implementation of ITER held in Moscow on 23-24 April 2002 and about the visit of Canadian officials and members of the Canadian delegation to RF research center 'Kurchatov Institute'

  12. ITER physics design guidelines: 1989

    International Nuclear Information System (INIS)

    Uckan, N.A.

    1990-01-01

    The physics basis for ITER has been developed from an assessment of the results of the last twenty-five years of tokamak research and from detailed analysis of important physics issues specifically for the ITER design. This assessment has been carried out with direct participation of members of the experimental teams of each of the major tokamaks in the world fusion program through participation in ITER workshops, contributions to the ITER Physics R and D Program, and by direct contacts between the ITER team and the cognizant experimentalists. Extrapolations to the present data base, where needed, are made in the most cautious way consistent with engineering constraints and performance goals of the ITER. In cases where a working assumptions had to be introduced, which is insufficiently supported by the present data base, is explicitly stated. While a strong emphasis has been placed on the physics credibility of the design, the guidelines also take into account that ITER should be designed to be able to take advantage of potential improvements in tokamak physics that may occur before and during the operation of ITER. (author). 33 refs

  13. ITER management advisory committee meeting

    International Nuclear Information System (INIS)

    Yoshikawa, M.

    2001-01-01

    The ITER Management Advisory Committee (MAC) Meeting was held in Vienna on 16 July 2001. It was the last MAC Meeting and the main topics were consideration of the report by the Director on the ITER EDA status, review of the Work Programme, review of the Joint Fund and arrangements for termination and wind-up of the EDA

  14. ITER CTA newsletter. No. 7

    International Nuclear Information System (INIS)

    2002-04-01

    This issue of ITER CTA newsletter contains information about the meeting of the ITER CTA project board, which took place in Moscow, Russian Federation on 22 April 2002 on the occasion of the Third Negotiators Meeting (N3), and about the meeting 'EU divertor celebration day' organized on 16 January 2002 at Plansee AG, Reutte, Austria

  15. Are trinuclear superhalogens promising candidates for building blocks of novel magnetic materials? A theoretical prospect from combined broken-symmetry density functional theory and ab initio study.

    Science.gov (United States)

    Yu, Yang; Li, Chen; Yin, Bing; Li, Jian-Li; Huang, Yuan-He; Wen, Zhen-Yi; Jiang, Zhen-Yi

    2013-08-07

    The structures, relative stabilities, vertical electron detachment energies, and magnetic properties of a series of trinuclear clusters are explored via combined broken-symmetry density functional theory and ab initio study. Several exchange-correlation functionals are utilized to investigate the effects of different halogen elements and central atoms on the properties of the clusters. These clusters are shown to possess stronger superhalogen properties than previously reported dinuclear superhalogens. The calculated exchange coupling constants indicate the antiferromagnetic coupling between the transition metal ions. Spin density analysis demonstrates the importance of spin delocalization in determining the strengths of various couplings. Spin frustration is shown to occur in some of the trinuclear superhalogens. The coexistence of strong superhalogen properties and spin frustration implies the possibility of trinuclear superhalogens working as the building block of new materials of novel magnetic properties.

  16. ITER EDA Newsletter. V. 3, no. 8

    International Nuclear Information System (INIS)

    1994-08-01

    This ITER EDA (Engineering Design Activities) Newsletter issue reports on the sixth ITER council meeting; introduces the newly appointed ITER director and reports on his address to the ITER council. The vacuum tank for the ITER model coil testing, installed at JAERI, Naka, Japan is also briefly described

  17. ITER ITA newsletter. No. 6, July 2003

    International Nuclear Information System (INIS)

    2003-09-01

    This issue of ITER ITA (ITER transitional Arrangements) newsletter contains concise information about ITER related activities. One of them was the farewell party for for Annick Lyraud and Robert Aymar, who will take up his position as Director-General of CERN in January 2004, another is information about Dr. Yasuo Shimomura, ITER interim project leader, and ITER technical work during the transitional arrangements

  18. ITER interim design report package documents

    International Nuclear Information System (INIS)

    1996-01-01

    This publication contains the Excerpt from the ITER Council (IC-8), the ITER Interim Design Report, Cost Review and Safety Analysis, ITER Site Requirements and ITER Site Design Assumptions and the Excerpt from the ITER Council (IC-9). 8 figs, 2 tabs

  19. Fabrication progress of the ITER vacuum vessel sector in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Kim, B.C., E-mail: bckim@nfri.re.kr [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Lee, Y.J.; Hong, K.H.; Sa, J.W.; Kim, H.S.; Park, C.K.; Ahn, H.J.; Bak, J.S.; Jung, K.J. [National Fusion Research Institute, Gwahangno 113, Yuseong-gu, Daejeon (Korea, Republic of); Park, K.H.; Roh, B.R.; Kim, T.S.; Lee, J.S.; Jung, Y.H.; Sung, H.J.; Choi, S.Y.; Kim, H.G.; Kwon, I.K.; Kwon, T.H. [Hyundai Heavy Industries Co. Ltd., Dong-gu, Ulsan (Korea, Republic of)

    2013-10-15

    Highlights: ► Fabrication of ITER vacuum vessel sector full scale mock-up to develop fabrication procedures. ► The welding and nondestructive examination techniques conform to RCC-MR. ► The preparation of real manufacturing of ITER vacuum vessel sector. -- Abstract: As a participant of ITER project, ITER Korea has to supply two ITER vacuum vessel sectors (Sector no. 6, no. 1) of total nine ITER VV sectors. After the procurement arrangement with ITER Organization, ITER Korea made the contract with Hyundai Heavy Industries (HHI) for fabrication of two sectors. Then the start of the manufacturing design was initiated from January 2010. HHI made three real scale R and D mock-ups to verify the critical fabrication feasibility issues on electron beam welding, 3D forming, welding distortion and achievable tolerances. The documentation according to IO and the French nuclear safety regulation requirement, the qualification of welding and nondestructive examination procedures conform to RCC-MR 2007 were proceed in parallel. The mass production of raw material was done after receiving ANB (agreed notified body) verification of product/parts and shop qualification. The manufacturing drawing, manufacturing and inspection plan of VV sector with supporting fabrication procedures are also verified by ANB, accordingly the first cutting and forming of plates for VV sector fabrication started from February 2012. This paper reports the latest fabrication progress of ITER vacuum vessel Sector no. 6 that will be assembled as the first sector in the ITER pit. The overall fabrication route, R and D mock-up fabrication results with forming and welding distortion analysis, qualification status of welding and nondestructive examination (NDE) are also presented.

  20. ITER - Progress in design and validating R and D. Note by the Director. ITER meeting, Moscow, 28 June 2000

    International Nuclear Information System (INIS)

    2001-01-01

    This note summarises, for the ITER Council, material presented in detail to TAC, in the report ''Progress in Resolving Open Design Issues from the ODR'', including: further investigation of physics issues raised in the TAC report or in the course of the Parties' domestic assessments; and features of the ITER design which reflect a resolution of choice of options. In addition, the note summarises and illustrates recent progress in the programme of validating technology R and D, as presented in the report to TAC, ''ITER Technology R and D Progress Report''